HomeMy WebLinkAbout740476.tiff UNITED STATES ST TE OF COLORADO
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ATOMIC ENERGY COMMISSION COUNTY OF WELD
Filed with the Clcrk of the LOU
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• WASHINGTON,D.C. 20545 of County Commissioners
j�' er 1 4 1974
it
e" Y CLERK AND EC04DER
By Deputy
PUBLIC SERVICE COMPANY OF COLORADO H
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET NO. 50-267
FACILITY OPERATING LICENSE
License No. DPR-34
1. The Atomic Energy Commission (the Commission) having found that:
A. The application for license filed by the Public Service Company
of Colorado (the licensee) complies with the standards and
requirements of the Atomic Energy Act of 1954, as amended (the
Act) , and the Commission's rules and regulations set forth in
10 CFR Chapter I and all required notifications to other agencies
or bodies have been duly made;
B. Construction of the Fort St. Vrain Nuclear Generating Station
(the facility) has been substantially completed in conformity
with Provisional Construction Permit No. CPPR-54 and the appli-
cation, as amended, the provisions of the Act, and the rules
and regulations of the Commission;
C. The facility will operate in conformity with the application, as
amended, the provisions of the Act, and the rules and regulations
of the Commission;
D. There is reasonable assurance: (i) that the activities authorized
by this operating license can be conducted without endangering
the health and safety of the public, and (ii) that such activities
will be conducted in compliance with the rules and regulations
of the Commission;
E. The licensee is technically and financially qualified to engage
in the activities authorized by this operating license in accord-
ance with the rules and regulations of the Commission;
F. The licensee has satisfied the applicable provisions of 10 CFR
Part 140, "Financial Protection Requirements and Indemnity
Agreements," of the Commission's regulations;
740476
E4100 / / et2.—1—
- 2 -
G. The issuance of this operating license will not be inimical to
the common defense and security or to the health and safety of
the public;
H. After weighing the environmental, economic, technical, and other
benefits of the facility against environmental costs and consider-
ing available alternatives, the issuance of Facility Operating
License No. DPR-34 (subject to the conditions for protection of
the environment set forth herein) is in accordance with 10 CFR
Part 50, Appendix D, of the Commission's regulations and all
applicable requirements of said Appendix D have been satisfied;
and
I. The receipt, possession, and use of source, byproduct and special
nuclear material as authorized by this license will be in accord-
ance with the Commission's regulations in 10 CFR Parts 30, 40, 70,
and 73.
2. Facility Operating License No. DPR-34 is hereby issued to the Public
Service Company of Colorado to read as follows:
A. This license applies to the Fort St. Vrain Nuclear Generating
Station, a high temperature gas-cooled nuclear reactor and asso-
ciated equipment (the facility) owned by the Public Service
Company of Colorado. The facility is located near Platteville
in Weld County, Colorado, and is described in the "Final Safety
Analysis Report" as supplemented and amended (Amendments 15
through 29) and the Environmental Report as supplemented and
amended (Supplements 1 through 3).
B. This license is subject, for the initial rise to power, to the
conditions set forth in Specification LCO 4.9-1 of the Technical
Specifications attached hereto as Appendix A.
C. Subject to the conditions and requirements incorporated herein,
the Commission hereby licenses the Public Service Company of
Colorado:
(1) Pursuant to Section 104b of the Act and 10 CFR Part 50,
"Licensing of Production and Utilization Facilities," to
possess, use and operate the facility at the designated
location near Platteville in Weld County, Colorado, in
accordance with the procedures and limitations set forth
in this license;
- 3 -
(2) Pursuant to the Act, 10 CFR Part 70, "Special Nuclear
Material," and 10 CFR Part 73, "Physical Protection of
Special Nuclear Material," to receive, possess, and use
at any one time in connection with operation of the
facility:
a. Up to 1700 kilograms of contained uranium 235,
b. Up to 3 curies of plutonium contained in encapsulated
plutonium-beryllium neutron sources,
8,
c. 0ne
Pu-239, -232a, U-233,calibrationsources
U234andU-235,Peach6 3
source not
to exceed 10 microcuries;
(3) Pursuant to the Act and 10 CFR Part 40, "Licensing of
Source Material," to receive, possess and use at any one
time up to 25,000 kilograms of natural thorium in connec-
tion with operation of the facility;
(4) Pursuant to the Act and 10 CFR Part 30, "Rules of General
Applicability to Licensing of Byproduct Material," to
receive, possess and use in connection with operation of
the facility:
a. The following without restriction as to chemical and/or
physical form:
1. Any byproduct material with Atomic Numbers 1
through 83, inclusive, not to exceed 5 millicuries
per radionuclide;
2. Americium 241, not to exceed 2.01 curies;
3. Americium 243, not to exceed 5 millicuries;
4. Cesium 137, not to exceed 11 curies;
5. Hydrogen 3, not to exceed 15 curies;
6. Krypton 85, not to exceed 100 millicuries;
7. Neptunium 237, not to exceed 5 millicuries;
b. Californium 252, 3 milligrams as sealed sources, not to
exceed 0.5 curie per source;
- 4 -
(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess,
but not separate, such byproduct and special nuclear
materials as may be produced by the operation of the
facility.
D. This license shall be deemed to contain and is subject to the
conditions specified in the following Commission regulations in
10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section
40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and the
appropriate sections of Parts 70 and 73; is subject to all
applicable provisions of the Act and to the rules, regulations,
and orders of the Commission now or hereafter in effect; and is
subject to the additional conditions specified or incorporated
below:
(1) Maximum Power Level
The licensee is authorized to operate the facility at
steady state reactor core power levels not in excess of
842 megawatts thermal.
(2) Technical Specifications
The Technical Specifications defining safety and environ-
mental conditions contained in Appendices A and B attached
hereto are hereby incorporated in this license. The
licensee shall operate the facility in accordance with
these Technical Specifications. They include the following
conditions for the protection of the environment which were
set forth in the Final Environmental Statement:
a. Stream temperatures will be monitored in accordance with
the conditions set forth in Specification SR 1.2 of
Appendix B.
b. Monitoring of chemical concentrations will be performed
in accordance with the conditions set forth in
Specifications LCO 1.1 and SR 1.1 of Appendix B.
c. Operational radiological monitoring will be performed
in accordance with Specification SR 5.9.1 of
Appendix A.
d. Ecological studies will be performed in accordance with
the conditions set forth in Specification SR 2.1 of
Appendix B.
- 5 -
e. The discharge of all demineralizer regeneration
effluents will be made in accordance with the con-
ditions set forth in Specification LCO 1.1 of
Appendix B.
f. The total quantity of effluent originating as cooling
tower blowdown will be limited in accordance with the
conditions set forth in Specification LCO 1.3 of
Appendix B.
g. When the discharge temperature of cooling tower
blowdown exceeds 80°F, the discharge will be made
in accordance with the conditions set forth in
Specification LCO 1.2 of Appendix B.
h. Cooling tower blowdown will he discharged in
accordance with the conditions set forth in
Specification LCO 1.1 of Appendix B.
E. This license is subject to all Federal, State, and local standards
imposed pursuant to the requirements of the Federal Water Pollution
Control Act of 1972.
4. This license is effective as of the date of issuance and shall expire
at midnight, September 17, 2008.
FOR THE ATOMIC ENERGY COMMISSION
Original signed by:
A. Giambusso
A. Giambusso, Deputy Director
for Reactor Projects
Directorate of Licensing
Attachments:
Appendices A and B - Technical Specifications
Date of Issuance: DEC 2 1 1973
�1%I - UNITED STATES
= ATOMIC ENERGY COMMISSION
WASHINGTON. D.C. 20545
December 27 , 1973
Docket No. 50-267
Mr. Glenn K. Billings , Chairman
Board of County Commissioners of frt., ,,,,,,, ,.
Weld County, Colorado
Greeley, Colorado 80631 11
': 1Subject: Public Service Company of Colorado
(Fort St. Vrain Nuclear Generating Station) ' - .. •ir 5,
The following documents concerning our review of the subject facility REEDY. coLoe ,,'1
are transmitted for your information: ••
Notice of Receipt of Application.
❑ Draft Environmental Statement, dated
Final Environmental Statement, dated
❑ Safety Evaluation, or Supplement No. , dated
❑ Notice of Hearing on Application for Construction Permit.
Notice of Consideration of Issuance of Facility Operating License.
5 Application and Safety Analysis Report, Vol. •
❑ Amendment No. to Application/SAR, dated
❑ Construction Permit No. CPPR- , dated
Facility Operating License No. DPR- 34 , dated 12-21-73
Xl Technical Specifications, or Change No. , dated 12-21-73
❑ Other:
Directorate of Licensing
Enclosures:
As stated
cc:
APPENDIX A
TO
OPERATING LICENSE NO. DPR - 34
TECHNICAL SPECIFICATIONS
FOR THE
FORT ST. VRAIN NUCLEAR GENERATING STATION
PUBLIC SERVICE COMPANY OF COLORADO
DOCKET NO. 50-267
Date of Issuance: DEC 2 1 1973
FORT ST. VRAIN NUCLEAR GENERATING STATION
TECHNICAL SPECIFICATIONS
TABLE OF CONTENTS
PAGE
1.0 INTRODUCTION 1-1
2.0 DEFINITIONS 2-1
3.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 3.0-1
3.1 REACTOR CORE - SAFETY LIMIT 3.1-1
Specification SL 3.1 - Reactor Core Safety Limit 3.1-1
3.2 REACTOR VESSRT PRESSURE - SAFETY LIMIT 3.2-1
Specification SL 3.2 - Reactor Vessel Pressure
Safety Limit 3.2-1
3.3 LIMITING SAFglY SYSTEM SETTINGS 3.3-1
Specification LSSS 3.3 - Limiting Safety System
Settings 3.3-1
4.0 LIMITING CONDITIONS FOR OPERATION 4.0-1
4.1 REACTOR CORE AND REACTIVITY CONTROL - LIMITING
CONDITIONS FOR OPERATION 4.1-1
Specification LCO 4.1.1 - Core Irradiation 4.1-1
Specification LCO 4.1.2 - Operable Control Rods 4.1-2
Specification LCO 4.1.3 - Rod Sequence 4.1-3
Specification LCO 4.1.4 - Partially Inserted Rods 4.1-7
Specification LCO 4.1.5 - Reactivity Change with
Temperature 4.1-8
Specification LCO 4.1.6 - Reserve Shutdown System 4.1-10
Specification LCO 4.1.7 - Core Inlet Orifice Valves 4.1-11
Specification LCO 4.1.8 - Reactivity Status 4.1-13
Specification LCO 4.1.9 - Core Region Temperature Rise 4.1-14
i
PAGE
4.2 PRIMARY COOLANT SYSTEM - LIMITING CONDITIONS 4,2_1
FOR OPERATION
4.2-1
Specification LCO 4.2.1 - Number of Operable Circulators 4 2-2
Specification LCO 4.2.2 - Operable Circulator 4 2-2
Specification LCO 4.2.3 - Turbine Water Removal Pump 4.2-3
Specification LCO 4.2.4 - Service Water Pumps 4 2-3
Specification LCO 4.2.5 - Circulating Water Makeup System 4 2-3
Specification LCO 4.2.6 - Firewater Pumps 4 2-4
Specification LCO 4.2.7 - PCRV Pressurization 4 2-4
Specification LCO 4.2.8 - Primary Coolant Activity 4.2-11
Specification LCO 4.2.9 - PCRV Closure Seals
Specification LCO 4.2.10 - Loop Impurity Levels, 4.2-13
High Temperatures
Specification LCO 4.2.11 - Loop Impurity Levels, 4.2-13
Low Temperatures 4.2-13
Specification LCO 4.2.12 - Liquid Nitrogen Storage 4.2-14
Specification LCO 4.2.13 - PCRV Liner Cooling System 4.2-15
Specification LCO 4.2.14 - PCRV Liner Cooling Tubes
Specification LCO 4.2.15 - PCRV Cooling Water System 4.2-16
Temperatures
4.3 SECONDARY REACTOR COOLANT SYSTEM - LIMITING CONDITIONS 4.3-1
FOR OPERATION 4.3-1
Specification LCO 4.3.1 - Steam Generators 4 3-1
Specification LCO 4.3.2 - Boiler Feed Pumps 4.3-2
Specification LCO 4.3.3 - Steam/Water Dump Tank Inventory
Specification LCO 4.3.4 - Emergency Condensate and Emergency 4,3-3
Feedwater Headers 4.3-3
Specification LCO 4.3.5 - Storage Ponds 4.3-4
Specification LCO 4.3.6 - Instrument Air System 4.3-4
Specification LCO 4.3.7 - Hydraulic Power System 4 3-5
Specification LCO 4.3.8 - Secondary Coolant Activity
4.4 INSTRUMENTATION AND CONTROL SYSTEMS - LIMITING 4.4-1
CONDITIONS FOR OPERATION
Specification LCO 4.4.1 - Plant Protective System 4.4-1
Instrumentation 4.4-13
Specification LCO 4.4.2 - Control Room Temperature 4.4-13
Specification LCO 4.4.3 - Area Radiation Monitors 4.4-15
3
Specification LCO 4.4.4 - Seismic Instrumentation
it
PAGE
4.5 CONFINEMENT SYSTEM - LIMITING CONDITIONS FOR 4.5-1
OPERATION
Specification LCO 4.5.1 - Reactor Building
4.5-1
Specification LCO 4.5.2 - Reactor Vessel Internal 4.5-3
Maintenance
4.6 AUXILIARY ELECTRIC POWER SYSTEM - LIMITING CONDITIONS 4.6-1
FOR OPERATION
Specification LCO 4.6.1 - Auxiliary Electric System
4.6-1
4.7 FUEL HANDLING AND STORAGE SYSTEMS - LIMITING 4.7-1
CONDITIONS FOR OPERATION
4.7-1
Specification LCO 4.7.1 - Fuel Handling in the Reactor 4 7_2
Specification LCO 4.7.2 - Fuel Handling Machine 4.7-2
Specification LCO 4.7.3 - Fuel Storage Facility yontainer 4.7-4
Specification LCO 4.7.4 - Spent Fuel Shipping
4.8 RADIOACTIVE EFFLUENT DISPOSAL SYSTEM - LIMITING 4.8-1
CONDITIONS FOR OPERATION
Specification LCO 4.8.1 - Radioactive Gaseous Effluent 4.8-1
Specification LCO 4.8.2 - Radioactive Liquid Effluent
4.8-5
Specification LCO 4.8.3 - Reactor Building Sump Effluent . 4.8-7
4.9 FUEL LOADING AND INITIAL RISE TO POWER - LIMITING CONDITIONS FOR OPERATION 4.9-1
Specification LCO 4.9.1 - Fuel Loading and Initial Rise 4.9-1
to Power
5.0 SURVEILLANCE REQUIREMENTS
5.0-1
5.1 REACTOR CORE AND REACTIVITY CONTROL - SURVEILLANCE 5.1-1
REQUIREMENTS
5.1-1
Specification SR 5.1.1- Control Rod Drives 5.1-2
Specification SR 5.1.2 - Reserve Shutdown System 5.1-4
Specification SR 5.1.3 - Temperature Coefficient 5.1-4
Specification SR 5.1.4 - Reactivity Status 5.1-5
Specification SR 5.1.5 - Withdrawn Rod Reactivity 5.1-5
Specification SR 5.1.6 - Core Safety Limit
iii
PAGE
5.2 PRIMARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS 5.2-1
Specification SR 5.2.1 - PCRV Overpressure Safety System 5.2-1
Specification SR 5.2.2 - Tendon Corrosion 5.2-3
Specification SR 5.2.3 - Tendon Load Cell 5.2-4
Specification SR 5.2.4 - PCRV Concrete Crack 5.2-4
Specification SR 5.2.5 - Liner Specimen 5.2-6
Specification SR 5.2.6 - Plateout Probe 5.2-7
Specification SR 5.2.7 - Water Turbine Drive 5.2-8
Specification SR 5.2.8 - Bearing Water Makeup Pump 5.2-9
Specification SR 5.2.9 - He Circulator Bearing Water
Accumulators 5.2-10
Specification SR 5.2.10 - Engine-driven Fire Pump 5.2-10
Specification SR 5.2.11 - Primary Reactor Coolant Radio-
activity 5.2-11
Specification SR 5.2.12 - Primary Reactor Coolant Chemical 5.2-11
Specification SR 5.2.13 - PCRV Concrete Helium
Permeability 5.2-12
Specification SR 5.2.14 - PCRV Liner Corrosion 5.2-12
Specification SR 5.2.15 - PCRV Penetration Interspace
Pressure 5.2-13
Specification SR 5.2.16 - PCRV Closure Leakage 5.2-13
5.3 SECONDARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS 5.3-1
Specification SR 5. 3.1 - Steam/Water Dump System Valves 5.3-1
Specification SR 5.3.2 - Main and Hot Reheat Steam Stop
Check Valves 5.3-2
Specification SR 5.3.3 - Bypass and Safety Valves 5.3-2
Specification SR 5.3.4 - Safe Shutdown Cooling Valves 5.3-3
Specification SR 5.3.5 - Hydraulic Power System 5.3-4
Specification SR 5.3.6 - Instrument Air System 5.3-4
Specification SR 5.3.7 - Secondary Coolant Activity 5.3-5
5.4 INSTRUMENTATION AND CONTROL SYSTEMS - SURVEILLANCE
REQUIREMENTS 5.4-1
Specification SR 5.4.1 - Protective Instrumentation
Checks, Calibrations, and Tests 5.4-1
Specification SR 5.4.2 - Control Room Smoke Detector 5.4-12
Specification SR 5.4.3 - Core Region Outlet Temperature
Instrumentation 5.4-12
Specification SR 5.4.4 - PCRV Cooling Water System
Temperature Scanner 5.4-13
Specification SR 5.4.5 - PCRV Cooling Water System Flow
Scanner 5.4-13
Specification SR 5.5.6 - Core AP Indicator 5.4-14
Specification SR 5.4.7 - Control Room Temperature 5.4-14
Specification SR 5.5.8 - Power To Flow Instrumentation 5.4-14
Specification SR 5.4.9 - Area & Miscellaneous Process
Radiation Monitors 5.4-15
Specification SR 5.4.10 - Seismic Instrumentation 5.5- 5
Specification SR 5.5.11 - PCRV Surface Temperature Indication 5.4-16
iv
PAGE
5. 5 CONFINEMENT SYSTEM - SURVEILLANCE REQUIREMENTS 5.5-1
Specification SR 5.5.1 - Reactor Building 5.5-1
Specification SR 5.5.2 - Reactor Building Pressure
Relief Device 5.5-1
Specification SR 5.5.3 - Reactor Building Exhaust Filters 5.5-2
5.6 EMERGENCY POWER SYSTEMS - SURVEILLANCE REQUIREMENTS 5.6-1
Specification SR 5.6.1 - Standby Diesel Generator 5.6-1
Specification SR 5.6.2 - Station Battery 5.6-2
5.7 FUEt HANDLING AND STORAGE SYSTEMS - SURVEILLANCE
REQUIREMENTS 5.7-1
Specification SR 5.7.1 —Fuel Handling Machine 5.7-1
Specification SR 5.7.2 - Fuel Storage Facility 5.7-2
5.8 RADIOACTIVE EFFLUENT DISPOSAL SYSTEMS - SURVEILLANCE
REQUIREMENTS 5.8-1
Specification SR 5.8.1 - Radioactive Gaseous Effluent
System 5.8-1
Specification SR 5.8.2 - Radioactive Liquid Effluent
System 5.8-1
5.9 ENVIRONMENTAL SURVEILLANCE - SURVEILLANCE REQUIREMENTS 5.9-1
Specification SR 5.9.1 - Environmental Radiation 5.9-1
6.o DESIGN FEATURES 6.0-1
6. 1 REACTOR CORE - DESIGN FEATURES 6.1-1
Specification DF 6.1 - Reactor Core 6.1-1
v
PAGE
6.2 REACTOR COOLANT SYSTEM AND STEAM PLANT SYSTEM -
DESIGN FEATURES 6.2-1
Specification DF 6.2.1 - PCRV 6.2-1
Specification DF 6.2.2 - Steam Generator Orifices 6.2-3
Specification DF 6.2.3 - Steam Safety Valves 6.2-3
6.3 SITE - DESIGN FEATURES 6.3-1
Specification DF 6.3 - Site 6.3-1
1.0 ADMINISTRATIVE CONTROLS 7.0-1
7.1 ORGANIZATION, REVIEW AND AUDIT - ADMINISTRATIVE CONTROLS 7.1-1
Specification AC 7.1.1 - Organization 7.1-1
Specification AC 7.1.2 - Plant Operations Review
Committee 7.1-3
Specification AC 7.1.3 - Nuclear Facility Safety
Committee 7.1-5
7.2 SAFETY LIMITS - ADMINISTRATIVE CONTROLS 7.2-1
Specification AC 7.2 - Action to be taken if a Safety
Limit is Exceeded 7.2-1
7.3 ABNORMAL OCCURRENCE - ADMINISTRATIVE CONTROLS 7.3-1
Specification AC 7.3 - Action to be Taken in the Event of
an Abnormal Occurrence 7.3-1
7.4 RECORDS - ADMINISTRATIVE CONTROLS 7.4-1
Specification AC 7.4 - Records 7.4-1
7.5 PROCEDURES - ADMINISTRATIVE CONTROLS 7.5-1
Specification AC 7.5 - Procedures 7.5-1
7.6 REPORTING - ADMINISTRATIVE CONTROLS 7.6-1
Specification AC 7.6 - Reporting 7.6-1
vi
1-1
1.0 INTRODUCTION
These Technical Specifications apply to the Fort St. Vrain Nuclear
Generating Station Unit No. 1. These Technical Specifications pertain
to certain features, characteristics and conditions governing the
operation of this facility which are important in protecting the
barriers in the facility that separate the radioactive materials in
the facility from the environs.
These Technical Specifications will not be changed except by
express permission and with the approval of the Atomic Energy Commission.
2-1
2.0 DEFINITIONS
The following frequently used terms are defined to provide a uniform
basis for interpretation of these Technical Specifications.
2.1 Abnormal Occurrences
An abnormal occurrence is defined as any of the following:
a) A safety system setting less conservative than the limiting
setting established in the Technical Specifications.
b) Violation of a limiting condition for operation established
in the Technical Specifications.
c) Failure of a component of an engineered safety feature or safety
system that causes or threatens to cause the feature or system
to be incapable of performing its intended function. Simultaneous
failure of more than one component making up a redundant system
shall be considered a failure under this definition. In addition,
any failure of a component of an engineered safety feature or
safety system shall be considered a failure under this definition
unless it can be shown that the fault was not generic in nature.
d) Abnormal degradation of one of the several boundaries designed to
contain the radioactive materials resulting from the fission process.
e) Significant uncontrolled or unanticipated changes in reactivity.
equal to or greater than 1% AK/K.
f) Observed inadequacies in the implementation of administrative or
procedural controls, such that the inadequacy causes or threatens to
cause, the existence or development of an unsafe condition in
connection with the operation of the plant.
2-2
2.2 Equipment Surveillance Test
A test of the functional capability of a piece of equipment to
determine that it is operable. This may consist of either an on line or
off line demonstration of the operability of the equipment.
2.3 Instrumentation Surveillance
a) Channel Check
A qualitative determination that the channel is operable.
The determination is made by observation of channel behavior
during operation or comparison with other channels monitoring
the same variable or related variables.
b) Channel Test
A test of the functional capability of the channel to determine
that it is operable. This may consist of the injection of a
simulated signal into a channel as close as possible to the
primary sensor to verify that it is operable.
c) Channel Calibration
The adjustment of a channel so that it corresponds within
acceptable range and accuracy, to known values of the
parameter which the channel monitors. Calibration shall
encompass the channel and alarms up to the bistable output.
. 4 Irradiated Fuel
Irradiated fuel is fuel that has a radiation level > 100 mr/hr
measured one foot from the element surface.
2. 5 Low Power Operation
Low Power Operation is any operation with the Wide Range Logarithmetic
i.nstrtunentation indicating greater tha❑ 10-3% and less than 2% of
rated thermal power.
2-3
2.6 Normal Operating Range
The range of all plant parameters which can normally be expected
to occur during power operation, low power operation, and
reactor shutdown.
2./ Operable
Operable means that the system or component is capable of
performing its design function.
2.8 Operating
Operating means that the system or component is performing
its design function.
2.9 Plant Protective System
The plant protective system is the reactor protective circuitry
and the circuitry oriented towards protecting various plant
components from major damage. This system includes (1) scram,
(2) loop shutdown, (3) circulator trip, and (4) rod withdraw
prohibit.
2.10 Power Operation (or Reactor Operated at Power)
Power operation is any operation with the Linear Power Range
• instrumentation indicating more than 2% of rated therma]. power.
2.11 Radioactive Effluent
An effluent released from the plant containing radioactivity
measurably in excess of natural background.
2.12 Rated Thermal Power
Rated thermal power is 842 Mw(th).
2-4
2.13 Reactor Pressures
a) Normal Working Pressure (NWP) = 688 psig
b) Peak Working Pressure (PWP) = 704 psig
c) Reference Pressure (RP) = 845 psig
RP is the maximum PCRV internal pressure allowed over its
30-year operating life except for the initial pressure test.
2.14 Reactor Shutdown
The reactor is considered shut down when either (1) there is
no fuel in the reactor, or (2) when the reactor mode switch
is locked in the "OFF" position simultaneous with either of
the following reactivity conditions:
a) Hot Shutdown
A sufficient amount of control is inserted (at any average
core temperature* >220°F) that will yield a 0.01 Ap shutdown
margin with the average core temperature at 220°F in a
xenon free condition.
b) Cold Shutdown
A sufficient amount of control is inserted (at any average
core temperature* > 80°F) that will yield a 0.01 Ap shutdown
margin with the average core temperature at 80°F in a xenon
free condition.
*Average Helium Circulator Inlet Temperature plus Average Core
Outlet Temperature Divided by Two (2)
2.15 Refueling Cycle
Refueling cycle is defined as that interval of time between successvie
scheduled refuelings of a significant (> one-tenth) portion of the core.
2-5
2.16 Refueling Shutdown
The reactor is considered shut down for refueling purposes when
the reactor mode switch is locked in the "Fuel Loading" position
simultaneous with either hot shutdown or the cold shutdown
reactivity conditions.
2.11 Safe Shutdown Cooling
Safe shutdown cooling refers to cooling of the core with Safe
Shutdown Equipment providing for removal of core stored energy
and for adequate sustained decay heat removal. The reactivity
condition in the core is either hot or cold shutdown.
2.18 Surveillance Interval
A surveillance interval is the interval of time between
surveillance check, tests, or calibration. Unless otherwise
stated, the surveillance interval can be adjusted by ± 25% to
accomodate normal operational schedules. No surveillance interval
shall exceed 15 months unless otherwise specifically stated.
Unless otherwise stated in these specifications, surveillance may be
terminated on those instruments or equipment not in normal use
during reactor shutdown or refueling shutdown if the surveillance
interval is one month or less.
2.19 Trip
Trip is defined as the switching of an instrument or a device
with two stable states from its normal state to its abnormal
state. The result of a trip on a system level may be control
rod scram, pressure relief, loop shutdown, etc. .
3.0-1
3.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
Safety Limits are defined to protect the fuel particle integrity
and the integrity of the primary reactor coolant system boundaries.
The integrity of these barriers will ensure that an uncontrolled
release of radioactivity could not occur. Exceeding a Safety
Limit will not necessarily result in a violation of one of these
barriers, but may reduce the safety margin by an undesirable
degree. Violation of one of these barriers may not in itself
result in the uncontrolled release of radioactivity, but would
seriously diminish the protection from such an occurrence.
Limiting Safety System Settings are established for instrumentation
and protection devices related to the process variables upon
which Safety Limits are based.
An adequate margin is provided between the Limiting Safety
System Settings and the Safety Limits so that Safety Limits
would not be exceeded in the event that protective action is
initiated if a Limiting Safety System Setting is exceeded.
3.1-1
3.1 REACTOR CORE - .SAFETY LIMIT
Applicability
Applies to the limiting combinations of core thermal power
and core helium flow rate.
Objective
To maintain the integrity of the fuel particle coatings.
Specification SL 3.1 - Reactor Core Safety Limit
The combination of the reactor core power-to-flow ratio and
the total integrated operating time at this power-to-flow ratio
during any refueling cycle shall not exceed the limit given in
Figure 3.1-1. This safety limit is exceeded when the combination
of operating parameters (power, flow, and time) lies above or
to the right of the line given in Figure 3.1-1.
For the purpose of obtaining the total effective integrated
operating time for Figure 3.1-1, only transients resulting in a
power to flow ratio above the curve of Figure 3.1-2, at the
appropriate core power level shall be used.
Basis for Specification SL 3.1
In order to assure integrity of the fuel particles as a fission
product barrier, it is necessary to prevent the failure of significant
quantities of fuel particle coatings. Failure of fuel particle
coatings can result from the migration of the fuel kernels through
their coatings. The dependence of the rate of migration of the
particle kernel upon temperature and temperature difference across
the particle kernel using 95% confidence levels on the experimental.
data was used. During power operation, there is a temperature
gradient across each fuel rod, the higher temperature being at the
3.1-2
center of the fuel rod and the lower temperature at the outer edge
of ';he fuel. In an overtemperature condition, fuel kernels can
move through their coatings in this temperature gradient, in the
direction of the higher temperature.
The Core Safety Limit has been constructed to assure that a
fuel kernel migrating at the highest rate in the core will penetrate
a distance less than the combined thickness of the buffer coating
plus the inner isotropic coating on the particle.
The quantity of failed particle coatings in the core at all
times is determinable by measurement of gaseous fission product
activity in the primary loop.
In Figure 3.1-1, the quantity, P, is the fraction of design core
thermal power; i.e. , core thermal power (MW) divided by 842. The quantity,
F, is the fraction of design core coolant flows at the circulators; i.e. ,
the total coolant flow at the circulators in (lb/hr) divided by 3.5 x 106 lb/hr .
The limiting combinations of core thermal power and core coolant
flow rate are established using a series of short time conservative
assumptions. All hot channel factors discussed in Section 3.6 and all
power peaking factors discussed in Section 3.5.4 of the FSAR were applied
in determining this limiting curve. The range of region radial power
peaking factors (average power density in any refueling region, Pre
g)
divided by average power density in the core, Pcore) was assumed to
he less than or equal to 1.83 and greater than or equal to 0.4 . The
maximum intra-region power peaking factor (average power density in a
fuel column, Pcol) divided by the average power density in a fuel region,
Preg) used was 1.46 ± 0.2 for regions with control rods inserted and
1.34 ± 0.2 for all unrodded regions. A conservative estimate of the most
3.1-3
unfavorable axial power distribution was also used. That is , the
ratio of power density in the bottom layer of fuel elements of a
core region, Plower layer' to the average power density of the region,
Prev is less than or equal to 0.90 ± 0.09 for regions with control
rods fully inserted or withdrawn, and 1.23 ± 0.12 for regions with
control rods inserted more than two feet. The measured region coolant
outlet temperature for the six regions with their orifice valves most
fully closed and all regions with control rods inserted more than two
feet, was assumed to be not more than 50°F greater than the core
average outlet temperature. The measured region coolant outlet temperature
for the remaining core regions was assumed to be not more than 200°F
greater than the core average outlet temperature. During normal operation,
a condition with any measured region outlet temperature more than 50°F
above average should not persist for longer than a few hours. A measure-
ment uncertainty for the core region outlet temperature of ± 50°F
was assumed. A 5% uncertainty in flow measurement and a 5% uncertainty
in reactor thermal power measurement was assumed in establishing the limit.
For the total fuel lifetime in the core, based on calculations
incorporating plant parameters and uncertainties appropriate for longer
times, migration of the fuel particle kernel through its coating would
be less than 20 microns for the fuel with the most damaging temperature
history and with the core operated constantly at any of the power-to-flow
ratios and power combinations shown on the curve of Figure 3.1-2. Out
of a total inner coating thickness of 90 microns, only 70 microns have
been used for the determination of fuel particle failure in setting
the limit curve in Figure 3.1-1.
As can be seen from Figure 3.1-1, sufficient time (at least 9 minutes)
3.1-4
is available for the operator to take corrective action to prevent
the core safety limit from being exceeded for power-to-flow
ratios less than or equal to 2.0. In order to reach a power-to-flow
ratio of this magnitude, significant equipment malfunction, or failure,
and/or one or more significant deviations from operating procedures
would have to occur.
For analyzed abnormal situations which might occur during the
life of the plant which could potentially result in a power-to-flow
ratio greater than 2.0, the reheat steam temperature scram at 1075°F,
or the high power scram at 140% of rated power, will prevent the
Core Safety Limit from being exceeded.
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PERCENTAGE OF DESIGN CORE
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FIGURE 3.1-2
3.2-1
3.2 REACTOR VESSEL PRESSURE - SAFETY LIMIT
Applicability
Applies to the internal pressure limits on the Prestressed
Concrete Reactor Vessel (PCRV) including the primary penetrations
to this vessel.
Objective
To ensure the integrity of the reactor vessel and primary
penetration closures.
Specification SL 3.2 - Reactor Vessel Pressure, Safety Limit
The PCRV internal pressure and penetration primary secondary
closure interspace pressure shall not exceed the Reference Pressure of
8145 psig when fuel is installed in the reactor.
Basis for Specification SL 3.2
In order to assure the integrity of the PCRV as a fission
product barrier, the steady-state and transient operating conditions
for the PCRV are described in Section 5.2 of the FSAR, and are
summarized as follows :
Normal Working Pressure (NWP) 688 psig
Peak Working Pressure (PWP) 701} psig
Reference Pressure (RP = 1.2 times PWP) 8115 psig
The Reference Pressure is the maximum PCRV internal pressure
allowed over its 30-year operating life, except for the initial
proof test pressure (IPTP).
From reactor startup, after compl,ticn of IPTP, pressurization
of the PCRV above Reference Pressure is positively prevented by
3.2-2
means of the safety valve installation, described in Section 6.8
of the FSAR and specified in Section 3.3 of these Technical Specifications.
The PCRV is designed such that , over its normal operational
life, its structural response to internal pressures up to Reference
Pressure is essentially elastic. When pressurized from atmospheric
to Reference Pressure, the PCRV concrete goes through an unloading
process from a full compressive state due to the effective prestressing
forces to a less compressive state. At the end of the established
design life of the vessel, the minimum stress condition will exist.
However, the concrete cross section will always remain in a net compressive
state when these design prestressing losses have occurred. Prior to
operation, the Fort St. Vrain PCRV is subjected to the IPTP (approx-
imately equal to 1.15 times Reference Pressure) to verify the structural
response of the vessel to an internal pressure greater than Reference
Pressure, and to demonstrate at an early age that the PCRV, when
pressurized to the Reference Pressure level, will remain in a net
compressive condition at the end of design life.
3.3-1
3.3 LIMITING SAFETY SYSTEM SETTINGS
Applicability
Applies to the trip settings for instruments and devices which
provide for monitoring of reactor power, hot reheat temperature,
reactor internal pressure, and moisture content of the helium
coolant.
Obj ective
To provide for automatic protective action such that the
principal process variables do not exceed a safety limit as a
result of transients.
Specification LSSS 3.3 — Limiting Safety System Settings
The Limiting Safety System Settings for trip shall be as
specified in Table 3.3.1.
3.3-2
Specification LSSS 3.3 - Limiting Safety System Settings
TABLE 3.3.1
Parameter Function Trip Setting
1. Reactor Core Limiting
Safety System Settings
a) High Neutron Flux Scram < 140% of rated
thermal power
b) High Reheat Steam Scram < 1075°F
Temperature
c) Low Primary Coolant Scram < 50 psi below rated,
Pressure programmed with load
2. Reactor Vessel Pressure
Limiting Safety System
Settings
a) High Primary Coolant Scram and Preselected < 53 psi above rated,
Pressure Loop Shutdown and programmed with load.
Steam/Water Dump Upper programmed limit
set to produce trip at
4 775 psia
b) High Moisture in the Scram, Loop Shutdown < 67°F dewpoint temp-
Primary Coolant and Steam/Water Dump erature (corresponds to
< 500 ppmv H2O @
700 psia pressure)
c ) PCRV Pressure Pressure Relief
Rupture Disc (Low 1 @ 812 psig ± 1%
Set Safety Valve)
Low Set Safety Valve 1 @ 796 psig ± 1%
Rupture Disc (High 1 @ 832 psig ± 1%
Set Safety Valve)
High Set Safety Valve 1 @ 812 psig ± 1%
d) Helium Circulator Pressure Relief
Penetration
Interspace
Pressure
Rupture Disc (2 per
825 psig ± 2%
penetration)
3.3-3
TABLE 3. 3. 1 (continued)
Parameter Function Trip Settii':;
Safety Valve (2 per 805 psig ± 3%
penetration)
e) Steam Generator Pressure Relief
Penetration
Interspace Pressure
Rupture Disc (2 for 825 psig ± 2%
each steam generator)
Safety Valve (2 for 475 psig ± 3%
each steam generator)
3.3-4
Basis for Specification LSSS 3. 3
Safety Limits have been established in Specifications SL 3.1 and
SL 3.2 to safeguard the fuel particle integrity and the reactor coolant
system barriers. Protective devices have been provided in the plant
design to ensure that automatic corrective action is taken when
required to prevent the Safety Limits from being exceeded during normal
operation, or during operational transients resulting from possible
operator errors, or as a result of equipment malfunction. This specification
establishes the trip settings for these automatic protective devices .
High Neutron Flux
The neutron flux trip setting has been established to protect the
fuel particle integrity during rapid overpower transients. The power
range nuclear channels respond to changes in neutron flux. However,
near rated thermal power, the channels are calibrated using a plant heat
balance so that the neutron flux that is sensed is read out as percent
of rated thermal power. For slow maneuvers, those where core thermal
power, surface heat flux, and the power transferred to the helium
follow the neutron flux, the power range nuclear channels will indicate
reactor thermal power. For fast transients, the neutron flux change
will lead the change in power transferred from the core to the helium
due to the effect of the fuel, moderator and reflector thermal time
constants. Therefore, when the neutron flux increases to the scram
trip setting rapidly, the percent increase in heat flux and power
transferred to the helium will be less than the percent increase in
neutron flux. A fixed trip setting is sufficient for the plant because
the negative temperature coefficient of reactivity and large heat
capacity of the reactor limit the transient increases in fuel and helium
3.3-5
temperature to acceptable values. Section 14.2 of the FSAR describes
postulated reactivity accidents and transient response. Based on a
complete evaluation of the reactor dynamic performance during normal
operation and expected maneuvers and during and following various
assumed mechanical failures, it was concluded that sufficient pro-
tection is provided by the single fixed point scram setting.
High Reheat Steam Temperature
High reheat steam temperature indicates either an increase in
thermal power generation without an appropriate increase in helium
cooling flow rate or a decrease in steam flow rate. Reheat steam
temperature in lieu of reactor outlet helium temperature is used because
of the difficulty in measuring gross helium temperature for protective
system purposes. The design of the steam generator is such that changes
in hot helium temperature due to a power increase first affect the
reheat steam temperature thus allowing the latter to serve as an index
of the helium temperature. A reheat steam temperature scram is
provided to prevent excessive ratio of power-to-helium flow due to a
power increase or steam flow imbalance.
(Section 14.2 of the FSAR. )
Low Primary Coolant Pressure
The low primary coolant pressure trip setting has been established
to protect the fuel particle coating integrity due to the loss of
primary coolant as the result of a coolant leak.
High Primary Coolant Pressure
The major potential source of primary coolant pressure increase
above normal operating range is due to water anclor steam inleakage by
means oL' a defective evaporator-economizer-superheater subheader or tube.
3.3-6
For a double ended offset tube rupture the rate of water and steam
inleakage will not exceed 35 lbs/sec initially, resulting in a
maximum rate of primary coolant pressure rise of approximately
1 psi per second. The normal plant protection system action upon
detection of moisture is reactor scram, loop shutdown, and steam/water
dump (FSAR Section 7.1.2.5) , occurring after approximately 12 seconds.
In this situation, the peak PCRV pressure at 100% reactor power is
limited to 705 psia. (FSAR Table 14.5-1) . Backup protective action
is provided by the high primary coolant pressure scram, loop shutdown,
and dump of a pre-selected loop and remaining loop steam depressurization.
(FSAR Section 7.1.2.4). The trip setting of < 53 psi above reactor pressure
between 25% and 100% of rated power is selected: (1) to prevent false
scrams due to normal plant transients, and (2) to allow adequate time
for the normal protective action (high moisture) to terminate the
accident while limiting the resulting peak PCRV pressure in the unlikely
event that the normal protective action were inoperative. In this case,
reactor pressure would continue to rise (for a time interval of
approximately 75 seconds in the case of the maximum leak rate) to the
high pressure trip setting. The resulting peak PCRV pressure would be
less than the PCRV reference design pressure (FSAR Table 14.5-1) . The
high pressure trip setting is programmed as a function of load, using helium
circulator inlet temperature as the measured variable indicative of toad.
The upper trip setting limit is set at 775 psia; this upper trip setting
limit would be reached at a programmed circulator inlet temperature
of 770°F, (FSAR Fig. 7.1-14) . The PCR/ safety valves provide the
ultimate protection against primary coolant system pressure exceeding
the PCRV reference design pressure of 845 psig.
3.3-7
High Moisture in the Primary Coolant
The high moisture trip setting corresponding to < 500 pp<av
moisture was established, considering the moisture monitor
characteristics and the necessity to minimize water inleakage
to the reactor system. A trip at 500 ppmv would be reached after
several hours of full power operation with a minimum water/steam
inleakage rate in excess of about 20 lbs/hr. Below that inleakage
rate, the trip setting would never be reached, but the indicating
instruments would show an abnormal condition. For maximum design
leakage rates, the system behavior is as discussed in the preceding
section on High Primary Coolant Pressure.
PCRV Pressure
If the pressure in the PCRV were to rise significantly above
the normal operation pressure, the low-set rupture disc would
rupture within the range of 804 psig (-1%) , to 820 psig (+1%).
The low set safety valve, set at 796 psig ± 1%, would be wide open
and flowing full capacity at or above 820 psig (3% accumulation).
If the pressure still continued to rise, the high-set rupture disc
would rupture between 824 _isig and 840 psig. The high-set safety
valve, set at 812 psig ± 1%, would be flowing full capacity above
836 psig (3% accumulation) . As the pressure decreased, the high-set
safety valve would close at a line pressure of approximately 69C
psig and the low-set safety valve at approximately 677 psig; the
corresponding primary system pressure would be approximately 737 psig
when the low-set safety valve closed. (FSAR Section 6.8.3)
3.3-8
Helium Circulator Penetration Interspace Protection
The penetration interspaces are protected against pressures
exceeding PCRV reference pressure. The safety valves are set at
805 psig and rupture discs are set at 825 psig (nominal) . The
rupture discs would burst in the pressure range of 809 psig (-2%)
to 842 psig (+2%) . The safety valves would open in the range of
781 psig (-3%) to 829 psig (+3%) and would relieve full capacity
at 886 psig (10% accumulation). The safety valves would reseat
at about 725 psig. The safety valve and rupture disc relieving pressures
were specified so as to comply with the ASME Boiler and Pressure
Vessel Code, Section III, Class 8, Nuclear Vessels, for over pressure
protection.
Steam Generator Penetration Interspace Protection
The six steam generator penetration interspaces in each loop
are provided with a common upstream rupture disc and safety valves
to protect against pressures exceeding PCRV reference design
pressure (845' psig) . A redundant safety valve and rupture disc
are provided. The rupture discs would burst in the pressure
range of 809 psig (+2%) to 842 psig (+2%) , with a nominal setting
of 825 psig. The safety valves are each set at 475 psig which allows
for a pressure drop in the inlet lines of 370 psi when relieving
at valve capacity.
4.0-1
4.0 LIMITING CONDITIONS FOR OPERATION
The Limiting Conditions for Operation, specified in this section,
define the lowest functional capability or performance levels necessary
to assure safe operation of the facility. These Limiting Conditions
for Operation provide for operation with sufficient redundancy so
that further, but limited, degradation of equipment capability or
performance, or the occurrence of a postulated incident will not prevent
a safe reactor shutdown.
These Limiting Conditions for Operation do not replace plant
operating procedures. Plant operating procedures establish plant
operating conditions with at least the capability and performance
specified in these Limiting Conditions for Operation. Violation
of a Limiting Condition for Operation shall be corrected as soon as
practicable. Unless otherwise stated in these specifications,
the condition would be corrected or the reactor shall be shutdown
in an orderly manner within a 24-hour period.
4.1-1
4.1 REACTOR CORE AND REACTIVITY CONTROL - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to the characteristics of the reactor core.
Objective,
To define minimum operable equipment and the characteristics
of the reactor core and reactivity control systems to ensure the
capability to control the reactivity of the core and to maintain
the fuel particle coatings as the primary fission product barrier.
Specification LCO 4.1.1 - Core Irradiation, Limiting Conditions
for Operation
The maximum irradiation of the fuel or reflector elements shall
not exceed either of the following conditions:
a) The incore irradiation lifetime of the fuel elements
and reflector elements immediately adjacent to the
active core shall be limited to the equivalent of 1800
effective days at rated thermal power.
b) The average burnup within a fuel region shall be limited
to 110,000 Mwd per tonne of initial uranium plus thorium.
Basis for Specification LCO 4.1.1
The basis of integrity of the coatings of the fuel particles and graphite
dimensional changes is dependent on many variables. Prime variables are the
total burnup accumulated by the coated fuel particle and the fast fluence.
Limiting the allowable irradiation times and burnup to those specified will
ensure that the coated fuel particles and graphite will remain within the
demonstrated irradiation test values. (See Section 3.4 and Appendix A.l.l
of the FSAR).
4.1-2
Specification LCO 4.1.2 - Operable Control Rods, Limiting Conditions
for Operation
The reactor shall not be operated at power unless a sufficient
number of control rod pairs are operable or fully inserted to assure
that cold shutdown can be achieved with a failure to insert the
highest worth operable withdrawn rod pair.
Basis for Specification LCO 4.1.2
The initial startup test program and the repair and/or replacement
of control rods require an assured subcriticality at room temperature
with one rod pair removed. This condition is always met as is shown
in Section 3.5.3.1 of the FSAR.
The allowable number of inoperable, withdrawn control rod pairs
will depend on the available total shutdown margin at various points
in life, as will be determined from measurements of control rod
worth, xenon worth, and knowledge of the worth of Pa-233 as a function
of operating and shutdown time.
The decay of Pa-233 occurs over a period of several weeks and is
predictable. Any time during power operation, after the Pa concentration
has equilibrated, cold shutdown capability is retained before the
decay of Pa-233 with any two rod pairs removed, or any two adjacent rod
pairs plus a third rod pair removed that is at least 3 regions from the
two unrodded regions. Sufficient time is available to repair or
replace at least one removed rod pair, or to activate the reserve
shutdown system before the Pa-233 decay reduces the cold shutdown
margin to an unsafe value. (FSAR Section 3.5.3.1)
4.1-3
Specification LCO 4.1.3 - Rod Sequence, Limiting Conditions
for Operation
Control rods shall be withdrawn or inserted in groups (3 rod pairs)
in a specified sequence. This sequence shall be followed except for rod
insertion resulting from a scram, rod runback, or during low power physics
testing, or as allowed in Specification LCO 4.1.4. This sequence shall be
approved by the Nuclear Facility Safety Committee.
Basis for Specification LCO 4.1.3
The following criteria shall be used as the basis to establish any
control rod withdrawal sequence:
a) The maximum calculated reactivity worth of any rod pair in any rod
configuration with the reactor critical shall not exceed 0.047 Ak.
b) The maximum allowable calculated single control rod pair worth,
at any core condition, during power operation shall depend on the
available core temperature coefficient. The accidental removal of
this maximum worth single rod-pair shall result in a transient
with consequences no more severe than the withdrawal of .012 Ak,
at rated power, from a core which has a temperature defect between
220°F and 1500°F of .028 Ak. (FSAR, Section 3.5.5.1 and 14.2.2.6)
c) Calculated power peaking factors in any normal operating configuration
shall be within the following specified range:
I. If the average core outlet temperature is > 950°F,
the calculated average peaking factor of any refueling
region shall be larger than 0.4 and less than 1.83:
0.4 < Preg < 1.83
Poore
4.1-4
II. If the average core outlet temperature is <950°F,
the calculated average peaking factor of any refueling
region shall be:
0.4 < Preg < 3.0
Pcore
III. The calculated maximum intra-region power peaking factor
shall be as follows: An uncertainty of ± 15% is applied
to this factor.
Tgas out > 950°F Regions with Control Pcol < 1.46 ± 0,2
Rods Fully Inserted _—
Preg
Regions with Control _
Rods Partially Pcol < 1.40 ± 0,2
Inserted, more than
two feet into the Preg
core
Regions with Control —
Rods not Inserted More Pao' < 1.34 ± 0.2
Than two feet into the --
core Preg
T < 950°F All Regions col < 1.61 ± 0,2
gas out
Preg
4.1-5
IV. The calculated axial peaking factor, in any region
shall be as follows: An uncertainty of 10% is to be
applied to this factor.
Regions With Control Plower layer ` 0.90 ± 0.09
Rods Fully Inserted
Preg
Regions With Control _
Rods Partially Plower layer ` 1.23 ± 0.12
Inserted, more than _
Two Feet into the Preg
Core
Regions With Control _
Rods Not Inserted Plower layer ` 0.90 ± 0.09
More than Two Feet _
into the Core Preg
The specification of a rod-pair withdrawal sequence to achieve low
power operation is required to:
a) Monitor the reactivity worth of rods withdrawn during the
approach to critical by measurements of changes in the
measured multiplied source neutrons.
b) Maintain an acceptable flux distribution at lower power
by keeping the flux level in the center of the core at
least as high as the average level.
c) Insure that the calculated maximum worth rod in low power
operation, if assumed accidentally withdrawn, would result
in a transient with consequences no more severe than the
rod withdrawal from low power accident analyzed in the
FSAR (Section 3.5.3.1 and 14.2.2.7) .
4.1-6
The specification of a rod-pair configuration during power operation
is required to yield an acceptable power distribution. In addition, the
sequence guarantees that the combination of maximum single rod-pair
worth and available core temperature coefficients, in the event of
an accidental rod withdrawal, will result in a transient with consequences
less severe than that analyzed in the FSAR. This configuration will
change from year to year as different regions of the core are refueled.
An example of typical operating configurations for the initial core
is shown in Table 3.5-8 of the FSAR (Section 3.5.3.4) .
The rod withdrawal accident analysis at rated power as described
in the FSAR was based on a maximum rod pair worth of 0.012 Ak, using
temperature coefficients equivalent to a reactivity defect from
refueling (220°F) to operating temperature (1500°F) of 0.028 Ak. For
power operation in the range from 2 to 100 percent power, the fuel
temperature may be lower than the full power operating fuel temperature
of 1500°F. This results in a greater number of control rod pairs inserted
for the critical configuration, and a larger maximum single rod-pair worth.
A value larger than .012 Ak for a single rod pair can be safely
accomodated if fuel temperatures are lower than 1500°F and/or the
temperature defect between refueling temperature (220°F) and operating
temperature (1500°F) is greater than .028 Ak. (FSAR Section 14.2.1.1)
Whenever the core temperature coefficient is more negative, whether
it be caused by lower core temperatures or related core characteristics
such as the C/Th atom ratio, a larger value of the maximum single
rod-pair worth is acceptable with no greater consequences daring the
rod withdrawal accident. (See FSAR Section 14.2.1.1)
4.1-7
Analyses have been performed (See FSAR Section 14.2.1.1) to
determine the allowable increased single rod-pair worths at fuel temp-
eratures between 220°F and operating temperature (1500°F) for various
core temperature coefficients which show that the consequences of
an accidental rod pair withdrawal are no more severe than the 0.012 Ak
rod withdrawal at rated power analyzed in the FSAR. (Section 14.2.2.6)
The specified range of power peaking factors was used in developing
the Core Safety Limit of Specification SL-3.1 since the limiting
combinations of core thermal power and core coolant flow rate are a
function of the region radial power peaking factors, the intraregion
power peaking factors and the axial power distribution. Specifying a
control rod sequence which has peaking factors within these power
peaking factor limits assures that the criteria upon which Specification
SL-3.1 is based are met.
Specification LCO 4.1.4 - Partially Inserted Rods, Limiting Conditions
for Operation
All rod-pairs must be either fully inserted or fully withdrawn
except that:
a) Within the rod sequence of Specification LCO 4.1.3, two groups
of three rod-pairs (six rod-pairs) in addition to the regulating
rod-pair may be partially inserted, provided the two groups
axial positions are separated by at least 10 feet.
b) Without regard to the rod sequence of Specification LCO 4.1.3,
a maximum of six rod-pairs may be partially inserted up to
2 feet in addition to a) above.
c) In addition to a) and b) above, two runback groups (six rod-pairs)
may be inserted tc any location for a time period, not to exceed 4 hours.
4.1-8
Basis for Specification LCO 4.1.4
The presence of too many partially inserted rod pairs in the core
will tend to push the flux into the bottom half of the core and raise
the fuel temperatures. The intra-region power peaking factors and
axial power peaking factors used in determining the rod withdrawal
sequence in LCO 4.1.3, will be maintained during normal operation if
the rods are inserted and withdrawn in sequence and if partially
inserted rods are limited as in a) and b) above. (See FSAR 3. 5.4) .
In addition, up to 6 additional rod-pairs may be inserted up tc
two feet into the core. This will permit the operator to move rods to
change the radial power distribution and minimize both fuel temperatures .
The insertion of six additional rod pairs up to two feet in the
core has a minimal effect on the axial power distribution, resulting
in an increase in the average power density in the lower layer of
fuel of less than 5 percent.
The runback inserts two pre-selected groups of three rod-pairs during
rapid load reductions (see FSAR Section 7.2.1.2) . The partial insertion
of these control rod-pairs, up to six feet into the core (FSAR Section 3.5 .4.3)
in addition to a) and b) above would increase the average axial power
peaking factor in the lower layer of fuel to about 0.85. Negligible fuel
particle migration (See SL 3.1) would occur with this condition in the
core for four hours.
Specification LCO 4.1.5 - Reactivity Change with Temperature, Limiting,
Conditions for Operation
The reactivity change due to an average core temperature increase
between 220°F and 1500°F, refueling temperature to rated power conditions,
in the absence of xenon must be at least as negative as .031 dk.
4.1-9
Basis for Specification LCO 4.1.5
The negative temperature coefficient is an inherent safety mechanism
that tends to limit temperature increases during power excursions. It
is a stabilizing element in flux tilts or oscillations due, for example,
to xenon transients.
System temperatures during a power excursion beginning from a high
power level are well within design limits regardless of the magnitude of
the negative temperature coefficient, provided protective action is
initiated by a power level signal. However, if protective action
occurs much later, such as from a manual scram or an activation of the
reserve shutdown system, peak system temperatures will be sensitive to
the magnitude of the temperature coefficient. Peak fuel temperatures
during a power excursion beginning from low (or source) power levels
also depend on the temperature coefficient, particularly the fuel or
Doppler coefficient.
The reactivity change due to an average core temperature increase
between refueling (220°F) and operating conditions (1500°F), in the
xenon free core, is calculated to be at least as negative as .028 Ak.
This reactivity change implies an isothermal coefficient, with
equilibrium xenon, of -4.2 x 10-5 Ak/°F at 220°F and of -1.1 x 10-5 Ak/°F
at 1500°F, the data used for the safety analysis presented in the FSAR,
(Sections 3.5.5.1 and 14.2.2) .
The uncertainty in the measured temperature defect is estimated
to be about ± 10%, or about .003 Ak. By requiring that the measured
temperature defect be at least .031 Ak, a temperature coefficient at
least as negative as that used in the safety analysis is assured,
even if the maximum uncertainty in the measurement is applied.
4.1-10
Specification LCO 4.1.6 - Reserve Shutdown System, Limiting Condition
for Operation
Six reserve shutdown units of the 7 hopper subsystem and 29 reserve
shutdown units of the 30 hopper subsystem shall be operable whenever the
reactor is in low power or power operation and the core helium inlet
temperature is above 250°F to assure that hot shutdown can be achieved
from an operating condition. Any inoperable units must be capable of
being made operable within 7 days following a reactor shutdown utilizing
the reserve shutdown system.
Basis for Specification LCO 4.1.6
The reserve shutdown system must be able to achieve hot shutdown
in the event of a situation that prevents the insertion of any normal
control absorber.
After extended power operation the reserve shutdown system has
to cover the temperature defect between operating and refueling
temperature (220°F) , the decay of Xe-135, the buildup of Sm-149, and
the decay of Pa-233 to U-233. The core reactivity increase due to
cooldown and Xe decay occur fairly rapidly and is worth .089 Ak at the
beginning of the initial cycle. At the end of the initial cycle and
at the middle of the equilibrium cycle, by the time that Pa-233 has
reached an equilibrium concentration, the cooldown and Xe decay is worth
.081 Ak in the initial core and .076 Ak in the equilibrium core. During
the first 7 days following a shutdown, the core reactivity will rise
about .002 Ak due to Pa-233 decay and Sm-149 buildup. If the Pa-233
concentration had reached equilibrium, the total worth of its decay and
Sm buildup would be about .030 Lk in the initial core and .024 Ak in the
equilibrium core.
4.1-11
The reactivity requirements for the reserve shutdown system can
be summarized as follows:
Total Total
Required Worth Full Pa Worth
Cooldown and Pa Decay (7d) Shutdown 35 Units Decay Sm 37 Unit
Xe Decay (Ak) Sm Buildup (Ak) Operable Buildup Operable
Beginning
of Initial Cycle .089 0.0 0.01 .099 0.0 .099
Middle of
Initial
Cycle .081 .002 0.01 .093 .030 .121
Equilibrium
Cycle .076 .002 0.01 .088 .024 .110
As stated in Section 3.5.3.3 of the FSAR, the nominal worth for all
37 channels of the reserve shutdown system is .12 Ak at all times in the
absence of control rods. It was calculated to be as large as .14 Ak in
the initial core and .13 Ak in the equilibrium core. This is sufficient
reactivity control to cover core cooldown, Xe decay, and full Pa decay.
With the maximum worth channel inoperative in each subsystem, the worth
of the 35 inserted units was calculated to be greater than .101 Ak in
the initial core and .088 Ak in the equilibrium core. This is sufficient
reactivity control to cover core cooldown, Xe decay, and the first 7 days
of Pa decay.
Specification LCO 4.1.7 - Core Inlet Orifice Valves, Limiting Conditions
for Operation
The core inlet orifice valves shall be adjusted for the following
conditions :
4.1-12
a) Core Average outlet temperature > 950°F
The measured individual region outlet temperature
for the six regions whose valves are most fully
closed, and any region with control rods inserted
more than two feet into the core, shall not exceed
the core average outlet temperature + 50°F. The
measured individual region outlet temperature for
the remaining regions shall not exceed the core
average outlet temperature + 200°F.
b) Core average outlet temperature < 950°F
The measured individual region outlet temperature
for all 37 regions shall not exceed the core
average outlet temperature + 400°F and the conditions
of L.C.O. 4.1.9 must be met.
Corrective action shall be initiated at the onset of the condition
exceeding the limits stated. If the above limits are exceeded by
1) 100°F or more, an immediate orderly shutdown shall be initiated;
2) more than 50°F, but less than 100°F, corrective action must be
successful within 2 hours or an orderly shutdown shall be initiated;
3) less than 50°F, corrective action must be successful or the reactor
shutdown within 24 hours.
Basis for Specification LCO 4.1.7
The maximum helium flow imbalances used in developing the core
Safety Limit of Specification SL 3.1 corresponds to measured core
region outlet temperature which, for the 3ix regions with their orifice
valves most fully closed, and all regions with control rods inserted
4.1-13
more than two feet into the core, is no greater than 50°F above the
average core outlet temperature and which, for the remaining regions,
is no greater than 200°F above the core average outlet temperature.
A measurement uncertainty of ± 50°F was assumed for the core region
outlet temperatures in the development of Specification SL 3.1.
Specifying these maximum deviations from the average core outlet
temperature will assure that the criteria upon which Specification
SL 3.1 is based is met.
During power operation with an average core outlet temperature
less than 950°F, sufficient overcooling of the core is provided with
a +400°F deviation between the maximum and average core outlet
temperature to assure that Specification SL 3.1 remains valid and that
the integrity of the fuel particles would be preserved.
The time at temperatures exceeding the limits given, represents
conditions significantly below the core safety limit.
Specification LCO 4.1.8 - Reactivity Status, Limiting Conditions
for Operation
If the difference between the observed and the periodically
renormalized expected reactivities of the core at steady state
conditions reaches 0.012 Ak, the reactor shall be shutdown and reactor
operations shall not be resumed until a satisfactory explanation has
been found for the reactivity anomaly and permission is received
from the NFSC.
Basis for Specification LCO 4.1.8
An unexpected and/or unexplained change in the observed core
reactivity could be indicative of the existence of potential safety problems
4.1-14
or of operational problems. An observed change of 0.012 Ak would
represent a significant deviation from expected core reactivity
conditions, but would not represent an addition of reactivity greater
than the maximum rod pair worth, 0.012 Ak, for which a rod withdrawal
accident was analyzed in the FSAR (Section 14.1) .
Specification LCO 4.1.9 - Core Region Temperature Rise, Limiting
Condition for Operation
The measured helium coolant temperature rise through any core region
shall not exceed the limits given in Figure 4.1.9 (at the appropriate power
level) whenever the reactor is pressurized to more than 50 psia. Below
50 psia reactor pressure, the measured helium coolant temperature rise
shall not exceed 350°F with the core inlet orifice valves set at any position,
and shall not exceed 600°F with the core inlet orifice valves set for equal flow.
If the measured helium coolant temperature rise exceeds these limits,
immediate corrective action shall be taken. If this corrective action is
not successful within fifteen (15) minutes, an immediate orderly shutdown
shall be initiated.
Basis for Specification LCO 4.1.9
A maximum core region helium coolant temperature rise as a function of
calculated reactor thermal power (including power from decay heat) , as
indicated by Figure 4.1.9, has been specified to prevent very low helium
coolant flow rates through any coolant channel. Very low helium coolant flow
rates may result in laminar flow conditions with resultant high friction
factors and low heat transfer film coefficients and potentials for possible
local helium flow stagnation, which could result in excessive fuel temperatures.
The maximum cora region helium temperature rise given in Figure k.1.9
has been developed based upon a number of conservative assumptions. It was
4.1-15
assumed that the primary system was pressurized to full inventory. At
less than full inventory, higher region delta T's are acceptable. The
core inlet helium temperature was assumed to be 100°F. At higher core
inlet temperatures, higher region delta T's are acceptable for the
condition with helium flow orifice valves adjusted to yield equal helium
flow to all fuel elements, it was assumed that all regions had a power
density equal to the core average (Preg/Poore equals 1.0) . Regions with
higher than average power densities could yield acceptable region delta T's
higher than the limits of Figure 4.1.9, but conservatively have been
restricted to those of an average core power density region. For the
condition with orifice valves at any position, the allowable region delta T
is based upon the lowest power density region (Preg/Pcore equals 0.4) . For
regions with higher power densities, higher region delta T's are acceptable.
Conservatively these have been restricted to those of an 0.4 power density
region.
For depressurized operations, limits are also specified to prevent very
low helium coolant flow rates through any coolant channel. These limits
have been established based upon a 50 psia reactor pressure, and all other
conservative assumptions stated above.
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4.2-1
4.2 PRIMARY COOLANT SYSTEM - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to the configuration and characteristics of the primary
(helium) reactor coolant system excluding the steam generators which
are included in Section 4.3.
Objective
To ensure the capability to cool the reactor core and to
preserve the integrity of the fission product barriers , by defining
the minimum operable equipment of the primary reactor coolant system
and its characteristics.
Specification LCO 4.2.1 - Number of Operable Circulators, Limiting
Conditions for Operation
There shall be at least one operable circulator in each loop
during power operation. If only one of the four circulators is
operable at any time the reactor shall be immediately placed in a low
power or shutdown condition.
Basis for Specification LCO 4.2.1
One circulator, operating with condensate or firewater motive
power, provides for sufficient primary coolant circulation to assure
safe shutdown cooling. One circulator, operating with feedwater motive
power would provide sufficient primary coolant circulation following a
postulated depressurization accident. One circulator in each loop is
specified to allow for a single failure in either the heat removal
equipment or circulator auxiliary equipment which provides services
to one loop. Safe shutdown cooling is discussed in the FSAR, Section 10.3.9.
4.2-2
Specification LCO 4.2.2 - Operable Circulator , Limiting Conditions
for Operation
A circulator shall not be considered operable unless the following
conditions or system requirements are met for that circulator:
a) Emergency Feedwater and Firewater are available to
drive the water turbine and the capability for turbine
water drainage exists. The Emergency Feedwater or Condensate
Header may be inoperable for up to 24 hours without the helium
circulators being considered inoperable.
b) The normal bearing water system is operable.
c ) The associated bearing water accumulator system is operable.
d) Both Bearing Water Makeup Pumps are operable to provide required
makeup. One of the bearing water makeup pumps may be inoperable
for 24 hours without the helium circulators being considered
inoperable.
Basis for Specification LCO 4.2.2
The requirements for an operable circulator specified above provide
for adequate circulator water turbine supply and circulator auxiliary
supplies to assure safe shutdown cooling. Operation of one circulator
on emergency feedwater would provide adequate helium circulation following
a postulated depressurization accident. Each independent bearing water
system provides a continuous supply of bearing water to the two circulators
in each primary cooling loop. In addition, a backup bearing water system
is provided which automatically introduces water to the circulators if
the normal supply fails. Two gas pressurized bearing water accumulators
(one each for the two circulators in each primary coolant loop) are provided.
4.2-3
These accumulators contain sufficient water to permit circulator
coast-down without circulator damage if both the normal and the
backup bearing water supplies should fail. The minor water makeup
requirements for the normal bearing water system is provided by the
bearing water makeup pumps.
Specification LCO 4.2.3 - Turbine Water Removal Pump, Limiting Conditions
for Operation
There shall be one operable turbine water removal pump during
power operation.
Basis for Specification LCO 4.2.3
One turbine water removal pump has sufficient capacity to remove
the water from two circulator water turbines. This is adequate for
a safe shutdown cooling.
Specification LCO 4.2.4 - Service Water Pumps, Limiting Conditions
for Operation
At least two service water pumps and the associated pump pit
shall be operable during power operation.
Basis for Specification LCO 4.2.4
The availability of the service water system ensures the capability
of supplying essential components with cooling water, as described in
FSAR Sections 1.4, 10.3, and 14.4.
Specification LCO 4.2.5 - Circulating Water Makeup System, Limiting
Conditions for Operation
At least two circulating water makeup pumps connectible to the
essential bus shall be operable during power operation.
Basis for Specification LCO 4.2.5
Circulating water system makeup to the service water and tire protection
system provides adequate makeup water to safely shut the reactor down from
4.2-4
any normal operating condition. For further explanation see FSAR Sections
1.4, 10.3 and 14.4.
Specification LCO 4.2.6 - Firewater Pumps, Limiting Conditions for
Operation
The engine-driven fire pump, motor driven fire pump, and associated
pump pits shall be operable and there shall be at least 325 gallons of
fuel in storage during power operation.
Basis for Specification LCO 4.2.6
Either of the fire pumps provide adequate capacity to operate a
circulator water turbine(s) and supply emergency cooling water for safe
shutdown cooling. With the 325 gallons of fuel in storage, the engine-
driven fire pump can operate at rated conditions for 24 hours which is
adequate time to have more fuel delivered to the site. For further explanation
see FSAR Sections 1.4, 10.3 and 14.4.
Specification LCO 4.2.7 - PCRV Pressurization Limiting Conditions
for Operation
The PCRV shall not be pressurized to more than 100 psis unless:
a) The PCRV safety valve installation is operable, and there
is less than 5 psig between the rupture disc and relief
valve, and both inlet block valves are locked open.
b) All primary and secondary penetration closures and hold
down plates are in place and operable, pet Specification
LCO 4.2.9.
c) The interspaces between the primary and secondary penetration
closures are maintained at a pressure greater than primary
system pressure with purified helium gas.
4.2-5
d) One set of rupture discs and safety valves protecting
each steam generator and circulator penetration is operable,
and there is less than 5 psig between the rupture discs
and the relief valves.
When the PCRV is pressurized to more than 100 psia, corrective
action shall be initiated at the onset of any condition exceeding
the limits. If corrective action is not successful within 12 hours,
the reactor, if operating, shall be put in a shutdown condition,
followed by PCRV depressurization to less than 100 psia.
Basis for Specification 4.2.7
The PCRV safety valve installation (consisting of two parallel
systems, each of which has a manual block valve and a rupture disc
mounted upstream of the safety valves and which discharge to atmosphere
via a single particulate filter) provides the ultimate protection
against overpressuring the PCRV. A single manually operated
block valve is provided, between the PCRV and each rupture disc ,
so that necessary maintenance and/or testing of the discs and safety
valves can be performed after shutdown and depressurization of the
plant. Redundant instrumentation, as well as mechanical locks
on the valve, ensure that the valves will always be open when the
PCRV is pressurized.
The secondary closures serve two purposes related to plant safety:
(1) to form an interspace between the inner and outer closures that
can be maintained above primary coolant pressure with clean helium
to positively prevent any small normal leakage of contaminated helium
4.2-6
from the primary coolant system through the primary closure, and
(2) to eliminate the possibility of a large primary coolant system
leak as a result of any failure of a primary closure. In this latter
function, the secondary closures are considered to be a form of secondary
containment since, as in most power reactor plants, the secondary
containment prevents escape of radioactivity in the event of a primary
coolant system rupture, corresponding to a failure of the PCRV primary
closures. Since no credible failure of the PCRV liner , reinforcement,
and concrete can result in significant leakage, the secondary closures
of the PCRV penetrations thus constitute secondary containment of the
primary coolant system.
As long as penetration interspace pressure is maintained above
primary coolant pressure, any leakage into the reactor building will
be purified helium and will thus have no radiological consequences.
Penetration pressurizing gas is obtained from the high pressure
helium supply tanks of the helium storage system.
The steam generator and helium circulator penetrations are provided
with rupture discs and safety valves to prevent overpressure should a
process line rupture within the penetration. (These are the only
penetrations which contain process fluids at pressures high enough to
require such protection. ) Separate overpressure protection trains are
provided for a) the six steam generator module penetrations of each
loop and b) each of the four helium circulator penetrations. Each
train consists of a pair of rupture discs, each of which is upstream
of a safety valve, with the two rupture disc-safety valve combinations
piped in parallel. A block valve is provided at the inlet tc each
rupture disc. The block valves serving each pair of rupture discs
4.2-7
and associated safety valves are mechanically interlocked so that only one
valve can be closed at any time. Design basis for the circulator penetration
interspace safety valves is the rupture of a bearing water supply line.
For the steam generators penetration interspaces, the design basis is the
rupture of a subheader (35 lb/sec of superheated steam at 1000°F) .
Specification LCO 4.2.8 - Primary Coolant Activity, Limiting Conditions
for Operation
The primary coolant gaseous and plateout activity levels shall be
limited to:
a) The product of primary coolant noble gas beta plus gamma activity,
times E, shall not exceed 2,40 curies-mev (where E is the weighted
lb.
average of the beta and gamma energies per disintegration in MeV) ,
when measured 15 minutes after sampling.
b) The primary coolant circulating halogen inventory shall not
exceed an 1311 thyroid dose equivalent of 24 curies.
c) The plateout halogen inventory shall not exceed an 1311
thyroid dose equivalent of 5000 curies/loop.
d) The plateout 90Sr inventory shall not exceed a "Sr bone
dose equivalent of 140 curies/loop.
e) Determination of E will be performed at least once a month, and in
any event will be performed each time the primary coolant radioactivity
concentration changes by 25% from the previous measurement at the
same reactor power level. Calculations required to determine r
will consist of the following:
1. Quantitative measurement in units of Ci of radionuclides making
lb
up at least 95% of the noble gas beta plus gamma decay energy
in the primary coolant measured 15 minutes after sampling.
4.2-8
2. A determination of the beta and gamma decay energy per disintegration
of each nuclide determined in (1) above, by applying known
decay energies and schemes.
3. A calculation of E by appropriate weighting of each nuclide's beta
and gamma energy with its concentration as determined in (1) above.
Basis for Specification LCO 4.2.8
The whole body dose is a direct function of the gross gamma activity
in the primary coolant. The whole body skin dose is a direct function
of the gas beta activity in the primary coolant.
Measuring the primary coolant beta plus gamma activity 15 minutes
after sampling would indicate that activity that would reach the EAB*
following the postulated accident, taking into account the decay of
short half life isotopes and short term atmospheric conditions. The
131I thyroid dose equivalent and the 90Sr bone dose equivalent are
defined as a ratio of the rem/millicuries effectivity values for the
respective isotopes times the activity of the subject nuclide in millicuries.
The limits on the primary coolant noble gas beta plus gamma
concentrations are based on the maximum credible accident (FSAR Section 14.8)
wherein the entire primary circulating inventory is carried out of the
PCRV and is released to the atmosphere through the plant vent system. The
primary coolant noble gas beta plus gamma concentration is calculated
based on a short-term atmospheric dilution factor of 2.7 x 10-3 sec/m3
resulting from downdraft of the exhaust plume at a wind speed of 0.3 m/sec
during atmospheric condition F, and based on a combined total external
beta plus whole body gamma dose of 8.6 rem at the exclusion area boundary (EAB) .
*Exclusion Area Boundary
4.2-9
The equivalent 1311 and 90Sr activity primary coolant and plateout
limits are based on the Design Basis Accident No. 2 (PCRV rapid
depressurization-FSAR Section 14.11) wherein the entire primary
circulating inventory, and conservatively 6% of the plateout halogens/loop
and 5% of the plateout 90Sr/loop, is carried out of the PCRV and out of
the reactor building through the louvers. These maximum equivalents
result in calculated site boundary doses which will be well below
10 CFR 100 limits. The maximum equivalent activity levels determined
by the Design Basis Accident No. 2 are summarized in the following table.
Activity Levels Determined by the Depressurization Accident
Dose Nuclide Equivalent Curies Equivalent Activity Resulting
Category Reference Plated Out Released to Environ. Dose (rem)
1311 5000/loop 320 curies 138)
Thyroid (Plateout 150
12)
(Circulating 1311 Not Plated Out 24 curies
90Sr 140/loop 7 curies 75
Bone
The shown activity levels are based on the resulting doses at the
EAB as shown in the table, assuming a dilution factor of 8.4 x 10-4
sec/m3 and effectivities of 1,480 rem per m Ci of 1311 inhaled, and
36,100 rem per m Ci of 90Sr inhaled. These effectivity values are based
on information in ICRP II, and the newer data, especially for 908r ,
given in the more recent ICRP VI , were ignored.
Should information become available which lean to a change in the
given dilution factors , or should the data given in ICRP VI become acceptable,
the allowable activity concentrations will be changed accordingly.
The noble gas inventories will be calculated from grab samples and
the readings of the gross gamma monitor. It has been demonstrated by
4.2-10
theoretical investigations and experiments that the steady state release
rate of noble gas fission products from failed fuel particles is
proportional to the square root of the fission product half-life.
Further information is given in Section 3.7.2 of the FSAR. The inventory
of any non-measured noble gas nuclide will be calculated by assuming that
the release rate is proportional to the square root of the fission product
half-life. Figure 3.7-1 of the FSAR will be used as a guide in making
such determinations.
The 90Sr inventory will be determined by an analysis of the plateout
probes. In the interim between probe removals , the 90Sr inventory shall
be tentatively estimated from
-At t -A(t-T)
A (t) = A (0)e + J A (T)e At
"Sr 90Sr 0 90Kr
where A (0) is the total 90Sr inventory in the loop, as determined by
"Sr
the most recent plateout probe analyses, t is the elapsed time since
this determination, A is the decay constant for 90Sr, and A (i ) is
90Kr
the time dependent 90Kr activity in the coolant stream based on the
reading of the gaseous activity monitor and grab samples. Note that ,
if the 90Kr activity is constant (or bounded, or can be averaged) , the
estimated 90Sr inventory would be given by
-At _ -At
A (t) = A (0)e + A (1 - e ) .
90Sr 90Sr 90Kr
This method of estimating the 90Sr inventory in the interim between
probe removals is based upon the consideration that the source of 90Sr
is anticipated to be predominantly from 90Kr. However, the inventory
will be periodically updated by the probe analyses to give the total measured
90Sr, regardless of origin, the probe to be removed as specified in SR 5.2.( .
4.2-11
Specification LCO 4.2.9 - PCRV Closure Leakage, Limiting Conditions
for Operation
The total helium leakage through all the Primary Closure Seals in
any group of penetrations shall not exceed an equivalent leak rate of
400 lbs/day at a differential pressure of 10 psi.
The total helium leakage through all the secondary closure seals shall
not exceed an equivalent leakrate of 400 lbs/day at a differential pressure
of 688 psi.
Basis for Specification LCO 4.2.9
Penetration closure interspace volumes are normally maintained at a
pressure greater than the Primary Coolant Pressure by supplying them with
clean helium from either the high pressure helium storage tanks or from
the helium purification system; therefore, any leakage through either the
primary or secondary closure seals will be clean helium.
The normal gas supply to all the penetration closure interspaces is
from the helium purification system and is continuously monitored for
flow so that an increase in closure leakage can be sensed and alarmed.
The penetration closure interspaces are supplied with pressurizing gas
in groups through the arrangement of the purified helium piping. The
grouping of the penetrations is as follows:
Group I : All penetrations in the top head of the PCRV
(37-control rod drive, 2-high temperature
filter-adsorber, and 1-top access) .
Group II: All instrument penetrations (20) plus the bottom
access penetration.
Group III: The six steam generator penetrations, Loop I.
Group IV: The six steam generator penetrations , Loop II.
Group V-VIII: Each helium circulator penetration.
4.2-12
To prevent the possible loss of all helium coolant by way of the
Helium Purification System, due to a complete failure of a secondary
closure, the piping supplying pressurizing gas to the failed closure is
automatically isolated if the pressurization gas flow exceeds 275#/hr.
The leakage rate limitations for the primary closures are based
on a differential pressure of 688 psi , which would be the differential
pressure across a primary closure in the event a secondary closure
should fail.
The calculated permissible leakage rate across the primary closure
would be well in excess of 1145 lbs/hr. at a differential pressure of 688 psi .
Converting the 1145 lbs/hr. leakage rate to normal operating conditions
of 10 psi differential pressure, indicates an operating limiting leakage
rate of 400 lbs/day, or 16.7 lbs/hr. This leakage flow can readily be
detected on the pressurizing gas flow indicator. It is assumed that
under these conditions, the entire inventory of primary coolant would leak
through the primary closure. (The associated activity release would be
similar to that release resulting from the maximum credible accident (MCA) ,
discussed in Section 14.8 of the FSAR) . Assuming the design primary
coolant activity and assuming a dilution factor of 2.7 x 10-3 sec/m3 , the
resultant dose is at least an order of magnitude less than the limits
of 10 CFR 100 at the exclusion area boundary.
Secondary seal leakage during normal operation is leakage of clean
helium. The secondary seal leakage is limited to 400 lb/day at the
normal operating differential pressure of 688 psi, to assure compliance
with LCO 4.2.7, part c, which specifies pressurization of the penetration
interspaces to a pressure greater than primary system pressure.
4.2-13
Specification LCO 4.2.10 - Loop Impurity Levels, High Temperatures,
Limiting Conditions for Operation
The reactor shall not be operated with an average core outlet
temperature > 1200°F, if chemical impurity concentrations in the
primary coolant exceed 10 ppm (by volume) for the sum of H20, CO,
and C02. However, these amounts may be exceeded by up to a factor
of 10 for a period of ten days, or by up to a factor of 100 for one
day from the time the limit is exceeded.
Basis for Specification LCO 4.2.10
For plant operation in the normal power range (25% to 100% of
rated thermal power) , maximum impurity levels have been established
to restrict carbon transport from the reactor core to cooler
portions of the primary coolant system to about 330 lb/yr.
Limiting the quantity of carbon transported from the reactor
core insures the integrity of the fuel element, insures the integrity
of the core support structure, and limits the effect on the steam
generator heat transfer properties. The carbon corrosion will be
fairly uniformly distributed throughout the outlet third of the
core, resulting in a rate of weight loss from this portion of the core
of about 0,3% per year. (See FSAR Section 9.4.2).
Specification LCO 4.2.11 - Loop Impurity Levels, Low Temperatures,
Limiting Conditions for Operation
With the reactor operating and an average core outlet temperature
below 1200°F, impurity levels shall not be allowed to exceed:
H2O - those limits as a function of average core
outlet temperature given in Figure 4.2.11-1.
C02 - 1000 ppm (by volume)
CO - 15,000 ppm (by volume)
4.2-14
In addition to those limits above, during reactor startups and
shutdowns, the total time when reactor average outlet temperatures are
between 725°F and 1200°F, and moisture exceeds 10 ppm (by volume) shall
not exceed a total of 90 days during any one refueling cycle.
Basis for Specification LCO 4.2.11
During plant startups, core average outlet temperatures will be below
1200°F until the final stages when steam temperatures are increased to rated
and the plant enters the normal power range. At these lower temperatures,
graphite corrosion by the various chemical impurities is minimal and there
is reduced concern for carbon transport. Therefore, maximum impurity levels
have been established to prevent corrosion of metals in the primary coolant
system.
The moisture level allowable as a function of average core outlet
temperature, Figure 4.2.11-1, was developed to minimize burnable poison
oxidation, particularly during plant startups following reactor refuelings
when moisture levels are expected to be the highest.
At high temperatures in the presence of moisture, boron carbide, B4C,
is subject to oxidation to boron oxide, B203. In the event of subsequent
significant steam inleakage, the boron oxide is converted to volatile
boric acid, which is capable of being steam-distilled from the core. Such
an occurrence could produce an increase in core reactivity proportional to
the loss of B10
The criterion used to establish the curve of Figure 4.2.11-1 was that
not more than 10% of the beginning of life (BOL) B10 loading can be present
as oxide over a refueling cycle. This criterion is based on the BOL B10
worth of 0.06 AK, and the fact that 10% worth, 0.006 AK, is substantially
4.2-15
less than the minimum core shutdown margin of 0.016 AK (FSAR Section 3.4.3.1),
and only one-half of the reactivity anomaly of 0.012 AR specified in Technical
Specification SR 5.1.4.
The limits of Figure 4.2.11-1 plus the stipulation that conditions of
high moisture, > 10 ppm, and reactor average outlet temperatures between
725°F and 1200°F shall not exceed a total of 90 days per refueling cycle
assures that no more than 10% of the BOL B10 loading can be oxidized to
B203.
Specification LCO 4.2.12 - Liquid Nitrogen Storage, Limiting Conditions
for Operation
The reactor shall not be operated at power if the liquid nitrogen
storage tank level drops below 500 gallons.
Basis for Specification LCO 4.2.12
Adequate liquid nitrogen storage is provided to permit depressurization
of the PCRV via the helium purification system, assuming complete loss of all
nitrogen recondensing capability. (FSAR, Section 9.6.6) .
Continued cooling of the low temperature adsorbers is not required in
the event all refrigeration is lost, insofar as the heatup due to decay heat
would take more than a week to reach a temperature level above design cond-
itions. This source of heat can be used to regenerate the LTA, transfixing
the source of heat to the waste system. (FSAR, Section 9.4.6) .
4.2-15A
•
PERMISSIBLE MOISTURE CONCENTRATION
versus AVERAGE CORE OUTLET TEMPERATURE
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Average Core Outlet Temperature, °F
Figure 4.2.11-1-
4.2- 16
a) With only one complete operating loop (both heat exchangers
and at least one pump in service) , reactor power may be retained
at rated power for up to 48 hours. If two loop operation cannot
be restored within 48 hours, the reactor must be shut down in an
orderly manner.
b) If one of the two heat exchangers in the single operating
loop becomes inoperable, an orderly shutdown must be started
immediately.
Basis for Specification LCO 4.2.13
With both loops of the PCRV Liner Cooling System functioning, and
with a minimum of one operating pump and one heat exchanger in each loop,
the heat flux in the concrete will be distributed as designed. Analytical _
calculations indicate that operation at full power with one cooling loop for
48 hours satisfies the criterion which specifies a maximum temperature
increase of 20°F in the bulk of the PCRV concrete. Operation under these
conditions also satisfies cooling requirements in the event of a loss-of-
forced-circulation accident.
Specification LCO 4.2.14 - PCRV Liner Cooling Tubes, Limiting Conditions
for Operation
The reactor shall not be operated at power with two adjacent PCRV
liner cooling tubes inoperable.
Basis for Specification LCO 4.2.14
PCRV liner cooling tube failures, whether the result of leakage or
blocking, do not affect the integrity of the PCRV as long as such a
failure is limited to a single tube in any set of four on the side walls,
or any set of six on the PCRV liner top head and core support floor
4.2- 17
top casing. In this case, the local temperature in the concrete would
be less than 250°F, an allowable and acceptable concrete temperature.
(FSAR Section 5.4.5.3) .
Specification LCO 4.2.15 - PCRV Cooling Water System Temperatures ,
Limiting Conditions for Operation
The limiting conditions for the PCRV cooling water system temperatures
utilize the following water temperature definitions:
Inlet Water Temperature - is the water temperature measured
at the common PCRV cooling water heat exchanger outlet in
each loop.
Outlet Water Temperature - is the water temperature measured
at the common PCRV cooling water discharge from the Core
Support Floor, Lower Barrel Section, and Upper Barrel Section
and Top Head in each loop.
The temperature of the PCRV cooling water system shall be maintained
within the limits stated below, irrespective of whether the reactor is
operating or shut down:
a) The maximum temperature difference between the outlet water
temperature of the PCRV cooling water system, and the PCRV external
concrete surface temperature, averaged over 24 hours, shall not
exceed 50°F.
b) The maximum outlet water temperature of the PCRV cooling water
system shall not exceed 120°F.
c) The maximum temperature difference between the outlet water
temperature and the inlet water temperature of the PCRV cooling
water system shall not exceed 20°F.
4.2-18
d) The maximum rate of change of the PCRV concrete temperature
shall not exceed 14°F per week, as indicated by the weekly
average outlet water temperature of the PCRV cooling water
system.
e) The minimum average of the inlet and outlet cooling water
temperatures shall be greater than or equal to 100°F.
Basis for Specification LCO 4.2.15
During normal operation the PCRV concrete will experience non-uniform
temperature distribution due to unavoidable heat losses from the primary
system. These non-uniform temperatures result in thermal stresses in
the self-strained structure, but these stresses tend to relax due to
creep and other inelastic effects, particularly in areas of local stress
concentration. Therefore, only the bulk temperature of the PCRV concrete
is considered in establishing the acceptable thermal loading of the PCRV.
In addition to temperature gradients through the walls, the
concrete temperature varies locally between cooling tubes; this, however,
involves only a small amount of concrete. The cooling system specification
has therefore been prescribed such that the temperature of the concrete
between cooling tubes is limited to 150°F. In certain cases, local
concrete temperatures of 250°F would be acceptable, if the affected
area is small, since the resulting possible small loss in concrete strength
can be tolerated.
Due to the very large bulk of concrete, and the relatively long
time-constant for response for temperature changes , short-term
variations in the temperature of the air in contact with the vessel,
or the PCRV liner, can be tolerated without development of undesirable
4.2— 19
stresses. Similarly, significant changes in the bulk concrete temperature
must be performed slowly such that the average bulk temperature changes
at a rate no greater than 14°F per week.
The most highly irradiated portions of the liner (at the top head)
will be subjected to an integrated neutron dose of approximately
2.3 x 1018 n/cm2 (E > 1 MEV) during the life of the plant. The liner
materials have an initial nil ductility transition (NDT) temperature
of at least minus 60°F which is 160°F below the minimum operating
temperature. The 160°F value allows for a shift in the NDT of 100°F
and provides for operation above the fracture transition elastic
temperature (FIE = NDT + 60°F). This provision will ensure that crack
propagation in the liner at any tensile membrane stress up to yield
stress would be incredible, and in this respect the liner meets the
same criteria as are prescribed for steel nuclear pressure vessels, but
is more conservative since the liner is in general compression for all
normal operation modes.
Specimens from plates representing several heats utilized in the
liner construction were irradiated and evaluated by the Naval Research
Laboratory in the Union Carbide Research Reactor at Oak Ridge. The
30 ft.-lb. transition temperature values in the cases of plates exposed
to 2.5 x 1018 n/cm2 did not rise above 0°F so that the transition
temperature + 60°F criteria limit would be far below the normal operating
temperature.
Limiting the average cooling water temperature to 100°F and the
maximum inlet to outlet temperature difference to 20°F ensures that
the top head liner material average temperatures will be in excess of J00°F
etL all times.
4.3-1
4.3 SECONDARY REACTOR COOLANT SYSTEM - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to the minimum configuration and characteristics of the
secondary (steam) reactor coolant system, including the steam generators
and turbine plant.
Objective
To ensure the capability of this system to cool the core and
prevent a safety limit from being exceeded by defining the minimum
operable equipment and characteristics of the secondary reactor
coolant system.
Specification LCO 4.3.1 - Steam Generators, Limiting Conditions for
Operation
The reactor shall not be operated at power unless both the reheater
section and the economizer-evaporator-superheater (EES) section of one
steam generator and either the reheater section or the EES section of
the other steam generator is operable for the removal of decay heat.
The operable EES sections shall be capable of receiving water
from either the emergency condensate header or the emergency feedwater
header. The operable reheater sections shall be capable of receiving
water from the emergency condensate header.
Basis for LCO 4.3.1
The steam generators provide the means for shutdown heat removal
from the primary coolant. Either the reheater section or the EES
section of one steam generator can be used for this purpose. The reheater
section can be supplied water from the emergency condensate header,
and the EES section can be supplied water from either the emergency
condensate header or the emergency feedwater header.
4.3-2
Specification LCO 4.3.2 - Boiler Feed Pumps, Limiting Conditions for
Operation
The reactor shall not be operated at power unless at least two
of the three boiler feed pumps are operable. If the motor driven feed
pump is not operable and cannot be made operable within 24 hours, the
auxiliary boiler shall be put into operation.
Basis for Specification LCO 4.3.2
Any one of the boiler feed pumps can furnish feedwater for helium
circulator motive power and steam generator heat removal to provide
for shutdown cooling of the plant. One circulator, operating with
feedwater motive power, would provide sufficient primary coolant
circulation following a postulated depressurization accident. In
order to guard against an accident involving rupture of the cold
reheat line, (i.e. , reactor steam cannot be supplied to the turbine-driven
feed pumps), either the motor driven feed pump must be operable or the
auxiliary boiler must be operated to supply motive steam to the feed
pump turbine if required for a plant shutdown (refer to FSAR Section
6.2 and 10.1).
Specification LCO 4.3.3 - Steam/Water Dump Tank Inventory, Limiting
Condition for Operation
The reactor shall not be operated at power if the steam/water dump
tank contains an inventory of condensate corresponding to a level
indication exceeding 45 inches.
Basis for Specification LCO 4..3.3
The condensate inventory maintained in the steam/water dump tank
serves to cool the fluid dumped from a steam generator in the event of a
tube failure. No minimum level is required since the final pressure
4.3-3
after a dump into a dry vessel would not lift dump tank safety valves.
A maximum level of 45 inches corresponding to about 2100 gallons,
is established to prevent operation of safety valves due to hydro-
statically filling the tank during a steam/water dump.
Specification LCO 4.3.4 - Emergency Condensate and Emergency Feedwater
Headers Limiting Conditions for Operation
The reactor shall not be operated at power unless the emergency
condensate header and the emergency feedwater header are operable.
Basis for Specification LCO 4.3.4
A safe shutdown of the plant can be performed with water supplied
to a steam generator via the normal feedwater line, the emergency
feedwater line, or the emergency condensate line. In the event of
failure of the normal feedwater line (FSAR Section 10.3.6) , the
availability of either the emergency feedwater or condensate lines
provides adequate shutdown capability. In the event of a maximum
tornado (FSAR Section 10.3.9) the emergency condensate line and the
emergency feedwater line provide redundant flow paths for steam
generator supply from the Firewater System.
Specification LCO 4.3.5 - Storage Ponds, Limiting Condition for Operation
The reactor shall not be operated at power unless the inventory in
the circulating water makeup storage ponds is at least 20 million gallons
of water.
Basis for Specification LCO 4.3. 5
The fire water system serves as a backup means of supplying emergency
motive capacity to operate a circulator water turbine(s) and supply
emergency cooling water to safely cool the reactor in the event of a
4.3-4
"Maximum Tornado" or"Safe Shutdown Earthquake" (FSAR Section 10.3.9).
The storage ponds are required to supply a source of water for the
fire water system.
Specification LCO 4.3.6 - Instrument Air System - Limiting
Condition for Operation
The reactor shall not be operated at power unless at least two
instrument air compressors, their associated air receivers, and two
main air headers to the reactor building and turbine building
are operable.
Basis for Specification LCO 4.3.6
The instrument air system is required for air supply to the
essential instrumentation required for safe shutdown cooling (as --
discussed in Section 10.3.9 of the FSAR). The description of this
system is presented in FSAR Section 9.9.
Specification LCO 4.3.7 - Hydraulic Power System, Limiting
Conditions for Operation
In each of the two hydraulic power loops at least one hydraulic
fluid pump, one hydraulic valve accumulator servicing each group
of valves, and the associated headers shall be operable during
power operation. If both hydraulic fluid pumps or both accumulators
servicing a group of valves should become inoperable in one hydraulic
power loop, the affected secondary coolant loop shall be shut down
immediately.
Basis for Specification LCO 4.3.7
One hydraulic fluid pump or cne hydraulic valve accumulator and
the associated header serving each group of valves in each loop
assures an adequate supply of hydraulic fluid for safe snutdown cooling,
4.3-5
or for actuation of the steam water dump valves in the event of a
steam leak requiring steam water dump.
Specification LCO 4.3.8 - Secondary Coolant Activity, Limiting
Conditions for Operation
The secondary coolant activity level shall be limited to 0.009
uCi/cc of 1131 and 6.8 uCi/cc of tritium.
Basis for Specification LCO 4.3.8
The limit on the secondary coolant activity has been established
to limit the exclusion area boundary dose to less than the suggested
limits in the event of the accident involving loss of outside power,
main turbine trip, and failure of one diesel generator to start
(FSAR Section 10.3.2) . In that event, about 52,000 gallons of water
would be vented to the atmosphere as steam. Assuming a dilution
factor of 2.7 x 10-3, no partition factor of the iodine between the
steam released and the water not released, a two hour exposure dose
of about 1.5 Rem to the thyroid would be obtained. Using the same
assumptions for tritium a two hour exposure dose of about 0.5 Rem
to the whole body would be obtained.
4.4-1
4.4 INSTRUMENTATION AND CONTROL SYSTEMS - LIMITING CONDITIONS FOR
OPERATION
Applicability
Applies to the plant protective system and other critical
instrumentation and controls.
Objective
To assure the operability of the plant protective system and other
critical instrumentation by defining the minimum operable instrument
channels and trip settings.
Specification LCO 4.4.1 - Plant Protective System Instrumentation, Limiting
Conditions for Operation
The limiting conditions for the plant protective system instrumentation
are shown on Tables 4.4-1 through 4.4-4. These tables utilize the
following definitions:
Degree of Redundancy - Difference between the number of operable
channels and the minimum number of operable channels which when tripped
will cause an automatic system trip.
Operable Channel - A channel is operable if it is capable of
fulfilling its design functions.
Inoperable Channel - Opposite of operable channel.
Tables 4.4-1 through 4.4-4 are to be read in the following manner: If
the minimum operable channels or the minimum degree of redundancy for each
functional unit of a table cannot be met or cannot be bypassed under the
stated permissible bypass conditions, the following action shall be taken:
For Table 4.4-1, the reactor shall be shut down within 12 hours.
4.4-2
For Table 4.4-2, the affected loop shall be shut down within
12 hours.
For Table 4.4-3, the affected helium circulator shall be shut
down within 12 hours.
For Table 4.4-4, the reactor shall be shut down within 24 hours.
4.4-3
Specification LCO 4.4-1
TABLE 4.4-1
INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM, SCRAM
MINIMUM MINIMUM PERMISSIBLE
TRIP OPERABLE DEGREE OF BYPASS
NO. FUNCTIONAL UNIT SEriING CHANNELS REDUNDANCY CONDITIONS
la. Manual (Control Room) -- 1 0 None
lb Manual (Emergency Board) -- 2 (f) 1 None
2. Startup Channel-High < 105 cps 2 1 Reactor Mode
Sw. in "RUN"
3a. Linear Channel-High, < 140% power 2 (f) 1 None
Channels 3, 4, 5 Ta)
3b. Linear Channel-High, < 140% power 2 (f) 1 None
Channels 6, 7, 8 (a)
4. Primary Coolant Moisture
High Level Monitor and < 500 vpm (a) 1 (f) 1 (c) None
Loop Monitor < 100 vpm 2/Loop (f) 1/Loop (h)
5. Reheat Steam Temperature < 1075°F (a) 2 (b) (f) 1 None
- High (b)
6. Primary Coolant Pressure < 50 psig below 2 (f) (k) 1 Less than 30%
- Low normal, load rated power
programmed (a)
7. Primary Coolant Pressure < 7.5% above 2 (f) (k) 1 None
- High normal rated,
load programmed
(a)
8. Hot Reheat Header > 35 psig 2 (f) 1 Less than 30%
Pressure - Low rated power
9. Main Steam Pressure > 1500 psig 2 (f) 1 Less than 30%
- Low rated power
10. Plant Electrical (d) 2 (e) (£) ]. None
System-Loss
11. Two Loop Trouble -- 2 1 None
12. High Temperature, _ 325°F 2 (f) 1 None
Pipe Cavity
4.4-4
Specification LCO 4.4.1
TABLE 4.4-2
INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM,
LOOP SHUTDOWN
MINIMUM MINIMUM PERMISSIBLE
TRIP OPERABLE DEGREE OF BYPASS
NO. FUNCTIONAL UNIT SETTING CHANNELS REDUNDANCY CONDITIONS
la. Steam Pipe Rupture Under ≤ 9 v. dc.
2 (f) (s) 1 None
PCRV, Loop 1 (j)
lb. Steam Pipe Rupture Under 2 (f) (s) 1 None
PCRV, Loop 2 (j) < 9 v. dc.
lc. Steam Pipe Rupture, North 2 (f) 1 None
Pipe Cavity Loop 1 (j) < 9 v. dc.
Id. Steam Pipe Rupture, South 2 (f) 1 None
Pipe Cavity Loop 1 (j) < 9 v. dc.
le. Steam Pipe Rupture, North 2 (f) 1 None
Pipe Cavity Loop 2 (j) .19 v. .dc.
If. Steam Pipe Rupture, South 2 (f) 1 None
Pipe Cavity Loop 2 (j) ≤ 9 v. dc,
2a. High Pressure, Pipe < 2.5" w.g. 2 (f) 1 None
Cavity (j)
2b. High Temperature, Pipe < 130°F 2 (f) 1 None_
Cavity (j)
2c. High Pressure, Under < 2.5" w.g. 2 (f) 1 None
PCRV (j)
2d. High Temperature, Under < 130°F 2 (f) 1 None
PCRV (j)
2 1 None
3a. Loop 1 Shutdown Logic --
2 1 None
. 3b. Loop 2 Shutdown Logic --
4a. Circulator IA and 1B Circ. lA & 1B 2
1 None
Shutdown -Loop Shutdown
Shutdown Logic
4b. Circulator 1C and 1D Circ. 1C & 1D 2
1 None
Shutdown -Loop Shutdown
Shutdown Logic
4.4-5
Specification LCO 4.4-1
TABLE 4.4-2 (continued)
MINIMUM MINIMUM PERMISSIBLE
TRIP OPERABLE DEGREE OF BYPASS
NO. FUNCTIONAL UNIT SEr1ING CHANNELS REDUNDANCY CONDITIONS
5a. Steam Generator < 810 psig 2 (f) 1 None
Penetration Overpressure
Loop 1
5b. Steam Generator < 810 psig 2 (f) 1 None
Penetration
Overpressure Loop 2
6a. High Reheat Header < 5 mr/hr Above 2 (f) 1 None
Activity, Loop 1 Background
6b. High Reheat Header < 5 nr/hr Above 2 (f) 1 None
Activity, Loop 2 Background
7a. Low Superheat Header >800°F 2 (f) 1 Less than 30%
Temperature, Loop 1 (p) Rated Power
7b. Low Superheat Header > 800°F 2 (f) 1 Less than 30%
Temperature, Loop 2 (p) Rated Power
7c. High Differential Temp. < 50°F 2 (f) 1 None
Between Loop 1 and
Loop 2 (p)
4.4-6
Specification LCO 4.4-1
TABLE 4.4-3
INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM, CIRCULATOR TRIP
MINIMUM MINIMUM PERMISSIBLE
TRIP OPERABLE DEGREE OF BYPASS
NO. FUNCTIONAL UNIT SETTING CHANNELS REDUNDANCY CONDITIONS
1. Circulator Speed-Low 1910 rpm Below 2 (f) 1 Less than 30%
(r) Normal as Rated Power
Programmed by
FW Flow
2a. Loop 1, Fixed Feedwater 20% of Rated 2 (f) 1 Less than 30%
Flow-Low (Both Full Load Rated Power
Circulators)
2b. Loop 2, Fixed Feedwater 20% of Rated 2 (f) 1 Less than.30%
Flow-Low (Both Full Load Rated Power
Circulators)
3. Loss of Circulator > 475 psid 2 (f) 1 None
Bearing Water (r)
4. Circulator Penetration < 810 psig 2 (f) 1 None
Trouble (r)
5. Circulator Drain 1 5 psid 2 (f) 1 None
Malfunction (r)
6. Circulator Speed-High < 11,000 rpm 2 (£) 1 None
Steam (r)
7. Manual --- 1 0 None
8. Circulator Seal > -10"HZO, or 2 (f) 1 Opposite loop
Malfunction (r) < 80"H2O d Shutdown
9. Circulator speed-
high water < 11,500 rpm 2 (f)* 1* None
* Minimum operable channels and minimum degree of redundancy must be maintained on at
least one helium circulator per loop. If the minimum number of channels and the
minimum degree of redundancy are not maintained as required, reactor power shall be
reduced to 50% of rated thermal power within 12 hours.
4.4-7
Specification LCO 4.4-1
TABLE 4.4-4
INSTRUMENT OPERATING REQUIREMENTS
FOR REACTOR PROTECTIVE SYSTEM, ROD WITHDRAWAL PROHIBIT (RWP)
MINIMUM MINIMUM PERMISSIBLE
TRIP OPERABLE DEGREE OF BYPASS
NO. FUNCTIONAL UNIT SETTING CHANNELS REDUNDANCY CONDITIONS
1. Startup Channel-Low > 2.5 cps 2 1 Above 10-3%
count rate — Rated Power
2a. Linear Channel-Low power RWP > 5% (m) 2 1 (g)
(Channels 3, 4 and 5)
2b. Linear Channel-Low power RWP > 5% (m) 2 1 (g)
(Channels 6, 7 and 8)
3a. Linear Channel-High power RWP < 30% (n) 2 (f) 1 None
(Channels 3, 4 and 5
3b. Linear Channel-High power RWP < 30% (n) 2 (f) 1 None
(Channels 6, 7 and 8)
v
4.4-8
Specification LCO 4.4.1
NOTES FOR TABLES 4.4-1 THROUGH 4.4-4
(a) See Specification LSSS3.3 for trip setting.
(b) Two thermocouples from each loop, total of four, constitute one channel.
For each channel, two thermocouples must be operable in at least one
operating loop for that channel to be considered operable.
(c) With one primary coolant high level moisture monitor tripped, trips of
either loop primary coolant moisture monitors will cause full scram.
Hence, number of operable channels (1) minus minimum number required
to cause scram (0) equals one, the minimum degree of redundancy.
(d) Both 480 volt buses lA and 1C less than 6Q% normal voltage for longer
than 30 seconds.
(e) One channel consists of one undervoltage relay from each of the two
480 volt buses (two undervoltage relays per channel). These relays
fail open which is the direction required to initiate a scram.
(f) The inoperable channel must be in the tripped condition, unless the
trip of the channel will cause the protective action to occur.
(g) RWP bypass permitted if the bypass also causes associated single
channel scram.
(h) Permissible Bypass Conditions:
I. Any circulator buffer seal malfunction.
II. Loop Hot Reheat Header High Activity.
(j) Items la. or lc. or ld. accompanied by 2a. , 2b. , 2c. , or 2d. on Table
4.4-2 are required for loop 1 shutdown. Items lb. or lc. or lf. ,
accompanied by 2a. , 2b. , 2c. , or 2d. on Table 4.4-2 are required for
loop 2 shutdown.
(k) One operable helium circulator inlet thermocouple in an operable loop
is required for the channel to be considered operable.
(m) Low Power RWP bistable resets at 4% after reactor power initially
exceeds 5%.
(n) Power range RWP bistable resets at 10% after reactor power initially
exceeds 30%.
(p) Item 7a. must be accompanied by item 7c for Loop 1 shutdown.
Item 7b. must be accompanied by item 7c. for loop 2 shutdown.
4.4-8a
Notes for Tables 4.4-1 through 4.4-4 (continued)
(r) Separate instrumentation is provided on each circulator for this
functional unit. Only the affected helium circulator shall be
shut down within 12 hours if the indicated requirements are not met.
(s) Each channel has 2 microphones running in parallel with one ultrasonic
amplifier. For the channel to be considered operable, both microphones
and the amplifier must be operable.
4.4-9
Basis for Specification LCO 4.4.1
The plant protection system automatically initiates protective functions
to prevent established limits from being exceeded. In addition, other
protective instrumentation is provided to initiate action which mitigates
the consequences of accidents. This specification provides the limiting
conditions for operation necessary to preserve the effectiveness of these
instrument systems.
If the minimum operable channels or the minimum degrees of redundancy
for each functional unit of a table cannot be met or cannot be bypassed
under the stated permissible bypass conditions, the following action shall
be taken:
For Table 4.4-1, the reactor shall be shut down within 12 hours.
For Table 4.4-2, the affected loop shall be shut down within 12 hours.
For Table 4.4-3, the affected helium circulator shall be shut down
within 12 hours.
For Table 4.4-4, the reactor shall be shut down within 24 hours.
If, within the indicated time limit, the minimum number of operable
channels and the minimum degree of redundancy can be reestablished, the
system is considered normal and no further action needs to be taken.
The trip level settings are included in this section of the specifi-
cation. The bases for these settings are briefly discussed below.
Additional discussions pertaining to the scram, loop shutdown and circulator
trip inputs may be found in Section 7.1.2.3, 7.1.2.4, and 7.1.2.6, respectively,
of the FSAR. High moisture instrumentation is discussed in Section 7.1.2.5•
4.4-10
a) Scram Inputs
Manual Scram is provided to give the operator means for emergency
shutdown of the reactor independent of the automatic reactor protective
system.
Startup Channel-High Countrate is provided as a scram input during
fuel loading and zero power operations.
Linear Channel Flux-High (See Technical Specification LSSS 3.3) .
High Reactor Moisture (See Technical Specification LSSS 3.3) .
High Reheat System Temperature (See Technical Specification LSSS 3.3) .
Low Reactor Pressure is an indication of possible helium leakage from
the system. A scram is required because the reactor is in danger of being
inadequately cooled which would increase the hazard associated with activity
release from the PCRV. The trip is programmed with plant load (similar to
the high pressure trip) to reduce the response time when the plant is at
high power. The low pressure trip point is 50 psi below normal during
operation between 30% and 100% rated power which is lower than the pressures
reached on normal transient conditions.
High Primary Coolant Pressure (See Technical Specification LSSS 3.3) .
Low Hot Reheat Steam Pressure is an indication of either a cold reheat
steam line rupture or a hot reheat steam line rupture and necessitates plant
shutdown due to the potential loss of steam turbine circulator motive
power. The trip point is selected to be below normal operating levels
which vary over a wide range.
Low Main Steam Pressure is an indication of main steam line rupture or
loss of feedwater flow and necessitates plant shutdown due to potential loss
of steam turbine circulator motive power. The trip point is selected to be
below normal operating levels.
4.4-11
Plant Electrical System Power Loss requires a scram to prevent any
power-to-flow mismatches from occuring. A 30-second delay is provided
following a power loss before the scram is initiated to allow the emergency
diesel generator to start. If it does start, the scram is avoided.
Two-Loop Trouble. Operation on one loop at a maximum of about 50%
power may continue following the shutdown of the other loop (unless
preceded by scram as in the case of high moisture.) Onset of trouble in
the remaining loop (two-loop trouble) results in a scram. Trouble is
defined as a signal which normally initiates a loop shutdown. Similarly,
simultaneous shutdown signals to both loops result in shutdown of one
of the two loops only and a reactor scram.
High Temperature in the pipe cavity would indicate the presence
of an undetected steam leak or the failure of the steam pipe rupture
detection system to differentiate in which loop the leak had occurred
and to shut the faulty loop down.
The setpoint has been set above the temperature that would be expected
to occur in the pipe cavity if the steam leak were detected and the faulty
loop shutdown for all steam leaks except those of major proportion or due
to an offset rupture of one of the steam lines.
An undetected steam leak or pipe rupture under the PCRV within the
support ring would also be detectable in the pipe cavity, therefore
only one set of sensors and logic is required to monitor both areas.
b) Loop Shutdown Inputs
Steam Pipe Rupture In The Reactor Building necessitates shutdown
of the leaky loop to terminate the pressure and temperature buildup
within the building. Ultrasonic noise caused by escaping steam in
conjunction with a pressure or temperature rise will cause the appropriate
loop to shutdown.
4.4-12
The trip of the ultrasonic detection system is set at a level which
corresponds to 9 v. dc. output from the ultrasonic amplifier. The
pressure and temperature trips are set above normal operating building
pressure and temperature levels.
Shutdown of Both Circulators is a loop shutdown input which is necessary
to insure proper action of the reactor protective (scram) system (through
the two-loop trouble scram) in the event of the loss of all circulators and
low feedwater flow.
The remaining loop shutdown inputs are equipment protection items
which are included because their malfunction could prevent a scram due to
loss of the two-loop trouble scram input.
c) Circulator Shutdown Inputs
All circulator shutdown inputs (except circulator speed high on water
turbines) are equipment protection items which are tied to two loop trouble
through the loop shutdown system. These items are included in Table 4.4-3
because a malfunction could prevent a scram due to loss of the two loop
trouble scram input. Circulator speed high on water turbines is included
to assure continued core cooling capability on loss of steam drive.
d) Rod Withdraw Prohibit Inputs
Startup Channel Countrate-Low is provided to prevent control rod
withdrawal and reactor startup without adequate neutron flux indication.
The trip level is selected to be above the background noise level.
Linear Channel (5% Power) directs the operator's attention to either
a downscale failure of a power range channel or improper positioning of the
I.S.S.
Linear Channel (30% Power) is provided to prevent control rod withdrawal
if reactor power exceeds the I.S.S. limit for the "Low Power" position.
4.4-13
Specification LCO 4.4.2 - Control Room Temperature - Limiting Condition
for Operation
The reactor shall not be operated at power if the control room
temperature exceeds 120°F.
Basis for Specification LCO 4.4.2
The limiting temperature in the control room is established to
assure no over temperature condition which might cause damage to
essential instrumentation and control equipment. Satisfactory operation
of safety related control and, electrical equipment located in the
control room for temperatures up to 120°F is discussed in FSAR Amendment
No. 17, Question 7.5.
Specification LCO 4.4.3 - Area Radiation Monitors - Limiting Condition
for Operation
At least one area radiation monitor from each group shall be operable.
If any area monitor becomes inoperable, a portable monitor equipped with
an alarm shall be placed in the area, and all personnel notified of the
condition.
Basis for Specification LCO 4.4.3
The grouping of area radiation monitors is such that each monitor
in the group supplements the others in the group.
The notification of personnel of any malfunction, coupled with the
provision of a portable instrument , or a replacement, adequately ensures
protection for personnel, and detection of abnormalities.
The detectors are grouped as follows:
4.4-14
GROUP NO. DETECTOR NO. LOCATION
1 RT-93250-1 4881 Refueling Machine Control Room
1 RT-93252-1 4881 East Wall
1 RT-93251-1 4864 Reactor Plant Exhaust Filter Room
1 RT-93252-2 4864 South Stairwell
a
2 RT-93250-3 4856 Hot Service Facility
2 RT-93251-3 4868 Hot Service Facility
3 RT-93250-2 4854 East Walkway
3 RT-93250-4 4839 East Walkway
3 RT-93251-4 4816 Office Building
3 RT-93252-4 4829 Analytic Instrument Room
4 RT-93250-13 4791 Condensate Demineralizers
1+ RT-93250-5 4829 Main Control Room
4 RT-93251-6 4791 Grade Floor North
4 RT-93252-6 4791 South Stairwell
5 RT-93251-5 4781 East Walkway
5 RT-93251-7 4781 Valve Operating Station - West
5 RT-93252-7 4781 Valve Operating Station - East
6 RT-93250-8 4771 Northeast Walkway
6 RT-93251-8 4771 Radiochem Lab
6 RT-93251-9 4740 North Stairwell
4.4-15
Specification LCO 4.4.4 - Seismic Instrumentation - Limiting Conditions
for Operation
The reactor shall not be operated at power unless three (3) of
the six (6) seismic instruments are operable.
Basis for Specification LCO 4.4.4
The monitoring provided by three (3) seismic instruments, in the
event of an earthquake, is adequate to determine the ground acceleration
at the site.
4.5-1
4. 5 CONFINEMENT SYSTEM - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to the minimum operable equipment of the reactor building
(confinement) , and the ventilation system.
OW ective
To assure the operability of the confinement systems.
Specification LCO 4.5.1 - Reactor Building, Limiting Conditions
for Operation
The plant shall not be operated at power; reactor vessel internal
maintenance shall not be performed with irradiated fuel in the PCRV; or
irradiated fuel handling shall not be performed within the reactor
building unless:
a) Reactor Building Integrity is maintained as follows :
1. Personnel access to the building is controlled.
2. The reactor building pressure is sub-atmospheric.
3. The reactor building louvers are closed and the
"pressure set point" is at 3 inches of water.
4. When the truck doors to the truck bay are open, the reactor
floor hatch, the deck hatch and all personnel doors in the
truck bay are closed.
5. When the reactor floor hatch and/or the deck hatch are open,
the truck doors and external personnel doors in the truck
bay are closed.
b) Two of the three reactor building exhaust fans are operable.
4.5-2
Basis for Specification LCO 4.5.1
The integrity of the reactor building and operation of the ventilating
system in combination limit the off-site doses under normal and abnormal
conditions. In the unlikely event of a major release of activity from the
PCRV, the combination of the reactor building and ventilation system would
act to keep off-site doses well below 10 CFR 100 limits (see FSAR Section
14.10.3.4).
The pressure in the reactor building is held slightly below atmospheric
pressure. Exfiltration would occur only above a wind velocity of about
30 mph. Wind conditions within the range of 0 to 25 mph prevail at the
site about 98% of the time. The mechanical turbulence from wind speeds
of 25 mph or higher would result in a dilution better than during lesser
wind speed conditions for any nuclides exfiltrated from the reactor
building. (FSAR Section 6.1.4.2)
The purpose of the pressure relief device is to maintain the integrity
of the reactor building by relieving the pressure inside the building when
it equals or exceeds 3 inches of water. In the unlikely event of the
occurrence of a rapid increase of pressure inside the building of or
exceeding 3 inches of water, the louvers would open, relieving the pressure,
and then be automatically closed at approximately atmospheric pressure
(or they can be manually closed) , restoring the integrity of the reactor
building (see FSAR 6.1.3.4) and maintaining the potential doses from the
occurrence to as low as practicable.
The building ventilation system maintains the reactor building
pressure slightly subatmospheric and reduces the amount of radioactivity
released to the environment, during normal operation or accident conditions.
4.5-3
Specification LCO 4.5.2 - Reactor Vessel Internal Maintenance,
Limiting Conditions for Operation
During any reactor vessel internal maintenance with irradiated fuel
within the vessel which requires removal of both primary and secondary
closures, the following conditions shall be met:
a) The reactor is depressurized to atmospheric pressure or slightly
below.
b) The reactor average helium gas inlet temperature is 165°F or less.
c) The reactor is maintained in a reactor shutdown or refueling
shutdown condition and the reactivity of the core is monitored
continuously by at least two neutron flux monitors capable of
continuously indicating the neutron flux level within the core.
If any of these conditions cannot be met, internal maintenance in
the reactor vessel shall be terminated and any remote operated mechanisms
shall be retracted and the opening through the PCRV closed as soon as
practicable.
Basis for Specification LCO 4.5.2
In order to prevent the outleakage of primary coolant and potential
release of activity during in-vessel maintenance, the reactor must be
depressurized and maintained at or slightly below atmospheric pressure.
4.6-1
4.6 AUXILIARY ELECTRIC POWER SYSTEM - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to the minimum operable equipment supplying electric power
to the plant auxiliaries.
ObJective
To ensure that the capability of supplying electric power to the
plant auxiliaries is maintained by defining the minimum operable equipment.
Specification LCO 4.6.1 - Auxiliary Electric System, Limiting Conditions
for Operation
The reactor shall not be operated at power unless the following
conditions are satisfied:
a) Both the Unit Auxiliary and Reserve Auxiliary Transformers are
operable.
The Reserve Auxiliary Transformer can be made inoperable
for 24 hours provided both diesel generator sets (two
engines and associated generator per set) are started
immediately prior to taking the transformer out of service
to verify their operability, are shut down and their controls
left in the automatic mode and all three 480 V a-c essential
buses are operable.
b) 4160 V a-c Bus 1B must be operable.
4160 V a-c Bus 1B may be made inoperable for 12 hours
providing the 480 V a-c Essential buses and both diesel
generator sets are operable, operability to be proved
as in a) above.
4.6-2
c) The auxiliary power 480 V a-c essential buses 1A, 1B and
1C must be operable.
Each essential bus may be inoperable for 12 hours provided
the following conditions are satisfied:
1. Only one 480 V essential bus is inoperable at a time.
2. 4160 V a-c bus 1B is operable.
3. Engine Driven Fire Pump is operable.
4. Emergency Condensate Header is operable.
5. The diesel-generator set(s) supplying the
remaining operable 48o V a-c essential buses
are operable.
6. All equipment supplied by the operable
essential buses, associated with Safe Shutdown Cooling
must be operable.
7. Reactor building exhaust fans supplied from the
operable essential buses must be operable.
d) Both the diesel-generator sets are operable, including the
following:
1. One fuel oil transfer pump from the diesel fuel oil storage
tank to the diesel fuel oil day tanks.
2. One starting air compressor and receiver per
diesel-generator set.
3. Associated automatic load shedding, load programming,
and auto diesel-generator set starting equipment.
4. 500 gallons of fuel in each day tank.
One diesel generator set may be inoperable for up to
7 days (total for both) during any month provided the
4.6-3
operability of the other diesel-generator set is
demonstrated immediately as in a) above and all
essential buses are operable.
e) The two station batteries and their associated buses and
battery chargers are operable.
One battery charger or battery may be inoperable for
24 hours. A battery and battery charger disconnected
from the bus to overcharge the battery is not considered
inoperable if the overcharge period does not exceed
24 hours provided the following conditions are satisfied:
1. The 480 V essential bus supplying power to the
battery charger and the battery charger of the
connected D.C. bus must be operable.
2. D.C. Bus tie breakers must be closed.
3. Diesel-generator set associated with the
480 V essential Bus of 1) above must be operable.
One D.C. bus may be inoperable for 12 hours provided the
following conditions are satisfied:
1. The 480 Volt essential bus supplying power to the
battery charger and the battery of the operable
D.C. bus must be operable.
2. Instrument power inverter supplied from the operable
D.C. bus must be operable.
3. Diesel-generator set associated with the operable
D.C. bus must be operable.
4.6-4
f) Both the instrument inverters are operable. One inverter
may be inoperable for 24 hours.
g) A minimum of 20,000 gallons of fuel in underground storage.
Upon reaching this minimum quantity, the auxiliary boiler
shall be shut down.
h) At least one Boiler Fuel Oil Pump operable between the
auxiliary boiler fuel supply and the diesel fuel oil day tanks.
Both Boiler Fuel Oil Pumps may be inoperable for up
to 24 hours if at least 5,500 gallons are in the
diesel oil storage tank and both fuel oil transfer
pumps between the diesel oil storage tank and the day
tanks are operable.
Basis for Specification LCO 4.6.1
The objective of this specification is to assure that an adequate
source of electrical power is available to operate the plant during
normal operation, for cooling during shutdown, and for operation of
engineered safeguards in emergency situations. There are three sources
of power available for shutdown: unit auxiliary transformer, reserve
auxiliary transformer, and the standby diesel-generator sets.
In the normal operating mode, the unit auxiliary transformer is in
operation, reserve auxiliary transformer is energized and the standby-
generator sets are operable. (FSAR Section 8.2.3.3).
The main turbine-generator can be used as a source of auxiliary
power in the event that outside electrical power is lost.
In the event of loss of all outside power and a turbine-generator
trip, the diesel-generator sets would come on automatically to provide
the required energy necessary to safely shut down the plant.
4.6-5
In the event of the loss of the Reserve Auxiliary Transformer
when the main turbine-generator is out of service, links in the
bus between the main turbine-generator and the main power transformer
can be removed, allowing the main power transformer and unit auxiliary
transformer to be returned to service.
The essential 480 V power source is supplied from three separate
buses, any two of which can supply adequate power to shut the plant
down. Under accident conditions, if the normal supply of power to
these three essential buses should fail, the diesel-generator sets would
come on and energize them. Bus load shedding, breaker closing, and
load sequencing on to the diesel-generator sets is handled automatically.
The station batteries supply power for the instrument power
inverters, protective devices and equipment operational control.
During normal operation, d-c power is supplied by a-c to d-c rectifiers
which also keep the batteries fully charged. (FSAR Section 8.2.2.4)
Backup electric power for the non-interruptible a-c instrument loads
is provided by bus ties from Instrument Bus No. 3 to Instrument Buses
1 and 2 which are normally fed by the two Instrument Power Inverters.
Bus No. 3 receives its power from redundant instrument transformers
which are supplied from the essential 480 volt switchgear.
(FSAR Section 8.2.2.3).
A redundant source of electric power for the d-c instrument loads is
available from a bus tie between the two d-c buses which allows one
battery or a-c to d-c rectifier to supply both buses.
These backups and redundancies permit the temporary removal from
service of an instrument power inverter, a battery, a d-c bus or an
a-c to d-c rectifier.
4.6-6
A diesel fuel storage capacity of 50,000 gallons is provided.
A supply of 20,000 gallons of diesel fuel is adequate to provide
for operation of the standby generators for at least seven days
under required loading conditions. This allows adequate time to
obtain additional fuel and to make provisions to restore the standby
source of power into the station.
4.7-1
4.7 FUEL HANDLING AND STORAGE SYSTEMS - LIMITING CONDITIONS FOR OPERATION
Applicability
Applies to minimum operable equipment and characteristics of the
fuel handling and fuel storage systems during handling and storage of
irradiated fuel.
Objectives
To prevent an uncontrolled release of radioactivity during irradiated
fuel handling and storage by defining the minimum operable equipment and
characteristics.
Specification LCO 4.7.1 - Fuel Handling in the Reactor, Limiting Conditions
for Operation
During any irradiated fuel handling in the reactor vessel, the
following conditions shall be met:
a) The reactor is depressurized to atmospheric pressure or
slightly below.
b) The reactor average helium gas inlet temperature is 165°F or less.
c) The reactor is maintained in a refueling shutdown condition and the
reactivity of the core is monitored continuously by at least two
neutron flux monitors capable of continuously indicating the
neutron flux level in the core.
If any of these conditions can not be met, fuel handling in the reactor
vessel shall be terminated and the Fuel Handling Mechanism will be retracted
into the Fuel Handling Machine and the isolation valve closed as soon as
practicable.
4.7-2
Basis for Specification LCO 4.7.1
In order to prevent the outleakage of primary coolant and potential
release of activity during refueling, the reactor must be depressurized
and maintained at or slightly below atmospheric pressure. In order to
prevent pressurization of the fuel handling equipment exceeding 5 psig
(the maximum allowable working pressure of the fuel handling equipment)
as a result of accidental inleakage of water into the vessel during
refueling, the reactor inlet gas temperature is limited to 165°F.
Specification LCO 4.7.2 - Fuel Handling Machine, Limiting Conditions for
Operation
During any irradiated fuel handling with the Fuel Handling Machine
the following conditions shall be met:
a) The pressure in the Fuel Handling Machine is at approximately
atmospheric pressure.
b) A continuous supply of helium to the Fuel Handling Machine is
available.
If a) or b) above cannot be satisfied, the Fuel
Handling Machine shall be retracted to its uppermost
position and the reactor isolation valve and fuel
handling machine cask valve closed.
c) The cooling water outlet temperature from the Fuel Handling
Machine is 150°F or less.
If c) above cannot be met, immediate action shall
be taken to return the irradiated fuel elements
within the Fuel Handling Machine to the reactor core
or to the Fuel Storage Facility, after which fuel
handling shall be terminated.
4.7-3
Basis for Specification LOO 4.7.2
In order to assure proper operation of the Fuel Handling Equipment
a continuous supply of helium must be provided. The normal operating pressure
of the fuel handling machine must be maintained near the same pressure as
the reactor in order to reduce the potential for a release of activity.
The capability of purging the refueling equipment with air or purified
helium is necessary for proper refueling operations.
The temperature of the irradiated fuel elements removed from the core
during the refueling operation and storage is to be maintained below 750°F
in order to prevent any significant graphite oxidation if there is any air
leakage into the Fuel Handling Machine or fuel storage well. A fuel handling
machine cooling water system with an outlet temperature of < 150°F provides the
proper flow and temperature to maintain the fuel elements below the 750°F.
Specification LCO 4.7.3 - Fuel Storage Facility, Limiting Conditions
for Operation
During storage of irradiated fuel in the Fuel Storage Facility, the
following conditions shall be met:
a) Both cooling water coils must be operating and their outlet
cooling water temperatures 150°F or less, for any storage
well containing irradiated fuel.
b) If only one cooling water coil is operable on a storage well,
irradiated fuel storage is permissable if the outlet cooling
water temperature is 150°F or less, and the ventilation system
is capable of supplying a total of 12,000 CFM to the Fuel
Storage Facility.
c) The fuel storage wells containing irradiated fuel are maintained
at approximately atmospheric pressure.
If the above conditions a) or b) and c) cannot be met for a well or wells
4.7-4
containing irradiated fuel immediate action shall be taken to re-establish
the desired conditions. If the desired conditions have not been re-established
within 24 hours, the irradiated fuel shall be transferred to a storage well
or wells for which the desired conditions can be met.
Basis for Specification LCO 4.7.3
To prevent oxidation of the irradiated fuel, the fuel storage wells are
designed to maintain the irradiated fuel under a dry helium atmosphere.
Overpressurization of a storage well is alarmed to the operator and protection
is provided by relief valves.
The storage well cooling water system is designed with two 100% capability
cooling coils supplied from independent water sources. In addition, in the
event of a complete interruption of cooling to one of the fuel storage wells
as a result of a rupture or blockage of both cooling coils, the affected
storage well would be cooled by increasing the normal ventilation air
flow through the storage vault containing the affected storage well. The
ventilation system is capable of moving air through the vault at a rate of
9000 CFM until water cooling is restored on the well emptied. Normal
ventilation flows of 1500 CFM are maintained through the other two vaults.
LCO 4.7.4 - Spent Fuel Shipping Container, Limiting Conditions for Operation
Loading and shipment of spent fuel prior to 100 days decay time is
allowable if all of the requirements of 10 CFR 71 are met.
Basis for Specification LCO 4.7.4
In complying with the radiation dose limits of 200 mrem/hr at the outer
surface of the cask, as specified in 10 CFR 71, the fuel contained in the
cask will have a total activity which will be less than or equal to the
activity of the most radioactive spent fuel elements contemplated to be
shipped from the plant.
4.7-5
The potential radiological consequences of an accident whereby the
spent fuel shipping cask breaks open while being lowered to the truck,
have been determined (Amendment 17, answer to question 9.5) assuming
that the cask is loaded with such fuel elements.
4.8-1
4.8 RADIOACTIVE EFFLUENT DISPOSAL SYSTEM - LIMITING CONDITIONS FOR
OPERATION
Applicability
Applies to the release of radioactive liquid and gaseous waste
from the plant.
Objective
To assure that the quantity of radioactive material released is
kept as low as practicable and, in any event, within the limits of
10 CFR 20, by defining the condition for release of radioactive
effluents from the plant vent and the radioactive liquid waste system
to the cooling tower blowdown line.
Specification LCO 4.8.1 - Radioactive Gaseous Effluent, Limiting
Conditions for Operation
a) The release of gaseous and airborne particulate effluents
shall be made on an isotopic basis and shall be limited
in accordance with the following equation:
Ci
E r < 3 x 101° cm3
(MPC)i sec
where Ci is the concentration in (pCi/std.cc) of any
radioisotope, i; (MPC)i is in units of uCi/cc as defined
in Column 1, Table II, of Appendix B to 10 CFR 20; and r is
the release rate from the holdup tanks in std.cc/sec.
b) For purposes of calculating permissible release rates by the
above formula, MPC for halogens and particulates with half lives
longer than 8 days will be reduced by a factor of 700 from their
listed value in Column 1, Table II, of 10 CFR 20, Appendix B.
4.8-2
c) Plant equipment in the helium purification system (the high
temperature filter adsorber, the low temperature absorber
and the titanium sponge) , in the gaseous waste system (vacuum
tank, liquid drain tank, filters, compressor, surge tanks and
reactor building filters) , and analytical and monitoring
instrumentation shall be utilized to keep releases of radio-
active materials to unrestricted areas as low as practicable
and to assure surveillance of radioactive gaseous waste produced
during normal reactor operations and expected operational
occurrences.
d) Gaseous radioactive effluents released from the vent shall be
continuously monitored and recorded.
1) During power operation one noble gas monitor and the halogen/
particulate monitor on the plant vent plus the recorder shall
be in operation. If both noble gas monitors, the sampling
mechanism, or recorder becomes inoperable, the reactor must
be shut down within 48 hours. If both halogen/particulate
detectors, the sampling mechanism, or recorder becomes
inoperable, the reactor shall be shut down within 24 hours.
2) If both noble gas monitors are inoperable, grab samples from
the plant vent effluent shall be taken once every twelve
hours and analyzed for gross beta-gamma activity.
e) Except for air ejector discharge, secondary coolant system relief
valves, and deaerator vent, all normal releases of gaseous waste
from the plant shall be filtered through the reactor plant exhaust
particulate and charcoal filters.
4.8-3
f) Under normal operations, the low temperature adsorber, in the
Helium Purification System, shall be isolated and held for
60 days decay prior to regeneration. If abnormal conditions
or equipment failure prevent 60 days holdup, the low temperature
adsorber may be regenerated to the gas waste system with
subsequent release of the noble gases to the environment as
per a) and b) above.
g) The maximum amount of gaseous radioactivity in a gas waste
surge tank shall not exceed 370 equivalent curies of 88Kr.
h) Prior to the release of gaseous radioactivity from the gas
waste surge tanks, the contents shall be sampled and analyzed
to determine compliance with a) and b) above.
i) If during power operation, the air ejector discharge monitor
becomes inoperable, the reactor must be shut down within 48 hours.
When there is indication of a primary to secondary leak through
the steam generator reheater section, a grab sample shall be taken
and analyzed for gross beta-gamma activity once every 12 hours.
j ) Under normal operating conditions, tritium from the H2 Getters
shall be disposed of as a solid waste on an adsorbent material.
k) At least one reactor building exhaust fan shall be operating
whenever releases from the gas waste system through the vent are
taking place.
If condition a) cannot be met, or the vent stack monitors are not
operable, immediate action shall be taken to terminate release from the
gas waste system. If the above conditions cannot be met with this termination
of gas waste system releases, the reactor shall be shut down. If condition
g) cannot be met, immediate action shall be taken to terminate any operations
which result in the production of radioactive gases for storage in the tanks.
4.8-4
Basis for Specification LCO 4.8.1
The major source of gaseous radioactive waste will be the regeneration
of the low temperature filter adsorbers of the Helium Purification System.
The design objective for the plant's radioactive gas releases is 4160
curies per year; 4,120 curies of this are predicted to be long lived
Kr-85 (halflife is 10.8 years) . The limiting value for radioactive gaseous
release is based on (1/annual average dilution factor) . The estimated
800 Ci per year of tritium evolved from H2 Getter regeneration will normally
be disposed of as solid waste. Under unusual conditions, such as a steam
generator tube leak, it may be necessary to release the tritium to the
atmosphere.
The limitation on the curie inventory of a waste gas surge tank is
to limit potential site exclusion radius whole body doses to less than
0.5 rem in the event of a tank rupture.
It is the intent that through these operating limits, the annual
releases from this plant will be as low as practicable and at the same
time the licensee is permitted flexibility of operation, compatible
with considerations of health and safety, to assure that the public is
provided a dependable source of power even under unusual operating
conditions which may temporarily result in releases higher than small
fractions, but still within limits specified in 20.106 of 10 CFR 20.
It is expected that in using this operational flexibility under unusual
operating conditions the licensee will exert his best efforts to keep
levels of radioactive material in effluents as low as practicable.
4.8-5
Specification LCO 4.8.2 - Radioactive Liquid Effluent, Limiting Conditions
for Operation
The following conditions shall be observed regarding the controlled
release of liquid effluent from the radioactive liquid waste system:
a) The maximum instantaneous release rate of radioactive liquid
effluents from the site shall be such that the concentration of
radionuclides in the cooling tower blowdown water discharge does
not exceed the values specified in Table II, Column 2, 10 CFR 20,
Appendix B, for unrestricted areas.
b) The liquid waste filters, monitor tank, receiver tanks and
demineralizers shall be utilized to provide the longest practicable
holdup time, to keep releases of radioactive materials to
unrestricted areas as low as practicable, and to assure a means
of surveillance of radioactive liquid waste produced during normal
plant operations maintenance and expected operational occurrences.
c) Prior to release, duplicate samples of liquid effluent from the
radioactive liquid waste system shall be analyzed isotopically
and for gross beta-gamma activity to determine compliance with a)
above. If the gross concentration of radioactivity in the undiluted
effluent from the radioactive waste monitor tank exceeds 2 x 10-6
uCi/cc, the liquids shall be treated by demineralization prior
to release to the environment.
d) All liquid effluent releases from the radioactive liquid waste
system shall be continuously monitored and recorded and equipment
shall be operable to automatically terminate the release on high
specific activity or low circulating water blowdown flow.
If the conditions of a) and d) above cannot be met, immediate action
shall be taken to terminate the release.
4.8-6
Basis for Specification LCO 4.8.2
Liquid waste from the radioactive waste disposal system are diluted
in the cooling tower blowdown flow. Interlocks from the waste treatment
system discharge valve with the discharge line radiation monitors and the
cooling tower blowdown flow meter will terminate the discharge of waste
in the event of high activity and/or low blowdown flow (less than 1100
gallons per minute (gpm) ).
It is expected that plant releases of radioactive materials and
effluents will be small fractions of the limits specified in 10 CFR 20.106
and will be held as near background levels as practicable.
The design objective of the liquid waste treatment system was to
limit annual liquid waste discharge from the plant to 0.2 Curies which
corresponds to 10 CFR 20 limits.
The liquid waste discharged from the plant will normally flow to
a farm pond (Goosequill Pond) on the north end of the Public Service
Company property near the confluence of St. Vrain Creek and the S. Platte
River. Goosequill Pond drains to the S. Platte River. An alternate
flow path is to a slough which drains to St. Vrain Creek. The operational
environmental monitoring program directs special attention to these areas
so that possible buildup of radioactivity will be detected. It is expected
that releases of radioactive materials in effluents will be kept to small fractions
of the limits specified in 20.106 of 10 CFR 20. At the same time the licensee
is permitted the flexibility of operation, compatible with considerations
of health and safety, to assure that the public is provided a dependable
source of power, even under unusual operating conditions which may
temporarily result in releases higher than such small fractions, but still
within the limits specified in 20.106 of 10 CFR 20.
4.8-7
It is expected that, in using this operational flexibility under
unusual operating conditions, licensee will exert his best efforts to
keep levels of radioactivity in effluents as low as possible.
Specification LCO 4.8.3 - Reactor Building Sump Effluent - Limiting
Conditions for OPeration
The following conditions shall be satisfied regarding the operation
of the reactor building sump pumps and effluent discharge:
a) The discharge from the reactor building sump pumps shall
be filtered and the flow limited to < 10 gpm when operated
in the automatic mode.
b) If effluent discharges from the reactor building sump at flow
rates >10 gpm are to be made, sampling and analysis shall be
required as indicated in LCO 4.8.2, part c).
c) Effluent discharges from the reactor building sump shall not
occur simultaneous with discharge from the Radioactive Liquid
Waste System.
d) All liquid effluent releases from the reactor building sump
shall be continuously monitored and recorded and equipment shall
be operable to automatically terminate the release on high specific
activity or low circulating water blowdown flow.
e) If discharge is to be made in the automatic mode, the sump
discharge line must be proportionally sampled on a continuous
basis. Analysis of the composite sample shall be made three times
per week.
4.8-8
f) If the continuous proportional sampler should be inoperable,
automatic discharge from the sump would be permitted provided
daily samples are taken from the sump and analysis made of the
composite sample three times per week.
g) If any of the above conditions cannot be met, immediate action
shall be taken to terminate discharge from the reactor building
sump. to the circulating water blowdown line.
Basis for Specification LC0 4.8.3
Limiting the discharge flow rate from the reactor building sump to
<10 gpm provides for effluent monitoring sensitivity to assure conformance
with the limits of 10 CFR 20, in the highly unlikely event that the sump
should contain any radioactive liquid.
Sampling and analysis performed prior to discharging from the sump
at flow rates >10 gpm and prohibiting discharge to the radioactive liquid
waste discharge line simultaneous with a discharge from the Radioactive
Liquid Waste System will assure conformance with the limits of 10 CFR 20.
4.9-1
4.9 FUEL LOADING AND INITIAL RISE TO POWER - LIMITING CONDITIONS FOR
OPERATION
Applicability
Applies to all phases of operation from initial fuel loading through
initial full power testing.
Objective
To assure that certain system modifications, testing and documentation
are completed prior to initiating certain phases of the power ascension
program. These phases include (1) fuel loading and low power physics testing
with an air environment, and (3) hot physics tests with a helium environment
and rise to full power testing.
Specification LCO 4.9.1 - Fuel Loading and Initial Rise to Power -
Limiting Condition for Operation
The attainment of full power operation shall be accomplished in the
following phases.
a) Phase 1 - Fuel loading and low power physics testing in an air
environment
These activities shall be performed in an air environment and
reactor operation shall be limited to a power level of 0.1
percent of rated thermal power (0.84 MWt) .
During Phase 1 activities, certain system modifications shall be
concluded and procedural items completed. Completion of these
activities must be approved by the AEC's Regional Regulatory
Operations Office prior to commencement of Phase 2 operations.
4.9-2
These activities, to be completed during Phase 1, include:
1. Completion of modifications to System 52 (Turbine Steam
System) , including physical construction, testing and
auditing.
2. Completion of modifications to System 91 (Hydraulic Piping
System) , including physical construction, testing, and
auditing.
3. Completion of all miscellaneous construction, testing and
auditing items except on-going maintenance items and tasks
to be accomplished during rise to power and during power
operation.
During Phase 1 operations:
1. The cold shutdown margin for the core with the highest worth
operable rod stuck in the withdrawn position shall be main-
tained at a minimum value of 0.025Ap.
2. All electrical power shall be disconnected from the control
rod drives not in use during testing.
b) Phase 2 - Hot physics tests with helium environment and rise to
full power testing.
Prior to initiation of Phase 2 activities, all system modifications,
testing and documentation, including that identified in Phase 1
above, must be completed and approved by the AEC's Regional
Regulatory Operations Office.
4.9-3
Following completion of Phase 1 activities, and the appropriate
approval by the AEC's Regional Regulatory Operations Office,
Specification LCO 4.9.1, shall not restrict operation of this
facility.
Basis for Specification LCO 4.9.1
The following of the two-phase approach identified ensures that all
system modifications, testing, and documentation necessary to protect the
health and safety of the public are completed in an orderly and timely
manner.
5.0-1
5.0 SURVEILLANCE REQUIREMENTS
The surveillance requirements specified in this section define the
tests, calibrations, and inspections which are necessary to verify the
performance and operability of equipment essential to safety during all
modes of operation, or required to prevent or mitigate the consequences
of abnormal situations.
5.1-1
5.1 REACTOR CORE AND REACTIVITY CONTROL - SURVEILLANCE REQUIREMENTS
Applicability
Applies to the surveillance of the reactor core and core reactivity
control mechanisms.
Oblective
To ensure the capability to control the reactivity and temperature
of the reactor core.
Specification SR 5.1.1 - Control Rod Drives Surveillance
The surveillance of the control rod drives shall be as follows :
a) All 37 control rod pairs will be scrammed from the full out to
the full in position once a year and the scram time measured.
Operable withdrawn control rods shall have a scram time less
than 160 seconds.
b) All control rods which are withdrawn during power operation will
be exercised a short distance (about 6 inches) once a month.
Operation of position indicators, motion indicators, and the
absence of slack cable indication shall be verified.
Basis for Specification SR 5.1.1
Tests will be performed on the control rod drives to assess their
capability to control the reactivity of the reactor core. On a yearly
basis, the control rods will be scrammed from the full out position and
the scram time measured. The drive mechanisms are designed for a normal
scram time of 140 ± 10 seconds. However, for safe reactivity control of
the reactor, scram times of the drive mechanisms may be as great as
160 seconds without altering the kinetics of the scram.
5.1-2
The drive mechanism will be used to exercise sequentially, all
withdrawn rods over a short distance (about 6 inches) once a month.
This test will assess the operability of the control rods and drives
and position indicating instrumentation. Any binding of the rods in
their channels can be determined by a slack cable indication.
Specification SR 5.1.2 - Reserve Shutdown System Surveillance
The surveillance of the reserve shutdown system shall be as follows:
a) The ability to pressurize each of the 37 reserve shutdown hoppers
to 10 psi above reactor pressure, as indicated by operation of
the hopper pressure switch, shall be demonstrated every three
months. Operable reserve shutdown hoppers shall be capable
of being pressurized.
b) The test pressurizing gas pressure indicator shall be calibrated
annually.
c) An off-line functional test of a reserve shutdown assembly shall
be performed in the hot service facility, or other suitable
facility, following each of the first five refueling cycles and
at two refueling cycle intervals thereafter. These tests will
consist of pressurizing reserve shutdown hopper to the point
of rupturing the disc and releasing the poison material. If a
reserve shutdown hopper rupture disc does not rupture at a
differential pressure less than 300 psi and release the poison
material, the reactor shall be placed in a shutdown condition
until it can be shown that LCO 4.1.6 can be met.
d) The instrumentation which alarms a low pressure in the reserve
shutdown actuating pressure lines shall be functionally tested in conjunction with, and at the same intervals specified in part a)
5.1-3
above, and calibrated once a year. Operable reserve shutdown
hoppers shall have an actuating bottle pressure > 1500 psig.
e) The reserve shutdown hopper pressure switches shall be calibrated
at the same interval that they are removed from the reactor for
maintenance.
Basis for Specification SR 5.1.2
The reliability of the reserve shutdown system to perform its function
will be maintained by a control system pressure test and actual off-line
rupture tests conducted in the hot service facility or other suitable facility.
The control system pressure test demonstrates the ability to pressurize
the hoppers and indicates the operability of the control system components.
A successful test will increase the hopper pressure about 10 psi above
reactor pressure. This differential is well below the minimum 150 psi
differential required to burst the disc.
The off-line tests consist of actual disc ruptures and poison drops.
These will be used to determine the reliability of the differential burst
pressure of the disc, and the tendency of the poison material to hang up
or deteriorate in the hoppers over extended periods of time.
This test information will be used to verify the capability to shut
down the reactor in an emergency situation. The reserve shutdown system
hoppers operate in two subsystems. The first consists of the seven hoppers
in refueling regions 1, 3, 5, 7, 22, 28 and 34; the second subsystem is
comprised of the remaining thirty hoppers in the remaining refueling regions.
Safe control of the reactor by the reserve shutdown system can be
accomplished with one of the seven hoppers inoperative, and one of the
remaining 30 hoppers inoperative. A differential pressure of from 585 to 315 psi
is available from the helium supply bottle with a pressure > 1500 prig.
5.1-4
Specification SR 5.1.3 - Temperature Coefficient Surveillance
The reactivity change as a function of core temperature change shall
be measured at the beginning of each refueling cycle.
Basis for Specification SR 5.1.3
The major shifts in reactivity change as a function of core temperature
change will occur following refueling. The specified frequency of
measurement following each major refueling will assure that the change of
reactivity as a function of changes in core temperature will be measured
on a timely basis to evaluate the limit specified in Specification LCO 4.1.5.
Specification SR 5.1.4 - Reactivity Status Surveillance
A surveillance check of the reactivity status of the core shall be
performed at each startup and once per week during power operation. If
the difference between the observed and the periodically renormalized
expected reactivities at steady state conditions reaches 0.012 Ok, this
discrepancy shall be considered an abnormal occurrence.
Basis for Specification SR 5.1.4
The specified frequency of the surveillance check of the core
reactivity status will assure that the difference between the observed
and expected core reactivity will be evaluated regularly.
This specification is designed to ensure that the core reactivity
level is monitored to reveal in a timely manner the existence of potential
safety problems or operational problems. An unexpected and/or unexplained
change in the observed core reactivity could be indicatite of such problems.
The periodic renormalization of the expected reactivity will eliminate
discrepancies due to manufacturing tolerances, analytical modeling
approximations, and deficiencies in basic data.
5.1-5
The subsequent comparison of predicted and observed reactivities
in a "basic configuration" (i.e. , at a constant rod configuration, at
full power, and with equilibrated xenon and samarium) , will ensure that
the comparison will be easily understood and readily evaluated. The
value of 0.012 AK is considered to be a safe limit since the reactivity
insertion of this amount has been studied in accident analysis and
been found to result in changes in system temperatures which is not excessive.
Specification SR 5.1.5 - Withdrawn Rod Reactivity Surveillance
The reactivity worth of the control rods which are withdrawn from
the low power condition to the operating condition, in the normal withdrawal
sequence, shall be measured at the beginning of each refueling cycle.
The measured rod worths will be used to insure that the criteria for the
selection of the rod sequence of Specification LCO 4.1.3 are met.
Basis for Specification SR 5.1.5
The measurement of control rod worths at the beginning of a refueling
cycle will provide for an evaluation of calculational methods for control
rod worths used in the prediction of the maximum worth rod in Specification
LCO 4.1.3.
Specification SR 5.1.6 - Core Safety Limit Surveillance
During power operation the total operating time of the fuel elements
within the core at power-to-flow ratios above the curve of Figure 3.1-2
will be evaluated once per week when the plant operation is within the
normal operating range, and as soon as practicable after any deviation
from the normal operating range. These operating times will be compared
to the allowable operating time of Specification SL 3.1 to assure that the
Core Safety Limit has not been exceeded.
5.1-6
Basis for Specification SR 5.1.6
Only during operation of the plant outside of the normal operating
range is there a potential for accumulating significant operating times
at power-to-flow ratios greater than the curve of Figure 3.1-2. Therefore,
weekly evaluations of the total accumulated operating time at power-to-flow
ratios greater than the curve of Figure 3.1-2 is sufficient during normal
operation. Following any significant deviation from the normal operating
range, the operation should be evaluated to determine the degree to which
the actual total operation of the core approached the Core Safety Limit.
5.2-1
5.2 PRIMARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS
Applicability
Applies to the surveillance of the primary (helium) reactor
coolant system excluding the steam generators.
Obi ect ive
To ensure the capability of the components of the primary reactor
coolant system to maintain the primary reactor coolant envelope as a
fission product barrier and to ensure the capability to cool the core
under all modes of operation.
Specification SR 5.2.1 - PCRV Overpressure Safety System Surveillance
(a) One of the two rupture disc and safety valve assemblies in
the PCRV overpressure safety system shall be tested each year
at shutdown on an alternate basis after initial power operation.
(b) Each safety valve protecting a steam generator or circulator
penetration shall be tested at intervals not to exceed every
five years.
(c) The instrumentation associated with the valves in (a) and (b)
above shall be tested and calibrated as follows:
1) The pressure switches and alarms for the interspace
between the rupture discs and the relief valves shall
be functionally tested monthly and calibrated annually.
2) The position indicating lights associated with the PCRV
relief valve shutoff valves shall be functionally tested
and calibrated annually.
5.2-2
Basis for Specification SR 5.2.1
The rupture disc installation can be tested when the reactor is
shut down and the primary system pressure has been reduced below 50 psig.
This allows closure of the manual block valve upstream of one of the safety
valve-rupture disc trains. The rupture disc assembly is then removed and
bench tested. The test will involve a bench test determination of the
pressure level at which the deflection of the rupture disc would cause
rupture to occur. The second safety valve-rupture disc system remains in
a fully operable condition during this testing procedure thus ensuring
pressure relief protection for the PCRV.
The safety valves are tested for setpoint activation without removing
them from the system. The reactor is shut down, and the primary system
pressure is reduced below 50 psig.
The steam generator and helium circulator penetrations are provided
with safety valves to prevent overpressure should a process line rupture
within a penetration. This testing of these valves at five year intervals
provides information on the capability of these valves to relieve under
design conditions.
The intervals specified for testing the associated instrumentation is
adequate to assure the reliability of the rupture disc and relief valve
operation.
5.2-3
Specification SR 5.2.2 - Tendon Corrosion Surveillance
The serviceability of the corrosion protection applied to and the
condition of the prestressing tendons shall be monitored as follows:
a) Corrosion-protected wire samples of sufficient length (at least
15 feet) shall be inserted with selected tendons (those tendons with load
cells). Corrosion inspection of at least one of these wires shall be
made after the end of the first and third calendar year after prestressing.
Additional inspections shall be conducted at five calendar year intervals
thereafter.
b) A sample of the atmosphere contained in a representative number
of tendon tubes (tendon tubes without load cells and tendon tubes with
load cells from which wire samples are examined) shall be drawn and
analyzed for products of corrosion at the end of the first and third
calendar year after prestressing. Additional samples shall be taken at
five calendar year intervals thereafter.
Basis for Specification SR 5.2.2
The corrosion protection provided for the PCRV prestressing components
is considered to be more than adequate to assure that the required
prestressing forces are sustained throughout the operational life of
the plant. The details of the corrosion protection system are described
in Section 5.6.2.5 of the FSAR.
Sampling tendon tube atmosphere for products will provide a secondary
check on the adequacy of the corrosion protection provided for the stressing
tendons.
5.2-4
Specification SR 5.2.3 - Tendon Load Cell Surveillance
Checks on the possible shift in the load cell reference points
for representative load cells shall be performed after the end of the
first and third calendar year after initial prestressing. Additional
checking shall be conducted at five calendar year intervals thereafter.
Basis for Specification SR 5.2.3
The PCRV tendons apply the force required to counteract the
internal pressure. Therefore, they are the PCRV structural components most
capable of being directly monitored and of indicating the capability of
the vessel to resist internal pressures. Since the relation between
effective prestress and internal pressure is directly and easily calculable,
monitoring tendon loads is a direct and reliable means for assuring that
the vessel always has capacity to resist pressures up to Reference Pressure.
Monitoring of the tendon loads will assure that deterioration of
structural components including progressive tendon corrosion, concrete
strength reduction, excessive steel relaxation, etc. , cannot occur undetected
to a degree that would Jeopardize the safety of the vessel. Each of these
phenomena would result in tendon load changes. These changes, as reflected
by the load cells, are monitored in the control room by an alarm system
which alerts the operator when the tendon load settings are exceeded. The
upper settings will be varied depending on the location of the tendon being
monitored, while the lower settings for all load cells will be set to
correspond to 1.25 times peak working pressure (PWP) .
Specification SR 5.2.4 - PCRV Concrete Crack Surveillance
Crack patterns on the visible surfaces of the PCRV shall be mapped
prior to and following the initial proof test pressure (IPTP). Concrete
cracks which exceed 0.015 inches in width shall be recorded. Subsequent
5.2-5
concrete surface visual inspections shall be performed after the end
of the third and fifth calendar year following initial prestressing.
Recorded cracks shall be assessed for changes in length and any new
cracks will be recorded. Additional inspections shall be conducted
at ten calendar year intervals thereafter.
Basis for Specification SR 5.2.4
Cracks are expected to occur in the PCRV concrete resulting from
shrinkage, thermal gradients, and local tensile strains due to mechanical
loadings. The degree of cracking expected is limited to superficial
effects and is not considered detrimental to the structural integrity
of the PCRV. Reinforcing steel is provided to control crack growth
development with respect to size and spacing. Model testing has also
shown that severely cracked vessels contain the normal working pressure
for extended periods of time as long as the effective prestressing forces
are maintained.
Cracks up to about 0.015 inches (limits of paragraph 1508b, ACI 318-63)
for concrete not exposed to weather are generally considered acceptable and
corrosion of rebars at such cracks is of negligible consequence. Large
crack widths will require further assessment as to their significance,
depending on the width, depth, length, and location of the crack on the
structure, and must be considered with reference to the observed overall
PCRV response.
Further discussion on the significance of concrete cracks in the PCRV
is given in Section 5.12.5 of the FSAR.
Observed crack development with time during reactor operation will be
related to the PCRV structural response as monitored by the installed sensors
and deflection measurements. Details of the PCRV structural monitoring
5.2-6
provisions are given in Section 5.13.4 and Appendix E.17 of the FSAR.
The interval for surveillance after the fifth year following
initial prestressing may be adjusted based on the analysis of prior results.
Specification SR 5.2.5 - Liner Specimen Surveillance
Specimens shall be placed adjacent to the outside surface of the
top head liner so that changes in notch toughness due to irradiation of
the steel can be measured during the life of the reactor.
Five years following initial power operation, three sets of 12 specimens
of the PCRV liner materials and weld material shall be removed and tested
to obtain Charpy impact data. The specimen holders shall contain dosimeters
to provide integrated neutron flux measurements. Additional specimen
removal and testing shall be conducted at ten year intervals thereafter.
Basis for Specification SR 5.2.5
A test program will be performed to survey and assess the shifts in
NDTT of the PCRV liner materials. The testing is to be accomplished by
placing Charpy impact test specimens, made from the liner materials, near
the liner and exposing them to appropriate neutron fluxes and temperatures.
The Charpy impact test specimens are to be removed, 36 at a time, during
the life of the vessel and tested to determine the condition of the
vessel steel. The total number of specimens placed in the reactor is '750;
which will allow the determination of a complete impact transition curve
for the plate metal, the weld metal and the heat affected zone at each
test interval.
This testing program will meet the requirements of ASTM-E-185-70, with
the following exceptions:
5.2-7
a. Tensile specimens are not included, since the liner is not
a load carrying member but only a ductile membrane.
b. No thermal control specimens have been provided, since
there is no appreciable temperature cycling of the liner.
The liner materials will normally be kept at or below 150°F
during all plant operation.
Tests performed on this liner material (see FSAR Section 5.7.2.2)
have indicated that no observable changes in material characteristics
developed during an exposure to a fluence equivalent to the first five
years of power operation. Further, these tests demonstrated no significant
damage after a fluence equivalent to 30 years of power operation. The
testing program prescribed for the Fort St. Vrain liner is in compliance
with the ASME Boiler and Pressure Vessel Code, Section III N-110.
The interval for specimen removal and testing subsequent to the fifth
refueling cycle may be adjusted based on the analysis of prior results.
Specification SR 5.2.6 - Plateout Probe Surveillance
One plateout probe shall be removed for evaluation coincident with the
first, third, and fifth refueling, and at intervals not to exceed five
refueling cycles thereafter. If, during the second or fourth refueling
cycle, or any refueling cycle following the fifth refueling, the primary
coolant noble gas activity (gamma + beta) should increase by 25% over
the average activity of the previous three months at the same reactor
power level and the primary coolant activity is greater than 25% of design,
the plateout probe shall be removed at the end of that refueling cycle.
The probes shall be analyzed for "Sr inventory in the reactor circuit.
The probes removed shall also be analyzed for 1311
5.2-8
Basis for Specification SR 5.2.6
The plateout probes are located in penetrations extending into
steam generator shrouds and then into the gas stream of each coolant
loop. One sample is accumulated by continuously bypassing a small
portion of the core outlet coolant stream through diffusion tubes and
sorption beds located in the probe body. Another sample can be accumulated
by continuously bypassing a portion of the circulator outlet coolant stream
through the probe. The core outlet sample can be used to determine the
concentrations of fission products in the coolant stream entering the
steam generator; the circulator outlet sample provides information about
the amount of cleanup in each pass around the circuit.
The probes shall be analyzed for 90Sr and the results shall be used
to establish the total 90Sr inventory in the reactor circuit to determine
compliance with LCO 4.2.8. Results of probe analyses shall be compared
with the calculated estimates of 90Sr which were made between probe
removals. The analysis for 1311 shall be made to determine the degree
of conservatism of the assumptions made regarding the circulating and
plated out iodine in the primary coolant circuit.
The interval for probe removal and analysis subsequent to the fifth
refueling cycle may be adjusted based upon the analysis of prior results.
Specification SR 5.2.7 - Water Turbine Drive Surveillance
Components of the helium circulator water turbine drive system shall
be tested as follows:
a) One circulator and the associated water supply valving in each
loop will be functionally tested by operation on water turbine
drive using feedwater, condensate, and condensate at reduced _,
5.2-9
pressure to simulate fire pump discharge pressure as motive
power, annually.
b) Safety valves (V-21522, V-21523, V-21542, and V-21543) ,
located in the water turbine supply lines , will be tested
for relieving pressure annually.
c) Both turbine water removal pumps and the turbine water
removal tank overflow to the reactor building sump shall be
functionally tested every three months.
d) The instrumentation and controls associated with c) shall
be functionally tested in conjunction with and at the same
intervals as the turbine water removal pumps and shall be
calibrated annually.
Basis for Specification SR 5.2.7
The circulator water turbine drives are normally operated during
an extended shutdown. Therefore the specified surveillance requirements
are adequate to ensure water turbine operability.
Specification SR 5.2.8 - Bearing Water Makeup Pump Surveillance
The circulator bearing water makeup pumps and associated instruments
and controls shall be tested as follows:
a) Normal Makeup Pump shall be operated in the recycle mode
every three months.
b) Emergency Makeup Pump shall be functionally tested every
three months.
c) The associated instruments and controls shall be functionally
tested in conjunction with and at the intervals specified in
parts a) and b) above, and calibrated annually.
5.2-10
Basis for Specification SR 5.2.8
During accident conditions described in FSAR Section 10.3.9, the
circulator bearing water makeup pump is required to operate inter-
mittently to make up bearing water. The specified testing interval
is sufficient to ensure proper operation of the pumps and associated
controls.
Specification SR 5.2.9 - He Circulator Bearing Water Accumulators
The helium circulator bearing water accumulators, instrumentation,
and controls shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.2.9
He Circulator bearing water is normally supplied from the bearing
water system and is backed up by the backup bearing water system
supplied from the Emergency Feedwater Header. In the event of a failure
in both of these systems, the water stored in the bearing water accumulators
is adequate to safely shut down both helium circulators in a loop. The
monthly test interval and annual calibration interval will assure proper
operation of the accumulator controls if they should ever be called upon
to function.
Specification SR 5.2.10 - Engine-driven Fire Pump Surveillance
The engine-driven fire pump shall be functionally tested once a week.
The associated instruments and controls shall be functionally tested
monthly and calibrated annually.
Basis for Specification SR 5.2.10
During accident conditions described in FSAR Section 10.3.9, one
of the fire pumps is required to operate. The specified testing interval
5.2-11
is sufficient to ensure proper operation of the pump and associated control.
The motor driven fire pump routinely operates intermittently.
Specification SR 5.2.11 - Primary Reactor Coolant Radioactivity Surveillance
A grab sample of primary coolant shall be analyzed a minimum of once
per week during reactor operation for its radioactive constituents and
shall be used to calibrate the continuous primary coolant activity monitor.
If the continuous primary coolant activity monitors is inoperable, the
primary coolant activity level reaches 25% of the limits of LCO 4.2.8, or
the primary coolant acitivity level increases by a factor of 25% over the
previous equilibrium value of the same reactor power level, the frequency of
sampling and analysis shall be increased to a minimum of once each day until
the activity level decreases or reaches a new equilibrium value (defined by
four consecutive daily analysis whose results are within ± 10%) at which time
weekly sampling may be resumed.
Basis for Specification SR 5.2.11
The design of the instrumentation is such that under normal operating
conditions the activity of the primary coolant is measured and indicated
on a continuous basis. The weekly sampling interval provides an adequate
check on the continuous monitoring equipment.
Specification SR 5.2.12 - Primary Reactor Coolant Chemical Surveillance
The primary coolant shall be analyzed for chemical constituents a
minimum of once per week. If the chemical impurity levels exceed 50 percent
of the limits of LCO 4.2.10 or LCO 4.2.11, whichever is applicable, the
frequency of sampling and analysis shall be increased to a minimum of once
each day until the level decreases or reaches a new equilibrium value
(defined by four consecutive daily analysis whose results are within 10%) ,
at which time weekly sampling may be resumed.
5.2-12
Basis for Specification 5.2.12
The chemical constituents in the primary coolant are routinely measured
on a continuous basis. The specification of an interval for surveillance
allows for routine maintenance of the chemical impurity monitoring
equipment. The presence of higher than nominal impurity levels of
chemical impurities is related to core materials corrosion which might
occur only with very high levels for sustained periods of time.
Specification SR 5.2.13 - PCRV Concrete Helium Permeability Surveillance
The permeability of the PCRV concrete to helium shall be measured
prior to the initial startup of the reactor and after the end of the
third year following initial power operation. Additional measurements
shall be made at five year intervals thereafter.
Basis for Specification SR 5.2.13
Measurements of the relative helium permeability throughout plant
life provides, as a supplement to other surveillance efforts, information
concerning the continued integrity of the PCRV concrete.
The interval for surveillance after the fifth year following the
initial power operation may be adjusted based on the analysis of prior
results.
Specification SR 5.2.14 - PCRV Liner Corrosion Surveillance Requirement
The PCRV liner shall be examined for corrosion induced thinning, using
ultrasonic inspection techniques at the end of the third and fifth years
following initial power operation. Additional examinations shall be
conducted at ten year intervals thereafter.
Basis for Specification SR 5.2.14
The ultrasonic inspection of the PCRV liner is provided to detect
the thinning of the liner due to corrosion or to detect defects within
the liner at representative areas. Although no corrosion is expected
to occur, this specification allows for detection of corrosion or liner
5.2-13
defects in the event of some unexpected and unpredicted changes in the
liner characteristics. The provisions are discussed in Section 5.13
of the FSAR.
The interval for surveillance after the fifth year following initial
power operation may be adjusted based on the analysis of prior results.
Specification SR 5.2.15 - PCRV Penetration interspace Pressure Surveillance
The instrumentation which monitors the pressure differential between
the purified helium supply header to the PCRV penetration interspaces and
the primary coolant system will be functionally tested once every month
and calibrated annually.
Basis for Specification SR 5.2.15
This calibration and test frequency is adequate to insure that the
purified helium being supplied to the PCRV penetration interspaces
shall be at a higher pressure than the primary coolant pressure within
the PCRV.
Specification SR 5.2.16 - PCRV Closure Leakage, Surveillance Requirements
The surveillance of PCRV closure leakage shall be as follows:
a) PCRV primary and secondary closure leakage shall be
determined once per month, or as soon as practicable
after an increase in pressurization gas flow is alarmed.
b) The instrumentation monitoring PCRV penetration closure
interspace pressurization gas flows, including alarms and
high flow isolation, shall be functionally tested monthly
and calibrated annually.
5.2-14
Basis for Specification SE 5.2.16
The interval specified for determining the actual primary and
secondary closure leakage is adequate to assure compliance with
LC0 4.2.9. •
In the determination of closure leakage at the reference differential
pressure, laminar leakage flow shall be conservatively assumed,
therefore in correcting the determined closure leakage to reference
differential pressure, the ratio of the reference differential pressure,
and test differential pressure shall be used.
The interval specified for functional testing and calibration of
the instrumentation and alarms monitoring the penetration closure
interspace pressurization gas flow will assure sensing and alarming
any change in pressurization gas flow.
5.3-1
5.3 SECONDARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS
Applicability
Applies to the surveillance of the secondary (steam) coolant system
including the steam generators and turbine plant.
Objective
To ensure the core cooling capability of the components of the
steam plant system.
Specification SR 5.3.1 - Steam/Water Dump System Valves, Surveillance
Requirements
The steam/water dump valves shall be tested individually every
three months.
The steam/water dump tank level indicators shall be checked daily,
functionally tested monthly, and calibrated at each refueling.
Basis for Specification SR 5.3.1
The steam/water dump system is provided to minimize water
inleakage into the core as a result of a steam generator tube rupture
(FSAR Section 6.3) . Satisfactory operation of the dump valves as is
sufficiently demonstrated by testing every three months, will minimize
core damage and primary coolant system pressure rise in the event of a
steam generator tube rupture.
The dump valve test will be accomplished by closing the (normally
locked open) block valve downstream of the dump valve to be tested.
After operation of the dump valve, the block valve will again be locked
open, returning the dump valve to service.
5.3-2
Specification SR 5.3.2 - Main and Hot Reheat Steam Stop Check Valves,
Surveillance Requirements
The main steam and hot reheat steam stop check valves shall be full
stroke tested once per year and partial stroke tested once per week.
Basis for Specification SR 5.3.2
The main steam stop check and hot reheat stop check valves will be
partially stroked once a week during plant operation. Full stroking
tests are impractical because complete closure of any one valve would
automatically shut down one or more circulators. Therefore, the valves
will be stroked during power operation by means of special electrical
circuitry in the hydraulic control system which limits closure to ten
percent without interfering with emergency closure action called for by
the plant protective system. This test will demonstrate that the valves
are free to close when required, without causing severe pressure,
temperature, flow, or power generation transients.
Specification SR 5.3.3 - Bypass and Safety Valves, Surveillance Requirements
The main steam and hot reheat steam electromatic valves, the main steam
bypass valves, and the six hot reheat steam bypass valves shall be tested
once per year.
Basis for Specification SR 5.3.3
The specified secondary (steam) coolant system bypass valves and
safety valves will be tested once per year during plant shutdown as follows :
a) The main steam and hot reheat steam electromatic valves will be
tested by exercising the relief.
b) The main steam bypass valves will be tested for operability by
cycling the valves.
c) The six hot reheat steam bypass valves will be tested by exercising
each valve to ensure freedom of movement.
5.3-3
The main steam bypass valves divert up to 77% steam flow (via
desuperheaters) to the bypass flash tank on turbine trip or loop
isolation, so that the steam is available for driving helium circulators ,
boiler feed pump turbines, etc. The main steam electromatic valves
divert the remaining steam flow to atmosphere.
The six hot reheat steam bypass and electromatic relief valves
ensure a continuous steam flow path from the helium circulators for
decay heat removal.
The tests required on the above valves will demonstrate that each
valve will function properly. Test frequency is considered adequate for
assuring valve operability at all times.
Specification SR 5.3.4 - Safe Shutdown Cooling Valves, Surveillance
Those valves that are pneumatically, hydraulically, or electrically
operated, that are required for actuation of the Safe Shutdown Cooling mode
of operation, shall be tested twice annually with the interval between tests
to be not less than four (4) nor greater than eight (8) months.
Basis for Specification SR 5.3.4
The safe shutdown cooling mode of operation utilizes systems or
portions of systems that are in use during normal plant operation. In many
cases, those valves required to initiate Safe Shutdown Cooling are not
called upon to function during normal operation of the plant except to
stand fully closed or open.
Testing of these valves by stroking them twice annually will assure
their operation if called upon to initiate the Safe Shutdown Cooling mode
of operation.
5.3-4
During reactor operation, the instrumentation required to monitor and control the Safe-Shutdown mode of cooling is normally in use and
any malfunction would be immediately brought to the attention of the
operator. That instrumentation not normally in use is tested at
intervals specified by other surveillance requirements in this Technical
Specification.
Safe Shutdown Cooling, the systems or portions of systems involved,
are discussed in Sections 10.3.9 and 10.3.10 of the FSAR and are
represented in FSAR Figure 10.3-4.
Specification SR 5.3.5 - Hydraulic Power System Surveillance Requirements
The pressure indicators and low pressure alarms on the hydraulic oil
accumulators pressurizing gas and on the hydraulic power supply lines shall
be functionally tested once every three months and calibrated once per year.
Basis for Specification SR 5.3.5
The hydraulic power system is a normally operating system. Malfunctions
in this system will normally be detected by failure of the hydraulic oil
pumps or hydraulic oil accumulators to maintain a supply of hydraulic oil
at or above 2500 psig. Functional tests and calibrations of the pressure
indicators and low pressure alarms on the above basis will assure the
actuation of these alarms upon a malfunction of the hydraulic power system
which may compromise the capability of operating critical valves.
Specification SR 5.3.6 - Instrument Air System - Surveillance Requirements
The pressure indicators and low pressure alarms on the instrument air
receiver tanks and headers shall be functionally tested monthly and
calibrated annually.
5.3-5
1 Basis for Specification SR 5.3.6
The instrument air system is a normally operating system. Malfunctions
in this system will be normally detected by failure of the instrument air
compressors to maintain the instrument air receiver tanks at a pressure
above the alarm setpoint. Functional tests of the pressure indicators
and low pressure alarms on a monthly basis and calibration on an annual basis
will assure the actuation of these alarms upon a malfunction of the
instrument air system which may compromise the capability of operating
critical values. •
Specification SR 5.3.7 - Secondary Coolant Activity, Surveillance Requirements
The secondary coolant system will be analyzed for 131I, tritium, and
gross beta plus gamma concentration once per week during reactor operation.
If the secondary coolant activity level reaches 25% of the limit of
LCO 4.3.8, or the secondary coolant activity level increases by a factor
of 25% over the previous equilibrium value at the same reactor power level,
the frequency of sampling and analysis shall be increased to a minimum of
once each day until the activity level decreases or reaches a new equilibrium
value (defined by four consecutive daily analysis whose results are within
±10%) , at which time weekly sampling may be resumed.
Basis for Specification SR 5.3.7
The specification surveillance interval is adequate to monitor the
activity of the secondary coolant.
l .
5.4-1
5.4 INSTRUMENTATION AND CONTROL SYSTEMS - SURVEILLANCE AND CALIBRATION
REQUIREMENTS
Applicability
Applies to the surveillance and calibration of the reactor protective
system and other critical instrumentation and controls.
Objective
To assure the operability of the reactor protection system and other
critical instrumentation and controls by specifying their surveillance
and calibration frequencies.
Specification SR 5.4.1 - Reactor Protective System and Other Critical
Instrumentation and Control Checks, Calibrations, and Tests
The surveillance and calibration tests of the protective instrumentation
shall be as given in Tables 5.4.1 through 5.4.4:
a) Table 5.4.1 - Minimum Frequencies for checks, calibrations,
and testing of scram system.
b) Table 5.4.2 - Minimum Frequencies for checks, calibrations,
and testing of Loop Shutdown System.
c) Table 5.4.3 - Minimum Frequencies for checks, calibrations,
and testing of Circulator Trip System.
d) Table 5.4.4 - Minimum Frequencies for checks, calibrations, and
testing of Rod Withdrawal Prohibit System.
Basis for Specification SR 5.4.1
The specified surveillance check and test minimum frequencies are
based on established industry practice and operating experience at
conventional and nuclear power plants. The testing is in accordance
with the IEEE Criteria for Nuclear Power Plant Protection Systems, and
in accordance with accepted industry standards.
5.4-2
Calibration frequency of the instrument channels listed in Tables
5.4.1, 5.4.2, 5.4.3, 5.4.4 are divided into three categories: passive
type indicating devices that can be compared with like units on a continuous
basis; semiconductor devices and detectors that may drift or lose
sensitivity; and on-off sensors which must be tripped by an external source
to determine their setpoint. Drift tests by GGA on transducers similar to
the reactor pressure transducers (FSAR Section 1.3.3.2) indicate insignificant
long term drift. Therefore a once per refueling cycle calibration was
selected for passive devices (thermo-couples, pressure transducers, etc. ).
Devices incorporating semiconductors, particularly amplifiers, will be
also calibrated on a once per refueling cycle basis, and any drift in
response or bistable setpoint will be discovered from the test program.
Drift of electronic apparatus is not the only consideration in determining
a calibration frequency; for example, the change in power distribution
and loss of detector chamber sensitivity require that the nuclear power
range system be calibrated every month. On-off sensors are calibrated
and tested on a once per refueling cycle basis.
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5.4-12
Specification SR 5.4.2 - Control Room Smoke Detector
The control room smoke detectors and alarms will be functionally tested
once per year.
Basis for Specification SR 5.4.2
The control roam smoke detectors provide for sensing of the smoke in
the outlet air ducts from both the control room and the auxiliary electrical
room. In the event of any fire or smoke in the control panels, alarms
will be initiated.
Specification SR 5.4.3 - Core Region Outlet Temperature Instrumentation
The output of two thermocouples measuring each region outlet temperature
will be checked daily during power operation. If the indicated temperatures
for a region differ by > ± 75°v, a calibration shall be made and the faulty
thermocouple replaced by an operable thermocouple. The core region outlet
thermocouple shall be calibrated once per year during power operation by
traversing a calibrated thermocouple along each of the seven coolant
thermocouple assemblies.
Basis for Specification 5.4.3
The long-term thermocouple drift is estimated to be < 15°F per year
and this drift was included in the measurement uncertainty of ± 50°F used
to establish LCO 4.1.7. With this measurement uncertainty, a root mean
square difference of > ± 75°F would be an indication of a faulty reading.
Daily checks and yearly calibrations are considered adequate since the
expected drift in calibration is small and has been included in establishing
LCO 4.1.7 (See FSAR Section 7.3.3).
5.4-13
Specification SR 5.4.4 - PCRV Cooling Water System Temperature Scanner-
Surveillance Requirement
PCRV Cooling System temperature scanner readings shall be checked by
comparison of representative liner cooling tube thermocouple outputs to
their respective subheader temperatures and associated alarms tested once
per month during power operation.
All thirty-six (36) outlet subheader temperature indicators shall be
calibrated annually. In addition, ninety-seven (97) liner cooling tube
outlet thermocouples shall be calibrated annually.
Basis for Specification SR 5.4.4
The temperature scanner for the PCRV cooling system provides for
continuous temperature monitoring of the outlet water temperature of each
individual liner cooling tube and alarming of high outlet temperatures.
The surveillance interval specified is sufficient to detect any drift
in the output of the individual thermocouples or scanner electronics to assure
the temperature limitations of the PCRV cooling system are not exceeded.
The ninety-seven (97) thermocouples shall be distributed among the
thirty-six (36) subheaders so that between 16.7% and 21.5% of the total in each
subheader are calibrated each year. Thus, the maximum time between calibration
of any one thermocouple, or any complete subheader, shall not exceed six (6)
years. The overall percentage of thermocouples calibrated per year exceeds 18%.
The surveillance interval for calibration, combined with that for
checking, assures sufficient accuracy of temperature measurement to adequately
protect the PCRV concrete.
Specification SR 5.4.5 --PCRV Cooling Water System Flow Scanner -
Surveillance Requirement
A PCRV Cooling System flow scanner readout shall be taken and alarms
5.4-14
functionally checked monthly. The scanner and alarms, and six (6)
subheader flow meters shall be calibrated annually.
Basis for Specification SR 5.4.5
The flow scanner acts as a backup to the temperature scanner and
initiates no automatic protective action, only an alarm. Because a
restriction or a leak in the system would develop over a period of time,
the monthly interval for comparing scanner readouts is sufficient to detect
any long term change in the system.
Specification SR 5.4.6 - Core AP Indicator - Surveillance Requirement
The core AP instrumentation shall be calibrated on a once per refueling
cycle interval.
Basis for Specification 23 5.4.6
Core differential pressure is an indication of gross blockage of
flow in the core.
Specification SR 5.4.7 - Control Room Temperature-Surveillance Requirement
The control room temperature control thermostat shall be functionally
tested monthly and calibrated annually.
Basis for Specificatiol SR 5.4.7
The surveillance interval specified for functional testing and
calibration of the control room thermostat will assure its ability to
not only control the room temperature as desired, but to also indicate
the correct room temperature within the accuracy of the instrument.
Specification SR 5.4.8 - Power to Flow Instrumentation - Surveillance Requirement
The power to flow indication shall be verified daily and shall be
calibrated once per refueling cycle.
5.4-15
Basis for Specification SR 5.4.8
The power to flow ratio indication is an indication of the balance
between the heat generation and removal within the primary coolant system.
A verification of the power to flow indication on a daily basis is adequate
to assure the instrument is indicating properly. In addition, any change
in reactor power level no matter how small, should produce a change in the
power to flow ratio indication. A lack of response by this instrumentation
would be noticed by the operator. Calibration of the instrumentation
on a once per refueling cycle basis is acceptable by industry standards
for this type of instrumentation.
Specification SR 5.4.9 - Area and Miscellaneous Process Radiation Monitors-
Surveillance Requirement
The area radiation monitors shall be functionally checked weekly and
calibrated annually.
Basis for Specification SR 5.4.9
The surveillance interval specified for functional testing and calibration
are adequate to assure the proper operation of these detectors.
Specification SR 5.4.10 - Seismic Instrumentation - Surveillance Requirement
The Seismic Instrumentation shall be functionally tested every six
months and calibrated every two years.
Basis for Specification 5.4.10
The intervals specified for testing and calibration of the Seismic
Instrumentation are recommended by the manufacturer to assure the
instruments operate as intended.
5.4-16
Specification SR 5.4.11 - PCRV Surface Temperature Indication -
Surveillance Requirement
The PCRV surface temperature indicators shall be functionally tested
monthly and calibrated annually.
Basis for Specification SR 5.4.11
The PCRV surface temperature indicators provide for continuous
monitoring of surface concrete temperatures to assure the proper temperature
gradient is maintained through the PCRV wall and heads.
The surveillance interval specified is adequate to detect any drift
or malfunction of this instrumentation.
5.5-1
5.5 CONFINEMENT SYSTEM - SURVEILLANCE REQUIREMENTS
Applicability
Applies to the surveillance of the reactor building (confinement)
and the reactor building ventilation system.
Ob.7 active
To ensure that the structure and components of the reactor building
and ventilation systems are capable of minimizing the release of radio-
activity to the atmosphere during potential abnormal conditions.
Specification SR 5.5.1 - Reactor Building, Surveillance Requirements
The instrumentation which monitors the reactor building sub-atmospheric
pressure will be functionally tested once every month and calibrated
once a year.
Basis for Specification SR 5.5.1
The reactor building atmosphere is normally maintained slightly below
atmospheric pressure by the ventilation system (see FSAR Section 6.1.3.2).
This requirement minimizes the amount and consequences of airborne activity
released from the plant under most conditions (see FSAR Section 14.12.8) .
The leak rate of the building itself is not a significant parameter as is
shown in FSAR Section 6.1.4.2.
Specification SR 5.5.2 - Reactor Building Pressure Relief Device, Surveillance
The reactor building overpressure relief system differential pressure
switches shall to functionally tested on a monthly basis and calibrated
annually.
The louvers shall be exercised annually.
5.5-2
Basis for Specification SR 5.5.2
The reactor building pressure relief device is designed to protect
the building in the event that pressure in the reactor building exceeds
the turbine building pressure by 3 inches of water. The device consists
of louvers installed in a number of individual modules operated by
mechanical linkages to pneumatic actuators (see FSAR Section 6.1.3.4) .
The specified test frequency shall ensure the operability of the reactor
building relief system.
Specification SR 5.5.3 - Reactor Building Exhaust Filters, Surveillance
The exhaust filters in the reactor building ventilation system shall
be tested as follows:
a) The charcoal filters shall be tested once a calendar year to
demonstrate an iodine removal efficiency of at least 99% for
elemental iodine.
b) A test shall be conducted once a calendar year to demonstrate
that gas bypassing of the charcoal filters does not exceed 1%.
The test shall be conducted at normal flow conditions.
c) The HEPA filters shall be tested once a calendar year to
demonstrate that the removal efficiency is at least 95% for
particulates 0.3 micron or greater in size.
d) The associated temperature instruments and controls shall be
functionally tested every three months.
Basis for Specification SR 5.5.3
The iodine filter efficiency was assumed to be 90% for accident
calculations which is quite conservative when compared to the test
efficiency of at least 98%. The HEPA filters were assumed to be 95%
efficient for accident calculations, which is a factor of 10 less than the
5.5-3
specified test efficiency of at least 99.5%. (See FSAR Section 14.12.3) .
The minimum efficiency for the charcoal in the iodine removal filters
for removal of elemental iodine is expected to be 99.9+% at a relative
humidity < 99% upon delivery.
The corresponding test conditions upon which this efficiency is based
are a superficial flow velocity of 40 fpm across a 2 inch thick absorber
bed, at a temperature of 170°F, and for a steam-air mixture at a pressure
of 26 inches of water.
Testing of gas bypassing of the charcoal filters will be by the Freon
method. The HEPA filters will be tested by the D0P method.
The HEPA filters meet all of the requirements of AEC Health and Safety
Bulletin 212, dated June 25, 1965, covering Military Specification
MIL-F-51068(A) , dated April 23, 1964. They also carry the UL label
indicating full compliance with the requirements of UL Standard UL-586.
The HEPA filters are certified to have a tested efficiency of 99.9% at
rated air flow. (FSAR Section 6.1.3.2.3).
The test requirements for the filters ensure that removal of halogens
and particulates would be adequate in the event hypothesized accident
situations should occur.
5.6-1
5.6 EMERGENCY POWER SYSTEMS - SURVEILLANCE REQUIREMENTS
Applicability
Applies to the surveillance of the equipment supplying electrical
power to the essential plant services.
Objective
To establish the minimum frequency and type of surveillance for equipment
supplying electric power to the plant auxiliaries to ensure that the motive
power sources required to safely shut down the plant is available.
Specification SR 5.6.1 - Standby Diesel Generator Surveillance
The surveillance of the standby diesel generators shall be as follows:
a) Each standby generator set will be started and loaded to at
least 50% of rated full load capacity once every week. The
test shall L.ontinue for at least two hours to enable the engine(s)
and the generator to attain their normal operating temperature.
b) A loss of outside source of power and turbine trip shall be
simulated twice annually with the interval between tests to be
not less than four (4) nor greater than eight (8) months to
demonstrate that the standby generators , automatic controls, and
load sequencers are operable.
c) The diesel engine protective functions shall be calibrated annually.
d) The diesel engine exhaust temperature "shutdown" and "declutch"
shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.6.1
The weekly test of the standby diesel generator is to exercise the
engine by operating at design temperature and to demonstrate operating
capability. These tests will allow for detection of deterioration and
failure of equipment.
5.6-2
Tests once a year during refueling will functionally test the
standby generator system.
Specification SR 5.6.2 - Station Battery Surveillance
The surveillance of the station batteries shall be as follows:
a) The specific gravity and voltage of the pilot cell and temperature
of adjacent cells and overall battery voltage shall be measured
every week.
b) The specific gravity and voltage to the Nearest 0.01 volt,
temperature of every fifth cell and height of electrolyte shall
be measured every three months.
c) The station batteries will be load tested to partial discharge
once a year during plant shutdown.
Basis for Specification SR 5.6.2
The type of station battery surveillance called for in this
specification has been demonstrated through experience to provide a
reliable indication of a battery cell initial breakdown well before it
becomes unserviceable. Since station batteries will deteriorate with
time, these periodic tests will avoid precipitious failure.
The manufacturer's recommendation for equalizing charge is vital to
maintenance of the ampere-hour capacity of the battery. As a check upon
the effectiveness of this charge, the battery will be loaded to determine
its ampere-hour capacity. In addition, its voltage is monitored as a
function of time. If a cell has deteriorated or if a connection is
loose, the voltage under load will drop excessively, indicating need for
replacement or maintenance.
5.7-1
ti 5.7 FUEL HANDLING AND STORAGE SYSTEMS - SURVEILLANCE REQUIREMENTS
Applicability
Applies to surveillance of the fuel handling and fuel storage systems
during irradiated fuel handling and storage.
Objective
To ensure the prevention of any uncontrolled release of radioactivity
during fuel handling and fuel storage by establishing the minimum frequency
and type of surveillance on the equipment for the fuel handling and
storage systems.
Specification SR 5.7.1 - Fuel Handling Machine Surveillance
The surveillance of the fuel handling machine will be as follows:
a) Prior to refueling, the fuel handling machine cooling water
leak detector will be functionally tested.
b) A functional test of the Fuel Handling Machine and Isolation
Valve movements, interlocks, limit switches, and alarms will
be performed or simulated prior to annual refueling periods.
Basis for Specification SR 5.7.1
The fuel handling machine provides for the safe refueling of the
reactor. To assure the reliability of the fuel handling machine during the
refueling operation, the machine and its associated interlocks , limit
switches and alarms will be tested prior to refueling. All motions of
the machine should be cycled, including the pick-up and release of a
dummy element. A test of the helium system and the cooling system will
be made. These checks will assure the capability to maintain the proper
atmosphere environment within the machine to prevent any uncontrollable
5.7-2
release of activity, proper purging and back filling capabilities ,
and the capability to maintain temperature of fuel elements within
the machine below 750°F.
Specification SR 5.7.2 - Fuel Storage Facility Surveillance
The surveillance of the fuel storage facility will be as follows :
a) The fuel storage facility helium pressure indicators and
alarms will be calibrated and functionally tested annually.
b) The fuel storage facility cooling system flow indicators, and
flow and temperature alarms shall be calibrated and functionally
tested annually.
Basis for Specification SR 5.7.2
The fuel storage wells are provided for safe storage of new and
irradiated fuel elements. The basic design of the wells is to provide
a low temperature dry helium environment. All conditions connected with
this requirement are monitored by pressure, temperature, and flow
sensitive devices. The temperature and flow detecting devices maintain
surveillance of the wells' two independent cooling systems and are set
to alarm at previously determined maximum or minimum values. The pressure
sensitive device is available to guard against any over-pressurization of
the wells. The specified annual surveillance interval is sufficient to
insure proper operation of the instrumentation.
5.8-1
5.8 RADIOACTIVE EFFLUENT DISPOSAL SYSTEMS - SURVEILLANCE REQUIREMENTS
Applicability
Applies to surveillance of the Radioactive Effluent Disposal Systems .
Objective
To establish the minimum frequency and type of surveillance on the
equipment of the Radioactive Effluent Disposal Systems to assure that
releases of radioactivity are within those specified in Section 4.8.
S ecification SR .8.1 — Radioactive Gaseous Effluent S stem Surveillance
The surveillance of the radioactive gaseous waste disposal system
shall be as follows:
a) Automatic vent high activity blocking and transfer functions
of the gaseous waste system shall be tested prior to each
controlled release or once a month, whichever is more frequent.
b) Automatic gaseous waste header high activity transfer to the
gas waste vacuum tank shall be tested once per month.
c) The gas waste header activity monitors shall be functionally
tested once per month and calibrated quarterly.
d) The vent monitor system shall be functionally tested weekly, calibrated
quarterly, and following maintenance on the detector system.
e) Flow recorders shall be calibrated annually.
f) The vent iodine/particulate monitor filter shall be analyzed
once per week.
Specification SR 5.8.2 - Radioactive Liquid Effluent System Surveillance
The surveillance of the radioactive liquid waste disposal system
shall be as follows:
5.8-2
a) The level alarms and pump interlocks on the two liquid
waste receiver tanks and monitoring tank shall be tested
once per year.
b) The liquid effluent discharge blocking valve shall be
functionally tested prior to each release or once a
month, whichever is more frequent.
c) The activity monitors of the liquid waste disposal line and
the low cooling water blowdown flow switch shall be
functionally tested prior to the controlled discharge of any
liquid wastes or once a month, whichever is more frequent. The
activity monitors shall be calibrated quarterly and following
maintenance on the detector system.
Basis for Specification SR 5.8.1 and 5.8.2
The frequency specified above is based upon industry experience
and minimal disposal requirements of the plant. Tests prior to
discharge using the installed check source mounted in the instrument
will provide both a check on the calibration as well as a dynamic test
of the various monitors, alarms, and protective functions.
5.9-1
5.9 ENVIRONMENTAL SURVEILLANCE - SURVEILLANCE REQUIREMENTS
Applicability
Applies to sampling for environmental radioactivity in the vicinity
of the plant.
Ob.tective
To establish a sampling schedule which will recognize changes in
radioactivity in the environs and assure that effluent releases are kept
as low as practicable and within the limits of Appendix B, Table II , 10 CFR 20.
Specification SR 5.9.1 - Environmental Radiation, Surveillance Requirements
1. Gaseous Release
Sampling of air, external gamma, milk, forage and crops shall be con-
ducted in accordance with Action Guide 3 during the first three years of operation
and thereafter in accordance with Table 5.9-1 and Table 5.9-2, as specified below:
a) If releases from the plant vent produced concentrations or
exposures less than 3% of those specified in 10 CFR 20 for
unrestricted areas and the general population during the
previous quarter, the environmental survey shall be conducted
in accordance with Action Guide 1 for the current quarter.
b) If the concentrations or exposures during the previous quarter
were greater than 3% but less than 10% of those specified in
10 CFR 20 for unrestricted areas and the general population,
the environmental survey shall be conducted in accordance with
Action Guide 2 for the current quarter. If the samples taken
under Action Guide 2 do not indicate any significant increase
in environmental radioactivity, the survey shall revert to
Action Guide 1.
5.9-2
1
c) If the concentrations or exposures during the previous quarter were
greater than 10% of those specified in 10 CFR 20 for unrestricted
areas and the general population, the environmental survey 'shall be
conducted in accordance with Action Guide 3 for the current quarter.
If the samples taken under Action Guide 3 do not indicate any
significant increase in environmental radioactivity, the survey
shall revert to Action Guide 2.
2. Liquid Release
Sampling of water and silt, potable water, and aquatic biota shall be con-
ducted in accordance with Action Guide 3 during the first three years of operation
and thereafter in accordance with Table 5.9-1 and Table 5.9-2 as specified below:
a) If the gross beta-gamma activity released from the station during
the previous quarter was less than 3% of MPCw, the environmental survey
shall be conducted in accordance with Action Guide 1 for the current quarter.
b) If the gross beta-gamma activity released from the station during the
previous quarter was greater than 3% MPCw but less than 10% MPCw,
the environmental survey shall be conducted in accordance with
Action Guide 2 for the current quarter. If the samples taken under
Action Guide 2 do not indicate any significant increase in environmental
radioactivity, the survey shall revert to Action Guide 1.
c) If the gross beta-gamma activity released from the station during
the previous quarter was greater than 10% of MPCw, the environmental
survey shall be conducted in accordance with Action Guide 3 for the
current quarter. If samples taken under Action Guide 3 do not
indicate any significant increase in environmental radioactivity,
the survey shall revert to Action Guide 2.
1
5.9-3
d) Results of the aquatic biota sampling program will be reviewed
with appropriate agencies after one year of sampling following
commercial operation to establish the required extent of future
sampling.
Basis for Specification SR 5.9.1
Programs for monitoring the environment in the vicinity of Ft. St. Vrain
will be conducted by Colorado State University under a contract from Public
Service Company of Colorado (the licensee) and by the Colorado Department
of Health with assistance by the Environmental Protection Agency's Western
Environmental Radiation Laboratory. The Colorado Department of Health program
includes sampling and analyses of air, water and milk. In addition, they
will have special programs for sampling tritium in surface water and
atmospheric concentrations of 85Kr.
A preoperational radiological monitoring program has been conducted
since March 1969. This program has established an adequate baseline to
which operational environmental data can be compared.
The operational environmental surveillance program will be maintained
on a continuous basis to verify that projected and anticipated concentrations
of radioactive materials in the environment are not exceeded. The extent
to which environmental monitoring programs are conducted should depend on
the actual release of radioactivity into the environment.
When the quantity of material released is small the environmental
monitoring program may be minimal. For larger releases of radioactive
material, a more comprehensive environmental monitoring program is appropriate.
The surveillance levels specified in Action Guide 1 and Action Guide 2 are
comparable to intake Range 1 and Range 2 as given in Federal. Radiation
Council Report No. 2.
5.9-k
The operational surveillance program provides for collection and
analyses for samples within an area extending to a twenty mile radius from
the reactor. A concentrated area of sampling within a one mile radius is
designated the facility zone; the area from one to ten miles is called the
adjacent zone, and the reference zone is from ten to twenty miles.
Table 5.9-2 gives the location of each sampling station and the types
of samples to be taken at each station. Table 5.9-3 gives the minimum
sensitivities for the various analyses and/or measurements made on the
samples. Figure 5.9-1 and Figure 5.9-2 indicate the sample station locations.
The aquatic biota sampling program is a supplemental part of the
Environmental Surveillance Monitoring Program and was not a factor in the
design of the basic sampling program which was designed on the basis of
critical pathways to man.
It is felt that sampling during the preoperational phase and for a
representative period following operation will adequately demonstrate any
potential effect of plant operation on aquatic biota. Therefore, it is
planned that the results of the aquatic biota sampling program will be
reviewed with representatives from interested agencies such as the Bureau
of Sport Fisheries and Wildlife and the Colorado Game, Fish and Parks
Department following one year of commercial operation to establish the
extent to which sampling should be continued beyond that point.
For further information, see FEAR Section 2.7.
5.9-5
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..._...- Revised 5-6-71.
5.9-10
05
87Tr
GREELEY
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Figure 5.9-2 Sampling stations in the adjacent and reference
areas (off site). Revised 5-6-71.
5.9-11
TABLE 5.9-3 .
Sensitivities for environmental radiation measurements.
Counting Analytical M.D.A.
Media Isotope Efficiency Technique 99% C.L.
Air Gross a 48% Int. Prop. Ctr. (a)
Gross 8 26% Low Beta G. M. 0.01 pCi/m3
137Cs 21% Gamma Spect. 0.03 pCi/m3
95Zr-99Nb 30% Gamma Spect. 0.02 pCi/m3
10611.u4.3% Gamma Spect. ' 0.20 pCi/m3
144Ce 6.8% Gamma Spect. 0.10 pCi/m3
1311 (a) Gamma Spect. (a)
3H 25% Liquid Scint. 6.8 pCi/m3
(a)
Extern Thermolumines-
Gamma cent Read-out 40 mR (b)
Forage Gross 8 8% Low Beta G.M. 3.0 pCi/g
137Cs (a) Gamma Spect. 35 pCi
3H 25% Liquid Scint. 2000 pCi/1
Milk 137Cs 6.7% Gamma Spect. 2.0 pCi/1
1311 (a) Gamma Spect. (a)
99, 90Sr 8% Chem. Separation 10.0 pCi/i
3H 25% Liquid Scint. 2000 pCi/1
137Cs 6.7% Gamma Spect. 2.0 pCi/1
Water Gross 9 (c) Low Beta G.M. 1.0 pCi/1
3H 25% Liquid Scint. 2000 pCi/1
Sediment Gro:;; 8 8% Low Beta G.M. 3.0 pCi/g
(a) To be determined.
(b) Radiation exposure necessary to produce a response equal to 3 sigma
of the background as determined over a three month period.
(c) Dependent upon amount of dissolved and suspended solids.
(d) Depends upon relative humidity.
6.0-1
6.0 DESIGN FEATURES
The design features specified in this section define the design
characteristics of special importance to each of the physical barriers
and to the maintenance of safety margins which have not been covered
in any other specifications. The principal objective of this section
is to control changes in the design of vital portions of the facility,
with particular emphasis on those components which are not covered in
any other specification of these technical specifications.
6.1-1
6.1 REACTOR CORE - DESIGN FEATURES
Avplic_ ate?
Applies to the general design features of the reactor core including
fuel, moderator, reflector and reactivity control.
Obiective
To define the vital design characteristics of the reactor core
to control changes in the design features of the fuel, moderator ,
reflector and reactivity control.
S ecification DF 6.1 - Reactor Core Desi n Features
The following discussion describes the design features which shall
be incorporated in the reactor core:
Reactor Asse ,
The reactor core consists of: (1) removable fuel elements which
contain the fuel (U & Th) , the moderator (graphite) and burnable poison
(boron) , and (2) radial and axial reflectors which consist of removable
reflector elements and permanent blocks which are made of graphite, and
in some cases incorporating boron or structural steel. The reactor core
assembly, including reflector, has an overall assembly height of about
23.9 feet and a diameter of about 27.3 feet. The approximate weight of
the core assembly is 1,348,000 pounds. The preceding description includes
the core support graphite blocks.
The reactor reactivity control consists of 37 pairs of control rods
containing boron carbide, which are supplemented by burnable poison (boron)
in selected fuel elements as required. A reserve shutdown system consisting
6.1-2
of 37 hoppers of boron carbide-graphite balls is also provided.
A variable orifice flow-control assembly is located at the inlet to
each of the 37 refueling regions to provide adjustment of the coolant
flow through the region.
Active Core
The active core consists of 1482 hexagonal graphite fuel elements
stacked in 247 vertical fuel columns. The fuel elements form the active
core which is essentially a right circular cylinder 15.6 feet in height
and 19.5 feet in equivalent diameter. The active core is completely
surrounded by a graphite reflector. Within the core array, the fuel
columns are grouped into 37 refueling regions containing seven fuel
columns each, except for six outer corner regions which contain five
fuel columns each.
The center fuel column of each of the 37 fuel regions is a control
rod column. Each control rod column contains two control rod channels
and one reserve shutdown absorber material channel. Each control rod
channel has a diameter of 4.o inches and the two channels have a
centerline pitch spacing of 9.7 inches. The reserve absorber material
shutdown channel has a diameter of 3-3/4 inches. The control rod channels
are continuous from the top face of the top reflector and terminate in the
bottom reflector at an elevation not greater than 27.0 inches above the
top face of the core support block. The reserve shutdown absorber channel
is continuous from the top face of the top reflector and terminates in the
bottom fuel element at an elevation not greater than 47.5 inches above the
top face of the core support block.
6.1-3
Each fuel element is a hexagonal right prism with nominal dimensions
of 14.2 inches across the flats by 31.2 inches high. The fuel beds and
coolant channels are distributed on a triangular array of about 3/4 inch
pitch spacing with an ideal ratio of two fuel beds for each coolant
channel. The bottom of the fuel beds in the bottom fuel element of the
control rod fuel column does not exceed a length of 23.1 inches from the
top face of the fuel element.
Fuel
The fuel consists of fissile uranium highly enriched (93.15%) in 235U
and fertile thorium. The initial fuel loading is about 773 Kg of uranium
and 16,000 Kg of thorium. The initial core is loaded with 13 fuel
compositions whose distribution within the core is designed to mock up the
fuel content of the equilibrium cycle refueling regions and to shape the
radial and axial power distribution. Fuel is designed for up to a six
year life. About one-sixth of the core will be replaced at each refueling
interval. The fuel loading in a reload segment will be about 200 Kg of uranium
and 2300 Kg of thorium.
All uranium and thorium is in the form of heavy metal carbide kernels
coated with silicon carbide and pyrocarbon, referred to as coated fuel
particles. The coatings form the primary fission product barrier. The
coated fuel particles consist of two general types, fissile particles
(Th:U C2) and fertile (ThC2) particles. The fissile particles shall
contain thorium and uranium in a weight ratio of about 4.25 to 1 of thorium
to uranium. The fertile particles shall contain only thorium. The coated
fuel particles are bonded together with a carbonaceous material to form
fuel rods. The fuel rods are completely surrounded and contained by
6.1-4
graphite which forms the structural part of the fuel element and, in
addition to the carbon contained within the fuel rods, also serves as
the sole moderator.
Reflector
Reflector elements above, below and immediately adjacent to the
side of the active core are hexagonal right prisms with nominal dimensions
of 14.2 inches across flats and 15.6, 23.4 or 31.2 inches high, as required.
The outer peripheral envelope of the reactor core reflector graphite
contains boron to minimize the neutron flux leaving the reflector.
The side reflector contains nominal 2.3 weight percent boron stainless steel
pins within the spacer blocks. The middle layer of lower reflector
elements excluding the central element in each core region contains
25 weight percent boronated graphite pellets enclosed in hastalloy-X cans. The
top layer of reflector above the hexagonal columns contains 1 weight
percent crushed boronated graphite. The top layer of reflector above the
permanent side reflector blocks contains 1 weight percent boronated
graphite enclosed in steel cans.
Basis for Specification DF 6.1
The above specifications form the general design bases and criteria
for the overall design features of the reactor core which were used to
evaluate its general performance. Further details concerning these design
features are given in Section 3.0 of the FSAR.
6.2-1
6.2 REACTOR COOLANT SYSTEM AND STEAM PLANT SYSTEM - DESIGN FEATURES
Applicability
Applies to the vital design characteristics of the primary and
secondary coolant systems.
Objective
To control changes in the primary and secondary coolant systems.
Specification DF 6.2.1 - PCRV, Design Features
The PCRV is constructed of high strength concrete reinforced with
bonded reinforcement steel and prestressed with steel tendons. Prestressing
tendons, located in conduits embedded in the concrete are used to prestress
the entire structure. Access to the tendons is provided so that most
tendons can, if necessary, be retensioned or selectively removed for
inspection and replaced. The following table gives the type and number
of tendons in the PCRV:
Number of 1/4"
Type of Tendon Wires per Tendon Number of Tendons
Longitudinal 169 90
Circumferential 100
a) Head 169
b) Wall 152 210
Bottom Cross Head 169 24
Top Cross Head 169 24
The temperature of the PCRV concrete is controlled by means of insulation
mounted on the inside surface of the liner, and cooling tubes welded to
the concrete side of the liner. The whole of the internal surface of the
liner is covered by the thermal barrier which uses Kaowool insulation,
6.2-2
a ceramic fiber blanket material of high chemical purity which is about
50% alumina and 50% silica.
The various PCRV penetrations required for refueling, maintenance,
control rods, or operation of circulators and steam generators are
provided with liners that are welded to the cavity liner and extend
through the concrete to provide leak-tight access to the reactor internals.
Each penetration is provided with two closures in series, a primary closure
and a secondary closure. The primary closures , together with the PCRV
structure, contain the radioactive primary coolant in a manner analogous
to a conventional primary vessel. The secondary closure encloses the primary
closure and contains the radioactive primary coolant that might be released
from a leaking primary closure in a manner analogous to a conventional secondary
containment. The primary and secondary closures are similar to conventional
pressure vessel closures and are flat or formed heads. The closures
incorporate elastomer or metallic seals, depending on the operating
temperatures, and are attached to the penetration liner flanges with bolts
or shear rings. Independently anchored flow restriction devices are in
the larger penetrations to limit the flow rate of primary coolant if both
primary and secondary closures were to fail completely and instantaneously.
Basis for Specification DF 6.2.1
The PCRV is the structure that contains the reactor core and the entire
primary coolant system including steam generators and helium circulators.
It functions as the primary coolant pressure boundary and its design results
in an exceedingly low probaoility of gross rupture or significant leakage
throughout its design life. In addition, it incorporates features to back
up the reactor coolant pressure boundary, such as secondary closures in all
6.2-3
penetrations, plus flow restriction devices in those penetrations
that require them.
Applicable design codes, PCRV technology development, design and
analysis, construction and quality control tests and inspections, which
provide the basis for the PCRV design are presented in Section 5.0 of
the FSAR. The performance objectives for the PCRV are to provide
adequate strength, leak-tightness, biological shielding, and predictable
safety during all normal operating and credible accident conditions.
Additional information in support of the PCRV design is given in
Appendix E of the FSAR.
Specification DF 6.2.2 - Steam Generator Orifices, Design Features
The steam generator modules are provided with two sets of orifices:
a) The variable feedwater ringheader trim valves, which include
mechanical stops to prevent total closure.
b) The fixed feedwater orifices in the economizer tube inlets.
These flow limiting devices are provided to limit water/steam inleakage
to the primary coolant through a subheader tube rupture as described in
FSAR Section 6.4.
Basis for Specification DF 6.2.2
The feedwater flow limiting devices were selected to limit the maximum
inleakage from a single tube rupture to limits specified in Sections 6.4
and 14.5 of the FEAR.
Specification DF 6.2.3 - Steam Safety Valves , Design Features
The steam plant contains the following steam pressure safety valves,
with set pressures and capacities as shown below:
6.2-4
Set
Press.
Valve Quantity psig Type and Capacity
Main Steam Three/Loop 2720 ASME Code, Section III-A,
Safety 2790 Spring-Loaded Valves;
2860 105% of Loop Flow (Total)
Reheater One/Loop 1100 ASME Code, Section III-A,
Safety Spring-Loaded Valve; 55,000
lb/hr of Saturated Steam
@ 1100 psig
Bypass Flash 6 975 ASME Code, Section VIII,
Tank to 1020 Spring-Loaded Valves;
105% Plant Capacity (Total)
Hot Reheat 6 700 ASME Code, Section VIII,
to 735 Spring-Loaded Valves;
105% Plant Capacity (Total)
Basis for Specification DF 6.2.3
Main steam safety valves are provided in accordance with ASME Code,
Section III, Class A requirements and, when all three are fully open, will
prevent superheater outlet pressure from exceeding 3025 psig (110% of design
pressure). These valves discharge to atmosphere. The reheater safety valve
is sized to prevent over-pressure in the event of an accident involving the
requirement for flooding a reheater with condensate, followed by an operator
error in closing both the reheater inlet and outlet valves. These safety
valves discharge to the reactor building ventilation system exhaust filters.
The bypass flash tank and hot reheat line safety valves prevent over-
pressure of the cold reheat and the hot reheat piping, respectively. As
long as either the cold reheat or the hot reheat block valves are open,
these valves also prevent over-pressure of the reheaters. These valves
discharge to atmosphere.
6.3-1
6.3 SITE DESIGN FEATURES
Applicability
Applies to the location and extent of the Reactor Site.
Objective
To define those aspects of the site which affect the overall safety
of the installation.
Specification DF 6.3 - Site, Design Features
The Fort St. Vrain Nuclear Generating Station, Unit No. 1, is
situated on a tract of land located about 3.5 miles northwest from the
center of Platteville, Colorado. The tract is situated in Weld County,
Colorado (See FSAR Section 2.1) .
The exclusion area is approximately 1 mile square and is defined in
FSAR Fig. 2.1-3. The closest distance from the reactor building to the
boundary of the exclusion area is 1,935 feet. The limits of 10 CFR 20
shall apply at the boundary of this exclusion area. The Low Population
Zone (LPZ) is defined by a radius of 16,000 meters. The exclusion area
is zoned industrial, and the area surrounding the exclusion area is zoned
agricultural. Agricultural activities may continue on the site including
a portion of the exclusion area, and an evacuation procedure will be
maintained. There are no permanent residences located within the exclusion
area.
A security fence surrounds the plant area, as shown in FSAR Fig. 1.2-2.
Fences inside the security fence limit routine access into the plant from
the parking lot inside the main gate to the main plant entrance. The
main gate is electrically operated and controllable from within the plant.
6.3-2
An Information Center is located within the exclusion area, but
outside the main gate. An evacuation procedure will be maintained for
the Information Center.
Basis for Specification DF 6.3
The site offers adequate distances and favorable seismologic,
meteorologic, geologic , hydrologic , and population characteristics
as described in Section 2 of the FSAR. The favorable characteristics of the
site and the design of the plant ensure that 10 CFR 100 and 10 CFR 20
requirements can be met satisfactorily.
7.0-1
7.0 ADMINISTRATIVE CONTROLS
Administrative controls described in this section specify the
procedures, record keeping, review and audit systems, and reporting
that are required to provide assurance and documentation that the
plant is managed in a safe and reliable manner. These controls also
specify the administrative action which must be taken in the event
that a prescribed limit, setting or condition specified in these
Technical Specifications is exceeded or violated.
7.1-1
7.1 ORGANIZATION, REVIEW AND AUDIT-ADMINISTRATIVE CONTROLS
Applicability
Applies to the lines of authority and responsibility for the operational
safety of the facility, and the organization for periodic review and audit
of facility operation.
Obj ect ive
To define the principal lines of authority and responsibility for
providing continuing review, evaluation and improvement of plant
operational safety.
Specification AC 7.1.1 - Organization, Administrative Controls
The organization and lines of responsibility which govern plant
operation shall be as follows:
a) The Superintendent is directly responsible for the safe
operation of the facility.
b) In all matters pertaining to operation of the plant and to
these Technical Specifications, the Plant Superintendent shall report
to and be directly responsible to the Superintendent, Outside Steam
Plants. The administrative and departmental organizations are shown
in Figures 7.1-1 and 7.1-2.
c) Organization for conduct of operations of the plant is shown
in Figure 7.1-3.
1. A licensed senior operator shall be present on site
at all times when there is fuel in the reactor.
7.1-2
2. A licensed operator must be in the control room at all times
when fuel is in the reactor. During reactor startup, shutdown,
and recovery from reactor trip, two licensed operators must be
in the control room.
3. A licensed senior or special "fuel handling" senior operator
shall be in charge of any refueling operation.
4. An operator or technician, qualified in radiation protection
procedures, shall be present at the facility at all times
that there is fuel on site.
Initial staffing of the plant shall be as described in the FSAR.
From beginning of fuel loading, and until such time as the start-up tests
and demonstration run has been completed, each operating shift shall consist
of at least six persons, including at least one licensed senior reactor
operator, two licensed reactor operators, and one or more staff members
from the NSSS vendor's staff or consultants, who, by virtue of their training
and experience, can provide competent technical support for the start-up
and power ascension program.
American National Standards Institute N18.1-1971, "Selection and Training
of Personnel for Nuclear Power Plants", shall be used as a guide to selecting
and training replacement plant personnel and to retraining requirements for
those persons presently on the staff during the first years of operation.
At the beginning of the fourth year following the start of commercial
operation, the staffing of the plant shall be in accordance with American
National Standards Institute N18.1-1971, "Selection and Training of Personnel
for Nuclear Power Plants."
Basis for Specification AC 7.1.1
The lines of responsibility for plant operation are consistent with
7.1-3
those for other plants on the Public Service Company of Colorado system.
The plant organization provides for a sufficient number of qualified personnel
to operate the plant in a safe manner.
Specification AC 7.1.2 - Plant Operations Review Committee, Administrative
Controls
There shall be a Plant Operations Review Committee. Its organization,
responsibilities, and authority shall be as follows:
a) Membership
1. Chairman: Plant Superintendent, or designated alternate
(Assistant Plant Superintendent)
2. Assistant Plant Superintendent
3. Senior Health Physicist
4. Senior Results Engineer
5. Maintenance Supervisor
6. Shift Supervisor (any one of five)
7. Union Representative (a licensed operator)
b) Meeting frequency: Monthly, and as required, on call of the Chairman.
c) Quorum: Chairman, or designated alternate, plus three members.
d) Responsibilities:
1. Review all proposed changes for normal and emergency
operating procedures, and any other proposed changes
or procedures that are determined by the Plant
Superintendent to affect Nuclear safety.
2. Review proposed changes to the Technical Specifications.
3. Review all proposed revisions or modifications to plant
systems or equipment, which revision would require a change
in procedure 1) above.
7.1-4
4. Investigate all alleged violations of Technical Specifications,
license provisions , administrative procedures, operating
procedures , and regulatory requirements; such investigations
to include reporting, evaluation, and recommendations to
prevent recurrence, to the Superintendent Outside Steam Plants
and to the Chairman of the Nuclear Facility Safety Committee.
5. Review all procedures required by these Technical Specifications
annually.
6. Review proposed tests and experiments and their results.
7. Review plant operations to detect any potential safety hazard,
or abnormal performance of plant equipment.
8. Assure itself that a daily review of plant operations is made
by the responsible supervisors.
9. Review abnormal occurrences.
10. Conduct annual drill of "off-site Emergency Plan", including
evacuation of the site and evaluate implementing procedures
and communications with off-site support groups.
11. Perform special reviews and investigations, as requested by
the Nuclear Facility Safety Committee.
e) Authority
1. The Plant Operations Review Committee will act in an advisory
capacity to the Plant Superintendent on all matters brought
before it.
2. The plant Operations Review Committee shall make tentative
determinations as to whether or not proposals considered by
the Committee involve unreviewed safety questions. This
determination shall be subject to review and approval by
the Nuclear Facility Safety Committee.
7.1-5
f) Records
Minutes shall be kept of all meetings of the Plant Operations
Review Committee and copies shall be forwarded to the Superintendent,
Outside Steam Plants, to the Chairman of the Nuclear Facility
Safety Committee, and all members of the Plant Operations
Review Committee.
Basis for Specification AC 7.1.2
The Plant Operations Review Committee will provide a mechanism for periodic
review of plant operation by plant personnel who are directly familiar with and
responsible for all aspects of plant operation. The activities of this
Committee will ensure that there is a periodic review of plant operations
affecting safety.
Specification AC 7.1.3 - Nuclear Facility Safety Committee, Administrative
Controls
There shall be a Nuclear Facility Safety Committee. Its organization,
responsibilities, and authority shall be as follows:
a) Membership
All members of the Committee shall be designated by the
appropriate Vice President, as follows:
1. Technically qualified persons who are not members of
the plant staff who collectively provide expertise in:
a. Gas-Cooled Power Reactor Engineering
b. Nuclear Power Plant Technology
c. Reactor Operations
d. Chemistry and Radio Chemistry
e. Instrumentation and Control Systems
f. Radiation Safety
7.1-6
g. Mechanical and Electrical Systems
h. Metallurgy and Radiation Damage
i. Others as required
2. Plant Superintendent or designated alternate (as an observer)
Members of the Nuclear Facility Safety Committee
may be from the owners organization, or be outside
consultants. An individual may possess expertise
in more than one specialty.
b) Meeting frequency: During the first year of operation, meetings
shall be held at least every six weeks and thereafter, at least
three times annually, the interval between meetings not to exceed
five months, and as required on call of the Chairman.
c) Quorum: Chairman or Vice Chairman, plus a majority of the
permanent members.
d) Responsibilities
1. Review proposed changes in Technical Specifications
and operating license.
2. Review minutes of meetings of the Plant Operations
Review Committee, to determine if matters considered
by that Committee involve unreviewed or unresolved
safety questions.
3. Review matters including proposed changes, or
modifications, to plant systems or equipment having
safety significance, or referred to it by the Plant
Operations Review Committee or by the Plant Superintendent.
7.1-7
4. Conduct audits no less than semi-annually of facility
operations for compliance with internal rules , procedures,
regulations, and license requirements, including Technical
Specifications.
5. Assure itself that one or more Committee members visit
the plant at least once per month.
6. Investigate all reported instances of alleged violations
of Technical Specifications, AEC regulations, License
requirements, internal procedures or instructions, and
abnormal occurrences or performance of plant equipment
and anomalies.
7. Follow-up action, including re-audit of deficient
areas, shall be taken as required.
8. Review any indication that there may be a deficiency
in some aspect of the design or operation of safety
related systems or components..
9. Insure itself that it receives all information
necessary for it to fulfill its obligations and
responsibilities on a time scale such that it can
take effective action.
10. Review Security and Emergency Plans and their
implementing procedures.
11. Review the results of the Environmental Monitoring Program.
12. Review proposed tests and experiments and their results.
e) Authority
1. The Nuclear Facility Safety Committee shall report to the
appropriate Vice President.
7.1-8
2. Approve proposed changes to the operating license, including
Technical Specifications and its revised basis, for
submission to the AEC.
3. Approve proposed changes or modifications to plant systems
or equipment, provided such changes or modifications do
not involve unreviewed safety questions.
4. Approve appropriate action to prevent recurrence of
any violations of Technical Specifications.
f) Records
Minutes shall be recorded of all meetings of the Committee. Minutes
of the meeting shall be approved before circulation. Approved copies
of the minutes shall be forwarded to the appropriate Vice Presidents
of Electric Operations and Engineering and Planning, Manager of
Production, Director of Quality Assurance, Superintendent Outside
Steam Plants, Plant Superintendent, and all members of the NFSC.
Basis for Specification AC 7.1.3
The Nuclear Facility Safety Committee will provide a mechanism
for periodic review of safety aspects of plant operation by personnel who are
not part of the plant staff. The activities of this Committee will ensure
that management has effective responsibility for the safe operation of the
plant through the diverse membership of the Committee.
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• 7.2-1
7.2 SAFETY LIMITS, ADMINISTRATIVE CONTROLS
Applicability
Applies to the administrative procedures to be followed in the
event that a Safety Limit is exceeded.
Objective
To define the administrative procedures which will be followed in
the event that a Safety Limit is exceeded.
Specification AC 7.2 - Action to be Taken If a Safety Limit is Exceeded,
Administrative Controls
If a Safety Limit is exceeded, as defined in Specification SL 3.1
and 3.2, the following action shall be taken:
a) The reactor will be shut down immediately and reactor operations
shall not be resumed until approval is received from the AEC.
b) An immediate report will be made to the Superintendent of Outside
Steam Plants and the Chairman of the Nuclear Facility Safety Committee.
c) The Vice President of Electric Operations will be responsible for
reporting of the circumstances to the Director, Regional Regulatory
Operations Office within 24 hours by phone and telegraph, as
specified in Section 7.6, Reporting.
d) A complete analysis of the circumstances leading up to and
resulting from the situation together with recommendations to
prevent a recurrence will be prepared by the Plant Operations
Review Committee. This report will be forwarded to the
Superintendent of Outside Steam Plants and the Chairman of the
Nuclear Facility Safety Committee. Appropriate analysis reports
will be submitted to the Director, Directorate of Licensing
1 .
7.2-2
with a copy to the Director, Regional Regulatory Operations Office, within 10
days, as specified in Section 7.6, Reporting.
Basis for Specification AC 7.2
The procedures specified in the Specification will ensure that the
reactor is maintained in a safe condition, and that a prompt report is
made to responsible individuals and organizations, in the event that a
Safety Limit is exceeded.
7.3-1
7.3 ABNORMAL OCCURRENCE, ADMINISTRATIVE CONTROLS
Applicability
Applies to the administrative procedures to be followed in the
event there is an Abnormal Occurrence.
OW ective
To define the administrative procedures which will be followed in
the event there is an Abnormal Occurrence.
Specification AC 7.3 - Action to be Taken In the Event of an Abnormal
Occurrence, Administrative Controls
In the event of an Abnormal Occurrence, as defined in Section 2.0
of these Technical Specifications, the following action shall be taken:
a) The abnormal occurrence shall be reported to the Superintendent
Outside Steam Plants immediately.
b) The Abnormal Occurrence shall be reviewed by the Plant Operations
Review Committee and a report shall be forwarded to the Superintendent
Outside Steam Plants and the Chairman of the Nuclear Facilities
Safety Committee, including an evaluation of the cause of the
occurrence and recommended action to be taken to prevent recurrence
c) The appropriate Vice President shall be promptly notified. He
shall be responsible for reporting of the circumstances to the
Director, Regional Regulatory Operations Office within 24 hours by
phone and telegraph, as specified in Section 7.6, Reporting.
Basis for Specification AC 7.3
The procedures specified in this Specification will ensure that
Abnormal Occurrences will be properly investigated and reported.
7.4-1
7.4 RECORDS - ADMINISTRATIVE CONTROLS
Applicability
Applies to the records of operation which will be maintained.
Objective
To ensure that an adequate record of plant operation is maintained
to verify that the plant is operated in a safe manner.
Specification AC 7.4 - Records, Administrative Controls
Records and logs relative to the operation of Fort St. Vrain Unit
No. 1 shall be maintained in accordance with present Public Service
Company of Colorado policy. Records and logs relative to the following
specific items shall be retained as indicated:
a) Retain for at least 6 years
1. Records of plant operation, including such items as
power level, fuel exposure, and shutdowns.
2. Records of periodic checks , inspections , tests, and
calibrations of components and systems, as related
to these Technical Specifications.
3. Records of changes made to the procedures or equipment,
and records of special reactor tests and experiments.
b) Retain for the Life of the plant
1. Records of radioactive shipments.
2. Records of liquid or gaseous radioactive releases to
the environs.
3. Records of radiation exposures.
4. Records of off-site environmental monitoring surveys.
7.4-2
5. Records of fuel accountability, including inventories
and transfers, and element histories.
6. Records of plant radiation and contamination surveys.
7. Records and print changes made to the plant, as described
in the Final Safety Analysis Report.
8. Records of principal maintenance activities, including
inspections, repairs, and substitution or replacement of
principal items of equipment pertaining to nuclear safety.
9. Records of Abnormal Occurrences.
10. Records of the following plant transients shall be maintained:
a. Reactor Scrams
b. Turbine Trips
c. Primary System Rapid Depressurization
d. Loop shutdowns
e. Loss of Helium Circulator
f. Reheater Cooling after Initial Cooldown
Basis for Specification AC 7.4
The records required by this Specification will be adequate to
verify that the plant is operated in a safe manner, and will provide an
historical record of plant operation which will be used for review
and audit of plant operation.
7.5-1
7.5 PROCEDURES - ADMINISTRATIVE CONTROLS
Applicability
Applies to administrative procedures which will govern plant operations.
Objective
To ensure that written procedures will be maintained to define
requirements for plant operation.
Specification AC 7.5 - Procedures, Administrative Controls
Approved written procedures with appropriate check off lists and
instructions shall be maintained for the following:
a) Plant Operations
1. Integrated startup, operation, and shutdown of the reactor
system and of all systems and components involving nuclear
safety of the facility.
�-' 2. Fuel Handling Operations
3. Actions to be taken to correct specific and foreseen potential
or actual malfunction of systems or components, including
responses to alarms, primary system leaks, and abnormal
reactivity changes, and including follow up actions required
after plant protective system actions have been initiated.
4. Emergency conditions involving potential or actual release
of radioactivity.
5. Surveillance testing and calibration of instrumentation, as
required by these Technical Specifications.
6. Emergency plan procedures.
b) Maintenance
1. Procedures shall be developed and approved prior to implementation
to cover each specific maintenance operation that could involve the
7.5-2
safety of the reactor and personnel, as they are
required.
c) Radiological
1. Radiation control procedures shall be maintained and made
available to all station personnel. These procedures shall
show permissable radiation exposure, and shall be consistent
with the requirements of 10 CFR 20. This radiation protection
program shall be organized to meet the requirements of
10 CFR 20.
2. Pursuant to 10 CFR 20, 103(c) , (1) and (3) , allowance shall
be made for the use of respiratory protective equipment in
restricted areas where individuals are exposed to concentrations
in excess of the limits specified in Appendix B, Table I,
Column 1, of 10 CFR 20, subject to the following conditions
and limitations:
a. The limits provided in Section 20.103(a) and (b)
are not exceeded.
b. If the radioactive material is of such form that
intake through the skin or other additional route is
likely, individual exposures to radioactive material
shall be controlled so that the radioactive content
of any critical organ from all routes of intake
averaged over 7 consecutive days does not exceed that
which would result from inhaling such radioactive
material for 40 hours at the pertinent concentration
values provided in Appendix B, Table I, column 1,
of 10 CFR 20.
7.5-3
c. For radioactive materials designated "Sub" in the
"Isotope" column of Appendix B, Table I , Column 1
of 10 CFR 20, the concentration value specified is
based upon exposure to the material as an external
radiation source. Individual exposures to these
materials shall be accounted for as part of the
limitation on individual dose in 20.101.
d. Respiratory protective equipment is selected and used
so that the peak concentrations of airborne radioactive
material inhaled by an individual wearing the equipment
does not exceed the pertinent concentration values
specified in Appendix B, Table I, Column 1, of 10 CFR 20.
For the purposes of this subparagraph, the concentration
of radioactive material that is inhaled when respirators
are worn may be determined by dividing the ambient airborne
concentration by the protection factor specified in
Table I, appended to this specification for the respiratory
protective equipment worn. If the intake of radioactivity
is later determined by other measurements to have been
different than that initially estimated, the later
quantity shall be used in evaluating the exposures.
e. The licensee advises each respirator user that he may
leave the area at any time for relief from respirator
use in case of equipment malfunction, physical or
psychological discomfort, or any other condition that
might cause reduction in the protection afforded the wearer.
7.5-4
f. The licensee maintains a respiratory protective program
adequate to assure that the requirements above are met.
g. The licensee uses equipment approved by the U. S. Bureau
of Mines under its appropriate Approval Schedules , as
set forth in Table I below.
h. Unless otherwise authorized by the Commission, the
licensee does not assign protection factors in excess of
those specified in Table I in selecting and using
respiratory protective equipment.
3. These specifications, with respect to the provisions of 20.103 shall
be superseded by adoption of proposed changes to 10 CFR 20, Section
20.103, which would make this specification unnecessary.
d) Procedures prepared for a) and c) (1) above shall be reviewed by the
Plant Operation Review Committee. Approval or disapproval will be by
the Plant Superintendent.
e) Temporary changes to procedures prepared for a) and c) (1) above,
which do not change the intent of the original procedures , may be
made with the concurrence of a shift supervisor and one other
person holding a senior operators license. Such changes shall be
documented and subsequently reviewed by the PORC. Final approval
or disapproval will be by the Plant Superintendent.
f) Practice of site evacuation exercises shall be conducted annually,
including a check of communications with off-site support groups.
Annual review of the Emergency Plan shall be performed.
Basis for Specification AC 7.5
This specification will provide for written procedures which will be
available for the guidance of operating personnel for all foreseeable normal
and emergency operating conditions.
7.5-5
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7.5-6
See the following symbols:
CF: Continuous flow
D: Demand
NP: Negative Pressure (i.e. , negative phase during inhalation)
PD: Pressure Demand (i.e. , always positive pressure)
R: Recirculating (closed circuit)
For purposes of this specification the protection factor is a
measure of the degree of protection afforded by a respirator,
defined as the ratio of the concentration of airborne radioactive
material outside the respiratory protective equipment to that inside
the equipment (usually inside the facepiece) under conditions of use.
It is applied to the ambient airborne concentration to estimate the
concentration inhaled by the wearer according to the following formula:
Concentration Inhaled = Ambient Airborne Concentration
Protection Factor
The protection factors apply:
(i) Only for trained individuals wearing properly fitted
respirators used and maintained under supervision in
a well-planned respiratory protective program.
(ii) For air-purifying respirators only when high efficiency
(above 99.9% removal efficiency by U. S. Bureau of
Mines type dioctyl phthalate (DOP) test) particulate
filters and/or sorbents appropriate to the hazard are
used in atmospheres not deficient in oxygen.
(iii)For atmosphere-supplying respirators only when supplied
with adequate respirable air.
7.5-7
3/ Excluding radioactive contaminants that present an absorption or
submersion hazard. For tritium oxide approximately half of the
intake occurs by absorption through the skin so that an overall
protection factor of not more than approximately 2 is appropriate
when atmosphere-supplying respirators are used to protect against
tritium oxide. Air-purifying respirators are not recommended for
use against tritium oxide. See also footnote .2/, below, concerning
supplied-air suits and hoods.
4/ Under chin type only. Not recommended for use where it might be
possible for the ambient airborne concentration to reach instantaneous
values greater than 50 times the pertinent values in Appendix B,
Table I, Column 1 of 10 CFR, Part 20.
V Appropriate protection factor must be determined taking account
of the design of the suit or hood and its permeability to the
contaminant under conditions of use. No protection factor greater
than 1,000 shall be used, except as authorized by the Commission.
6/ No approval schedules currently available for this equipment.
Equipment must be evaluated by testing or on basis of available
test information.
3/ Only for shaven faces.
NOTE 1: Protection factors for respirators, as may be approved by the
U. S. Bureau of Mines , according to approved schedules for respirators
to protect against airborne radionuclides, may be used to the extent
that they do not exceed the protection factors listed in this Table. The
protection factors in this Table may not be appropriate to circumstances
when chemical or other respiratory hazards exist in addition to radioactive
hazards. The selection and use of respirators for such circumstances
7.5-8
should take into account approvals of the U. S. Bureau of Mines in
accordance with its applicable schedules.
NOTE 2: Radioactive contaminants for which the concentration
values in Appendix B, Table 1 of this part, are based on internal
dose due to inhalation may, in addition, present external exposure
hazards at higher concentrations. Under such circumstances, limitations
on occupancy may have to be governed by external dose limits.
• 7.6-1
7.6 REPORTING - ADMINISTRATIVE CONTROLS
Applicability
Applies to reports of plant operation required by the AEC.
Ob.t ective
To specify information which is required to be reported to the
AEC on a periodic basis.
Specification AC 7.6 - Reporting, Administrative Controls
Reports shall be submitted to the Director, Directorate of
Licensing, summarizing results of facility operations , including the
following:
SEMI-ANNUAL
a) Power Generation
A summary of power generated during the reporting period including:
1. Gross thermal power generated
2. Gross electrical power generated
3. Net electrical power generated
4. Number of hours the reactor was critical
5. Number of hours the generator was on line
6. Histogram of thermal power vs. time
b) Shutdowns
Descriptive material covering all outages occurring during the
reporting period. For each outage, information shall be
provided on:
1. The cause of the outage
2. The method of shutting down the reactor; e.g. , scram
L
7.6-2
or controlled deliberate shutdown.
3. Duration of the outage
4. Plant status during the outage; e.g. , cold shutdown or
hot shutdown.
5. Corrective action taken to prevent repetition, if appropriate.
c) Maintenance
Discussion of electrical, mechanical, and general maintenance
(except preventative maintenance) performed during the report
period and having potential effects on the safety of the facility.
The specific systems involved shall be identified and information
shall be provided on:
1. The nature of the maintenance (e.g. , routine, emergency, or
corrective).
2. The effect, if any, on the safe operation of the reactor.
3. The cause of any malfunction for which emergency or
corrective maintenance was required.
4., The results of any such malfunction.
5. The corrective and preventative action taken to preclude
recurrence.
6. The time required for completion.
d) Primary Coolant Chemistry
A tabulation on a monthly basis of the maximum, average, and minimum
values for the following primary coolant system parameters:
(i) Gross radioactivity in pCi/scc
(ii) Gross tritium in pCi/scc
(iii) Iodine 131 in pCi/scc
(iv) Ratio of Iodine - 131 to Iodine - 133
7.6-3
(v) Hydrogen ppm
(vi) Carbon Monoxide ppm
(vii) Carbon Dioxide ppm
(viii) Moisture (H20) ppm
(e) Occupational Personnel Radiation Exposure
(i) A tabulation of the number of occupational personnel exposures
for plant operations personnel (permanent and temporary) in the
following exposure increments for the reporting period: less
than 100 mrem, 100-250 mrem, 250-500 mrem, 500-750 mrem,
750-1000 mrem, 1-2 rem, 2-3 rem, 3-4 rem, 4-5 rem, 5-6 rem,
and greater than 6 rem.
(ii) A tabulation of the number of personnel receiving more than 500
mrem exposure in the reporting period according to duty function
[e.g. , routine plant surveillance and inspection (regular duty) ,
routing plant maintenance, special plant maintenance (describe
maintenance) , routine fueling operation, special refueling
operation (describe operation) , and other job-related exposures.]
(iii) A tabulation annually of the number of personnel receiving more
than 3 rem and the major cause(s) .
(f) Surveillance
A tabulation of results of surveillance as required by these
Technical Specifications.
(g) Radioactive Liquid Effluent Release (summarized on a monthly basis)
1. Total radioactivity (in curies) released, other than Tritium
and dissolved gases, and average concentration (in pCi/ml
above background) before dilution.
7.6- 4
2. Total Tritium and alpha radioactivity (in curies) released
and average concentration (in UCi/ml above background)
before dilution.
3. Total dissolved gas radioactivity (in curies) released
and average concentration (in TCi/ml above background)
before dilution.
4. Total volume (in gallons) of liquid waste discharged
before dilution.
5. Total volume (in gallons) of dilution water used.
6. Maximum concentration of total radioactivity (in UCi/ml
above background) other than tritium and dissolved gases
released in any single batch after dilution.
7. Estimated total radioactivity (in curies above background)
released, by nuclide, based on gamma isotopic analysis.
8. Percentage of MPC for total activity released, and the
MPC values used, calculated in accordance with the method
of Appendix B of 10CFR20.
h) Radioactive Gaseous Effluent Release (summarized on a monthly basis)
1. Total radioactivity released, excluding natural radioactivity,
(in curies) of:
a) Noble Gases
b) Tritium
c) Iodine
d) Particulates
e) Particulate Alpha Emitters
2. Maximum hourly average release rate (for any one hour
period) In curies/hr. above background.
7.6-5
3. Estimated total radioactivity (in curies above background)
released, by nuclide, based on the results of the required
gamma isotopic analysis.
4. Percentage of MPC and the MPC values used, calculated in
accordance with the method of Appendix B of 10 CFR 20:
a) Noble Gases
b) Tritium
c) Iodine
d) Particulates
5. Average meterological conditions during release periods,
including wind speed and relative frequency with which
wind was blowing from 16 directions for each batch release
from the radioactive gas waste storage system.
i) Solid Waste (Low Level)
1. Total volume (in cubic feet) of hydrogen getter adsorbent
shipped off site for disposal.
2. Curies of tritium involved.
3. Total volume (in cubic feet) of waste other than hydrogen
getter adsorbent shipped off-site for disposal.
4. Curies of miscellaneous waste involved.
5. Dates and disposition of material shipped off-site.
6. Total curies involved in off-site shipments.
j ) Environmental Monitoring (summarized on a quarterly, basis) .
1. Results of environmntal surveys performed during the
reporting period.
2. A map list of the list of the sampling locations, the total
number of samples, number of locations at which levels
7,6- 6
•
are significantly above local background and the highest,
lowest and average concentrations for the location with
greatest concentration, and designation of that point with
respect to the site.
3. Estimates of the likely resultant exposure to individuals
and to population groups and assumptions on which estimates
are made will be made if levels of radioactive materials in
the environment indicate the possibility of public intakes
in excess of 3% of those that could result from continuous
exposure to the concentration values listed in Appendix B,
Table II, Part 20.
k ) Facility Changes, Tests and Experiments
A summary description of changes in the facility or in procedures
C` and of tests and experiments carried out under the conditions of
Section 50.59(a) of 10 CFR 50. (Records shall be kept and a written
safety evaluation shall be made of all changes, tests and experiments
performed that do not require prior Commission approval. Under the
conditions of Section 50.59(b) , 10 CFR 50, a brief description of
the change and a summary of the safety evaluation will be submitted
to the Commission as a part of the Semiannual Report . )
Non-Routine Reports
a) Reporting of Abnormal Occurrence
In the event of an abnormal occurrence, a notification shall be made
within 24 hours by telephone and telegraph, to the Director,
Regional Regulatory Operations Office, followed by a written
report within 10 days to the Director, Directorate of Licensing,
C_-
with a copy to the Director, Regional Regulatory Operations Office.
7.6-7
The written report on abnormal occurrences, and to the extent
possible, the preliminary telephone and telegraph notification,
should: (a) describe, analyze, and evaluate safety implications,
(b) outline the measures taken to assure that the cause of the
condition is determined, and (c) indicate the corrective action
(including any changes made to the procedures and to the quality
assurance program) taken to prevent repetition of the occurrence
and of similar occurrences involving similar components or systems.
In addition, the written report should relate any failures or
degraded performance of systems and components for this incident
to similar equipment failures that may have previously occurred
at the facility. The evaluation of the safety implications of the
incident should consider the cumulative experience obtained from
the record of previous failures and malfunctions of the affected
systems and components or of similar equipment.
b) Off-Site Threats
Occurrences or conditions involving an off-site threat to the
safety of operation of the facility, including tornadoes, earthquakes,
flooding, repetitious aircraft overflights, attempted sabotage, or
civil disturbances shall be reported within 24 hours by telephone
and telegraph, to the Director, Regional Regulatory Operations Office,
followed by a written report within 10 days to the Director, Directorate
of Licensing, with a copy to the Director of the appropriate Regional
Compliance Office.
1
7.6-8
c) Reporting of Unusual Events
A written report shall be forwarded within 30 days to the Director,
Directorate of Licensing, and to the Director, Regional Regulatory
Operations Office, in the event of:
1. Discovery of any substantial errors in the transient or
accident analyses, or in the methods used for such analyses,
as described in the Safety Analysis Report or in the basis for
the Technical Specifications.
2. Any substantial variance in an unsafe direction of measured
values of thermal or nuclear characteristics , or of the
performance of a system or component from predicted
characteristics, or from performance specifications
contained in the Technical Specifications or in the
Safety Analysis Report.
3. Any condition involving a possible single failure which,
for a system designed against assumed single failures,
could result in a loss of the capability of the system to
perform its safety function.
d) Special Reports
1. Startup Report
A summary report of plant startup and power escalation testing,
and the evaluation of the results from these test programs ,
should be submitted for the newly licensed facilities and
modifications to an extent that the nuclear, thermal, or
hydraulic performance of the plant may significantly altered.
The test results should be compared with design predictions and
specifications. Startup reports should be submitted within
7.6-9
_ � 1
6o days following commencement of commercial power operation
(i.e. , following synchronization of the turbo-generator to
produce commercial power).
2. First Year Operation Report
A report should be submitted within 60 days after completion of
the first year of operation (this first year begins with the
synchronization of the turbo-generator to produce commercial
power). This report may be incorporated into the semiannual
operating report and should cover the following:
a. An evaluation of plant performance to date, in
comparison with design predictions and specifications;
b. A reassessment of the safety analysis submitted with
the license application in light of measured
operating characteristics when such measurements
indicate that there may be substantial variance
from prior analysis;
c. An assessment of the performance of vital equipment
as this performance relates to the safe operation of
the facility;
d. A progress and status report on any items identified
as requiring additional confirmatory information dui ng
the startup of the facility. This category includes
items addressed in the public safety evaluation, those items
that were established as conditions of the license and
those items identified in the startup report.
7.6- 10
Basis for Specification AC 7.6
The information specified to be reported periodically by this
Specification is adequate to document the operation of the plant
related to safety.
Hello