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- NUCLEAR REGULATORY COMMISSION
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June 5, 1984 .47-`7 '
-� In Reply Refer To: > ;;���r
Docket 50-267 ✓ , , , .4,";,
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/77 Mr. 0. R. Lee, Vice President c�` <C> ��
Electric Production OO<o. c/'
Public Service Company of Colorado
P.O. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
SUBJECT: FORT ST. VRAIN NUCLEAR GENERATING STATION, AMENDMENT NO. 43
TO FACILITY OPERATING LICENSE DPR-34
The Commission has issued the enclosed Amendment No. 43 to Facility Operating
License DPR-34 for the Fort St. Vrain Nuclear Generating Station in response
to your application for amendment dated March 14, 1984 (P-84067). This
application superseded, in its entirety, your April 19, 1983 (P-83148)
application.
The amendment revises the Technical Specifications to: 1) clarify the actions
to be taken when various instruments are inoperable; 2) correct typographical
and reference errors; 3) allow flexibility in the required moisture monitoring
instrumentation at low power levels; and 4) allow the use of previously
approved compensatory procedures during maintenance on the moisture monitors.
A copy of the Safety Evaluation supporting this amendment is also enclosed.
The notice of issuance will be included in the Commission' s next monthly
Federal Register notice.
Sincerely,
- , ' 0 SCI ,, ,
Philip C. Wagner, Project Manager
Reactor Project Branch 1
Enclosures:
1. Amendment No. 43 to DPR-34
2. Safety Evaluation
cc w/encis:
(See next page)
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Fort St. Vrain
cc list
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. 0. Box 840 Regional Representative
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
David Alberstein , 14/159A Denver, Colorado 80203
PA Technologies , Inc.
P. n. Box 85608 Don Warembourg
San Diego, CA 92138 Nuclear Production Manager
Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. 0. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O' Donnell
P. 0. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
,,Ns REGociUNITED STATES
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4,2
NUCLEAR REGULATORY COMMISSION
REGION IV
�Iff 611 RYAN PLAZA DRIVE, SUITE 1000
O 2� e40 ARLINGTON,TEXAS 76011
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PUBLIC SERVICE COMPANY OF COLORADO
DOCKET 50-267
FORT ST. VRAIN NUCLEAR GENERATING STATION
AMENDMENT TO FACILITY OPERATING LICENSE
Amendment No. 43
License DPR-34
1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Public Service Company of Colorado
(the licensee) dated March 14, 1984, 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;
B. The facility will operate in conformity with the application, the
provisions of the Act, and the rules and regulations of the
Commission;
C. There is reasonable assurance (i ) that the activities authorized by
this amendment can be conducted without endangering the health and
safety of the public, and (ii ) that such activities will be
conducted in compliance with the Commission's regulations;
D. The issuance of this amendment will not be inimical to the common
defense and security or to the health and safety of the public; and
E. The issuance of this amendment is in accordance with 10 CFR Part 51
of the Commission' s regulations and all applicable requirements have
been satisfied.
- 2 -
2. Accordingly, Facility Operating License DPR-34 is amended by changes
to the Technical Specifications as indicated in the attachment to this
license amendment, and paragraph 2.D. (2) is hereby amended to read as
follows:
(2) Technical Specifications
The Technical Specifications contained in Appendices A and B, as
revised through Amendment No. 43 , are hereby incorporated in the
license. The licensee shall operate the facility in accordance with
the Technical Specifications.
3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
JA Eric H. Johnson, Chief
0"` Reactor Project Branch 1
Attachment:
Changes to the Technical Specifications
Date of Issuance: June 5, 1984
ATTACHMENT TO LICENSE AMENDMENT
AMENDMENT NO. 43 TO FACILITY OPERATING LICENSE DPR-34
DOCKET 50-267
Replace the following pages of the Appendix A Technical Specifications with the
attached pages as indicated. The revised pages are identified by amendment
number and contain vertical lines indicating the areas of change.
Remove Insert
ii ii
4.4-1 4.4-1
4.4-2 4.4-2
4.4-8 4.4-8
4.4-8a ---
4.4-9 4.4-9*
4.4-10 4.4-10
4.4-11 4.4-11*
4.4-12 4.4-12*
4.4-13 4.4-13*
4.4-14 4.4-14
4.4-15 4.4-15*
4.4-16 4.4-16
4.4-17 4.4-17
4.4-18*
*page number change only
PAGE
4.2 PRIMARY COOLANT SYSTEM - LIMITING CONDITIONS
•
FOR OPERATION (Continued) .
•
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 - Fire Water System/Fire Suppression
Water System 4.2-4
Specification LCO 4.2.7 - PCRV Pressurization 4.2-5
Specification LCO 4.2.8 - Primary Coolant Activity 4.2-7
Specification LCO 4.2.9 — PCRV Closure Leakage 4.2-11
Specification LCO 4.2.10 — Loop Impurity Levels, High
Temperatures 4.2-13
Specification LCO 4.2.11 - Loop Impurity Levels, Low
Temperatures 4.2-13
Specification LCO 4.2.12 - Liquid Nitrogen Storage 4.2.15b
Specification LCO 4.2.13 - PCRV Liner Cooling System 4.2-15b
Specification LCO 4.2.14 - PCRV Liner Cooling Tubes 4.2-16
Specification LCO 4.2.15- PCRV Cooling Water System
Temperatures 4.2-17
Specification LCO 4.2-16 - DELETED
Specification LCO 4.2.17 - Diesel-Driven Generator for ACM 4.2-21
Specification LCO 4.2.18 - Primary Coolant Depressurization 4.2-21
Specification LCO 4.2.19 - Firewater Booster Pumps 4.2-22
•
4.3 SECONDARY REACTOR COOLANT SYSTEM - LIMITING CONDITIONS
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 4.3-2
Specification LCO 4.3.4 - Emergency Condensate and
Emergency Feedwater Headers 4.3-3
Specification LCO 4.3.5 - Storage Ponds 4.3-3
Specification LCO 4.3.6 - Instrument Air System 4.3-4
Specification LCO 4.3.7 - Hydraulic Power System 4.3-4
Specification LCO 4.3.8 - Secondary Coolant Activity 4.3-5
Specification LCO 4.3.9 - DELETED
Specification LCO 4.3.10 - Shock Suppressors (Snubbers) 4.3-7
4.4 INSTRUMENTATION AND CONTROL SYSTEMS - LIMITING
CONDITIONS FOR OPERATION 4.4-1
Specification LCO 4.4.1 - Plant Protective System
Instrumentation 4.4-1
Specification LCO 4.4.2 - Control Room Temperature 4.4-14
Specification LCO 4.4.3 - Area Radiation Monitors 4.4-14
Specification LCO 4. 4.4 — Seismic Instrumentation 4.4-16
Specification LCO 4.4.5 - Analytical System Primary Coolant
Moisture Instrumentation 4.4-16
Specification LCO 4.4.6 — Room Temperature, 480 Volt
Switchgear 4.4-1Q
>> Amendment No. 1$,2$ ,2$ , 43
Fort St. Vrain #1
Technical Specifications
Amendment #43
Page 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, except that to facilitate maintenance on the
Plant Protective System (PPS) moisture monitors, the
moisture monitor input trip functions to the Plant
Protective System which cause scram, loop shutdown,
circulator trip, and steam water dump may be disabled for
up to 72 hours. During the time that the Plant Protective
System moisture monitor trips are disabled, an observer in
direct communication with the reactor operator shall be
positioned in the control room in the location of
pertinent instrumentation. The observer shall
continuously monitor the primary coolant moisture levels
indicated by at least two moisture monitors and the
primary coolant pressure indications, and shall alert the
reactor operator to any indicated moisture or pressure
change.
For Table 4.4-2, the affected loop shall be shut down
within 12 hours.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-2
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.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 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 loss of voltage for no longer than 35
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. Failure
to trip the inoperable channel requires taking the appropriate
corrective action as listed on Pages 4.4-1 and 4.4-2 within the specified
time limit.
(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.
III. As stated in LCO 4.9.2
(j) Items la. , lc. , or id. accompanied by 2a. , 2b.; 2c. , or 2d. on
Table 4.4-2 are required for loop 1 shutdown. . Items lb. , le. or
if. , 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 bistables automatically reset at 10% after reactor
power is decreased from greater than 30%. The RWP may be manually
reset between 10% and 30% power.
(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.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-9
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.
(t) A primary coolant dew point moisture monitor shall not be considered
operable unless the following conditions are met:
1 ) Reactor Power Range Minimum Sample Flow
Startup to 2% 1 scc/sec.
>2% - 5% 5 scc/sec.
>5% - 20% 15 scc/sec.
>20% - 35% 30 scc/sec.
>35% - 100% 50 scc/sec.
2) Minimum flow of item 1 ) is alarmed in the control room and the alarm
is set in accordance with the power ranges specified.
3) The ambient temperatures indicated by both temporary thermocouples
mounted on the flow sensors in penetrations B1 and 83 are less than
185°F.
4) Fixed alarms of 1 scc/sec. and 75 scc/sec. are operable.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-10
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 actions 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 of the FSAR. High moisture
instrumentation is discussed in Section 7.3 of the FSAR.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-11
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) .
Hiah 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.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-12
Plant Electrical System Power Loss requires a scram to prevent any
power-to-flow mismatches from occurrino. 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 Buildins 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.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
page 4.4-13
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 charnel 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.
Fort St. Vrain #1
Technical Specifications
Amendment # 43
Page 4.4-14
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. 19, 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:
Fort St. Vrain
Technical Specification
Amendment # 43
Page 4.4-15
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 148614 Reactor Plant Exhaust Filter Room
1 RT-93252-2 14864 South Stairwell
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-93251-14 14816 Office Building
3 RT-93252-14 4829 Analytic Instrument Room
14 RT-93250-13 14791 Condensate Demineralizers
a RT-93250-5 4829 Main Control Room
•
RT-93251-6 14791 Grade Floor North
a RT-93252-6 14791 South Stairwell
5 RT-93251-5 4781 East Walkway
5 RT-93251-7 14781 Valve Operating Station - West
` RT-93252-7 14781 Valve Operating Station - East
6 RT-93250-8 14771 Northeast Walkway
6 RT-93251-8 14771 Radiochem Lab
6 RT-93251-9 4740 North Stairwell
Fort St. Vrain #1
Technical Specifications
Amendment #43
Page 4.4-16
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.
Specification LCO 4.4.5 - Analytical System Primary Coolant Moisture
Instrumentation - Limiting Condition for Operation
The reactor shall not be operated between a shutdown condition and
5% power during startup unless the primary coolant is being sampled by
two monitors, normally from the Analytical System.
If one of the two moisture monitors above becomes inoperable while
increasing reactor power between shutdown and 5%, a second monitor shall
be made operable or the reactor shall be shut down within 12 hours.
If all available moisture monitors become inoperable, during the
above-mentioned power increase, the reactor shall be shut down immediately.
During reactor power reduction from 5% power to shutdown conditions,
at least one moisture monitor must be in operation. If all available
moisture monitors become inoperable, the reactor shall be shut down
immediately.
Basis for Specification LCO 4.4.5
During reactor operation, primary coolant moisture monitors are
required below 5% reactor power for administration of LCO 4.2.11. One
moisture monitor is sufficient to detect primary coolant moisture content
on a continual basis.
Fort St. Vrain #1
Technical Specifications
Amendment #43
Page 4.4-17
Two analytical system moisture monitors will normally be in service
sampling primary coolant. These analytical moisture monitors do not
provide any automatic action (other than an alarm function) . Alternate
moisture monitors can also be placed in service sampling primary coolant,
such as through re-alignment of a moisture monitor in the analytical
system or utilization of operable (as defined in LCO 4.4.1 , Note (t)) plant
protective system dewpoint moisture monitors placed in the "indicate" mode
(note that in the "indicate" mode a trip is input to the PPS) . Operator
action is required to take corrective action in the event of high moisture
levels in the primary coolant in the shutdown to 5% reactor power range.
Operator reaction time to shut down the reactor in the event of
high moisture levels in the primary coolant system at reactor power
levels of 5% or less are acceptable. As indicated by Figure 4-2 in
Document GA-A13677, Test and Evaluation of the Fort St. Vrain Dew Point
Moisture Monitors System, one of the limiting parameters for determining
required response times to shut the reactor down in the event of high
primary coolant moisture is graphite oxidation. The allowable weight loss
of the hottest fuel element in the core is 1%.
At operating temperatures experienced at 5% reactor power, response
times to scram the reactor to limit oxidation to 1% by weight is
approximately 6700 seconds, well within the capabilities of an operator.
Fort St. Vrain #1
Technical Specifications
Amendment #43
Page 4.4-18
Specification LCO 4.4.6 - Room Temperature, 480 Volt Switchgear
The reactor shall not be operated at power if the 480 V switchgear
room temperature exceeds 120°F.
Basis for Specification LCO 4.4.6
The most limiting temperature in the 480 V switchgear room is
120°F. This limit is established to assure satisfactory operation
of safety-related control and electrical equipment located there
during reactor power operation.
,) ,pH HEG°<yr UNITED STATES
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NUCLEAR REGULATORY COMMISSION
N3 REGION IV
{r�
611 RYAN PLAZA DRIVE. SUITE 1000
' ARLINGTON, TEXAS 76011
SAFETY EVALUATION BY THE NUCLEAR REGULATORY COMMISSION
RELATED TO AMENDMENT NO. 43 TO FACILITY OPERATING LICENSE DPR-34
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET 50-267
INTRODUCTION.
By letter dated April 19, 1983, Public Service Company of Colorado (PSC or
the licensee) submitted an application to amend the Fort St. Vrain (FSV)
Technical Specifications (TS). The proposed changes would: 1) clarify a
footnote for the reactor protection system (RPS) instrumentation tables
(Tables 4.4-1 through 4.4-4 of LCO 4.4.1) to remove confusion, and 2) allow
the use of alternate moisture monitoring instrumentation during low power
operation (LCO 4.4.5) . Following a detailed review of the submitted TS
pages, we recommend some corrections and changes to clarify requirements
and facilitate future operations. The licensee agreed with our recommenda-
tions and superseded that application with a new application dated March 14,
1984. This new application, in addition to including the above changes
and correcting some editorial problems , provides for the use of temporary,
compensatory measures to allow maintenance on the moisture monitoring
instrumentation.
EVALUATION
The licensee's March 14, 1984 application proposed the following four areas
of change to the FSV TS dealing with instrumentation and control systems:
1. Editorial Changes
PSC proposed a correction to a typographical error in footnote (j ) of
Table 4.4-1 and to correct the FSAR references cited in LCOs 4.4. 1 and
4.4.2. These changes were reviewed and found to be acceptable,
editorial corrections.
Additional editorial changes were required because page numbers of
some proposed TS pages, submitted with the application, were inconsistent
with the existing page numbering. We corrected this problem by
renumbering the existing pages in an effort to improve the formatting
of this section of the TS. We also corrected the numbering of LCO
4.4.3.
- 2 -
2. Moisture Monitor Maintenance
PSC proposed the use of compensatory measures in the form of a
dedicated observer monitoring alternate moisture monitoring
instrumentation and primary coolant pressure instrumentation while
allowing the reactor protective system (RPS) moisture monitors
(dew-point monitors) to be out of service for corrective maintenance.
The proposed compensating provisions are similar to the operating
condition allowed by LCO 4.9.2, "Plant Protective System Dew Point
Moisture Monitoring Tests During Phase 2." (These tests involve
the injection of moisture-laden gas into the primary coolant to verify
the proper operation of the monitoring instruments. ) In addition,
a temporary (10-day) change was made to the TS by Amendment 31, which
was issued on January 20, 1983, that allowed continued plant
operation during maintenance of the dew-point monitors provided
compensatory measures similar to those proposed were taken. Since
this change is in accordance with previously approved conditions of
operation, is limited to short periods of time (72 hours) to allow
maintenance on the monitors, and acceptable levels of protection will
be provided by the required compensatory measures during the periods
of inoperability, we find it to be acceptable.
3. Clarification of footnote (f)
Footnote (f) of Table 4.4-1 states, "The inoperable channel must be
in the tripped condition, unless the trip of the channel will cause
the protective action to occur." This statement had been misinterpreted
by operating personnel as allowing continued plant operation without
the required instrumentation and was reported in Reportable Occurrence
83-001 which was transmitted by letter dated January 17, 1983. In an
effort to remove any confusion, PSC proposed some additional wording
in the April 19, 1983 application. Following discussions with the
NRC Project Manager, PSC agreed to further clarify the intent of the
footnote and provided that clarification in this application.
We have reviewed this change and have determined that it will better
define the correct operator action in the event of out-of-service
instrumentation and, therefore, find it to be an acceptable
administrative change.
4. Low Power Moisture Monitoring
The present LCO 4.4.5, "Analytical System Primary Coolant Moisture
Instrumentation," is worded in such a way that only two of the
analytical monitors were considered to be acceptable in fulfilling
the requirements. This condition is discussed in IE Inspection
Report No. 82-31, dated January 21, 1983, and is considered an
unresolved item (6231-01) . PSC proposed a change to LCO 4.4.5, in
accordance with the agreement stated in the above Inspection Report,
- 3 -
which allows the use of alternate moisture monitoring instrumentation.
The alternate monitors can be either the analytical instrument
installed in the analytical instrumentation panel or the RPS dew-
point monitors.
We have reviewed this change and found it acceptable because the
moisture limitations are not being changed and acceptable levels
of moisture monitoring will be provided to ensure that moisture
levels can be adequately determined to comply with the limitations.
ENVIRONMENTAL CONSIDERATION
We have determined that the amendment does not authorize a change in effluent
types or total amounts nor an increase in power level and will not result in
any significant environmental impact. Having made this determination, we have
further concluded that the amendment involves an action which is insignificant
from the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4) ,
that an environmental impact statement or negative declaration and environmental
impact appraisal need not be prepared in connection with the issuance of this
amendment.
CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public
will not be endangered by operation in the proposed manner, and (2) such
activities will be conducted in compliance with the Commission's regulations
and the issuance of this amendment will not be inimical to the common
defense and security or to the health and safety of the public.
Dated: June 5, 1984
The following NRC personnel have contributed to this Safety Evaluation:
Philip C. Wagner
PR R E:.c.
UNITED STATES
_ NUCLEAR REGULATORY COMMISSION
REGION IV
All, 611 RYAN PLAZA DRIVE. SUITE 1000
� M1c
ARLINGTON, TEXAS 76011
. .t o
June 4, 1984 !0
In Reply Refer To: Q *IPA
Docket 50-267 4 t ,
Mr. 0. R. Lee, Vice President I 1
Electric
Public Service
eCompany of Colorado tion �<F
P.O. Box 840 Oc N
Denver, Colorado 80201 �/
Dear Mr. Lee:
SUBJECT: FORT ST. VRAIN NUCLEAR GENERATING STATION, AMENDMENT NO. 42
TO FACILITY OPERATING LICENSE DPR-34
The Commission has issued the enclosed Amendment No. 42 to Facility Operating
License DPR-34 for the Fort St. Vrain Nuclear Generating Station in response
to your application for amendment dated March 6, 1984 (P-84068).
The amendment revises the Administrative Controls Technical Specifications
to incorporate the reporting requirements of 10 CFR Parts 50.72 and 50.73.
A copy of the Safety Evaluation supporting this amendment is also enclosed.
The notice of issuance will be included in the Commission's next monthly
Federal Register notice.
Sincerely,
e—afkif o Li 61
Philip C. Wagner, Project Manager
Reactor Project Branch 1
Enclosures:
1. Amendment No. 42 to DPR-34
2. Safety Evaluation
cc w/encls:
(See next page)
LI I/cc: - kg d'-i
Fort St. Vrain
cc list
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. 0. Box 840 Regional Representative
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
David Alberstein, 14/159A.
Denver, Colorado 80203
GA Technologies , Inc.
P. 0. Box 85608 Don Warembourg
San Diego, CA 92138 Nuclear Production Manager
Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. 0. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell
P. 0. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
UNITED STATES
NUCLEAR REGULATORY COMMISSION
'L-31:=7
REGION IV
611 RYAN PLAZA DRIVE, SUITE 1000
ARLINGTON, TEXAS 76011
**'.*
PUBLIC SERVICE COMPANY OF COLORADO
DOCKET 50-267
FORT ST. VRAIN NUCLEAR GENERATING STATION
AMENDMENT TO FACILITY OPERATING LICENSE
Amendment No. 42
License DPR-34
1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Public Service Company of Colorado
(the licensee) dated March 6, 1984, 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;
B. The facility will operate in conformity with the application, the
provisions of the Act, and the rules and regulations of the
Commission;
C. There is reasonable assurance (i ) that the activities authorized by
this amendment can be conducted without endangering the health and
safety of the public, and (ii ) that such activities will be
conducted in compliance with the Commission's regulations;
D. The issuance of this amendment will not be inimical to the common
defense and security or to the health and safety of the public; and
E. The issuance of this amendment is in accordance with 10 CFR Part 51
of the Commission's regulations and all applicable requirements have
been satisfied.
- 2 -
2. Accordingly, Facility Operating License DPR-34 is amended by changes
to the Technical Specifications as indicated in the attachment to this
license amendment, and paragraph 2.D. (2) is hereby amended to read as
follows:
(2) Technical Specifications
The Technical Specifications contained in Appendices A and B, as
revised through Amendment No. 42 , are hereby incorporated in the
license. The licensee shall operate the facility in accordance with
the Technical Specifications.
3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
j/1 Eric H. Johnson, Chief
0 Reactor Project Branch 1
Attachment:
Changes to the Technical Specifications
Date of Issuance: June 4, 1984
ATTACHMENT TO LICENSE AMENDMENT
AMENDMENT NO. 42 TO FACILITY OPERATING LICENSE DPR-34
DOCKET 50-267
Replace the following pages of the Appendix A Technical Specifications with the
attached pages as indicated. The revised pages are identified by amendment
number and contain vertical lines indicating the areas of change.
Remove Insert
7.1-10 7.1-10
7. 1-16 7. 1-16
7.3-2 7.3-2
7.5-12 - 7.2-25 7.5-12 - 7.5-18
port St. v^aln pi
Tec cal Specifications
Amenament No. 42
Page 7.1-10
a. Review of all procedures required by Technical
Specification 7.4(a) , (b) , and (c) and changes
thereto, and any other proposed procedure or changes
to approved procedures as determined by the Station
Manager to affect nuclear safety.
b. Review of all proposed tests and experiments that
affect nuclear safety.
c. Review of all proposed changes to the Technical
Specifications.
d. Review of all proposed changes or modifications to
plant systems or equipment that affect nuclear
safety.
e. Investigation of all violations of the Technical
Specifications including the preparation and
forwarding of reports covering the evaluation and
recommendations to prevent recurrence to the
Manager, Nuclear Production and to the Chairman of
the Nuclear Facility Safety Committee.
f. Review of all Reportable Events.
g. Review of facility operations to detect potential
nuclear safety hazards.
-ort t. i'ra' n F:
Tec, cal Specifications
Amendment No. 42
Page 7. 1-16
(7) All Reportable Events.
(8) Any indication that there may be a deficiency in
some aspect of design or operation of
structures, systems, or components, that affect
nuclear safety.
(9) Reports and meeting minutes of the PORC.
b. Audits of facility activities shall be performed
under the cognizance of the Nuclear Facility Safety
Committee. These audits shall encompass:
(1) The conformance of facility operation to all
provisions contained within the Technical
Specifications and applicable license conditions
at least once per year.
(2) The performance, training, and qualifications,
of the facility staff at least once per year.
(3) The results of actions taken to correct
deficiencies occurring in facility equipment,
structures, systems, or method of operation that
affect nuclear safety at least once per six
months.
• Fort St. Vrain r_
Tec cal Specifications
Amendment No. 42
Page 7.3-2
3) Licensee Event Reports (LER).
4) Records of surveillance activities,
inspections and calibrations required by these
Technical Specifications.
5) Records of reactor tests and experiments.
6) Records of changes made to Operating
Procedures.
7) Records of radioactive shipments.
8) Records of sealed source leak tests and
results.
9) Records of annual physical inventory of all
source material of record.
b) The following records shall be retained for the
duration of the Facility Operating License:
1) Record and drawing changes reflecting facility
design modifications made to systems and
equipment described in the Final Safety
Analysis Report.
2) Records of new and irradiated fuel inventory,
fuel transfers and assembly burnup histories.
3) Records of facility radiation and
contamination surveys.
Fort St. Vrain F1
Ted cal Specifications
Amendment No. 42
Page 7.5-12
the license application and amendments
thereto;
5. An evaluation of the change, which shows
the expected maximum exposures to
individuals in the unrestricted area and
to the general population that differ
from those previously estimated in the
license application and amendments
thereto;
6. A comparison of the predicted releases of
radioactive materials, in liquid and
gaseous effluents and in solid waste, to
the actual releases for the period prior
to when the changes are to be made;
7. An estimate of the exposure to plant
operating personnel as a result of the
change; and
8. Documentation of the fact that the change
was reviewed and found acceptable by the
Plant Operations Review Committee.
7.5.2 Reportable Events
a) Notification Requirements
1 The NRC shall be notified pursuant to the
conditions and requirements of 10 CFR 50.72.
-ort. St. v-al.
Tec cal Specifications
Amendment No. 42
Page 7.5-13
b) Licensee Event Reports (LER)
Licensee Event Reports will be submitted to
the NRC pursuant to the conditions and
requirements of 10 CFR 50.73.
7.5.3 Non-Routine Radiological Reports
a. Radioactive Gaseous Effluent
1. If the calculated dose from the release
of gaseous effluents pursuant to
ESR 8.1.1. 1 ) exceeds any of the limits in
ELCO 8.1.1.h) , in lieu of a Licensee
Event Report, a special report that
identifies the cause(s) for exceeding the
limit and defines the corrective actions
that have been taken to reduce the
releases and the proposed corrective
actions to be taken to assure that
subsequent releases will be in compliance
with the above limits will be prepared
and submitted to the NRC within 30 days.
Fort St. Drain F=
Tec, cal Specifications
Amenament No. 42
Page 7.5-14
2. If gaseous waste is discharged without
treatment and in excess of the limits, in
lieu of a Licensee Event Report, a
special report that includes the
following information shall be prepared
and submitted to the NRC within 30 days:
(a) Explanation of why gaseous radwaste
was being discharged without
treatment, identification of any
inoperable equipment or subsystems,
and the reason for the
inoperability,
(b) Action(s) taken to restore the
inoperable equipment to operable
status, and
(c) Summary description of action(s)
taken to prevent a recurrence.
b. Radioactive Liquid Effluent
1. If the calculated dose from the release
of radioactive materials in liquid
effluents pursuant to ESR 8.1.2.e)
exceeds any of the limits specified in
ELCO 8.1.2.9), in lieu of a Licensee
Event Report, a special report that
identifies the cause(s) for exceeding the
=ort St. Vrain #1
Tecl cal Specifications
Amendment No. 42
Page 7.5-15
limit(s) and defines the corrective
actions that have been taken to reduce
the releases and the proposed corrective
actions to be taken to assure that
subsequent releases will be in compliance
with the above limits will be prepared
and submitted to the NRC within 30 days.
2. If radioactive liquid waste is discharged
without treatment pursuant to
ELCO 8. 1.2.h) , and in excess of the
limits, in lieu of a Licensee Event
Report, a special report that includes
the following information shall be
prepared and submitted to the NRC within
30 days:
(a) Explanation of why liquid radwaste
was being discharged without
treatment, identification of any
inoperable equipment or subsystems,
and the reason for the
inoperability,
(b) Action(s) taken to restore the
inoperable equipment to operable
status, and
Fort t. Vrain #1
TecL cal Specifications
Amendment No. 42
Page 7.5-16
(c) Summary description of action(s)
taken to prevent a recurrence.
c. Radioactive Effluents - Total Dose
1. If the limits of ELCO 8. 1.5.a) have been
exceeded, in lieu of a Licensee Event
Report, a special report that defines the
corrective action to be taken to reduce
subsequent releases to prevent recurrence
of exceeding the above limits and
includes the schedule fof achieving
conformance with the above limits shall
be prepared and submitted to the NRC
within 30 days. This special report, as
defined in 10CFR Part 20.405c, shall
include an analysis that estimates the
radiation exposure (dose) to a member of
the public from uranium fuel cycle
sources, including all effluent pathways
and direct radiation, for the calendar
year that includes the release(s) covered
by this report. It shall also describe
levels of radiation and concentrations of
radioactive material involved, and the
cause of the exposure levels or
concentrations. If the estimated doses)
exceeds the above limits, and if the
Fort ct. Vrain F_
Tech :al Specifications
Amendment No. 42
Page 7.5-17
release condition resulting in violation
of 40CFR Part 190 has not already been
corrected, the special report shall
include a request for a variance in
accordance with the provisions of 40CFR
Part 190. Submittal of the report is
considered a timely request, and a
variance is granted until staff action on
the request is complete.
d. Radiological Environmental Monitoring
1. If the level of radioactivity as a result
of plant effluents in an environmental
sample medium at a specified location
exceeds the reporting levels of
Table 8.2-3 of ELCO 8.2.1, when averaged
over any calendar quarter, in lieu of a
Licensee Event Report, pursuant to
Specification ELCO 8.2.1.c), a special
report that identifies the cause(s) for
exceeding the limit(s) and defines the
corrective actions to be taken to reduce
radioactive effluents such that the
potential annual dose to a member of the
public is less than the calendar year
limits of Specifications ELCO 8.1.1.h)
and ELCO 8.1.2.g) will be prepared and
Port ct. Vrain #1
Tech :al Specifications
Amendment No. 42
Page 7.5-18
submitted to the NRC within 30 days.
When more than one of the radionuclides
in Table 8.2-3 are detected in the
sampling medium, this report shall be
submitted if:
Concentration (1) Concentration (2)
Reporting Level (1) + Reporting Level (2) + . . . ≥ 1.0
When radionuclides other than those in
Table 8.2-3 are detected and are the
result of plant effluents, a report shall
be submitted if the potential annual dose
to a member of the public is equal to or
greater than the calendar year limits of
Specifications ELCO 8. 1.1. i) and
ELCO 8.1.2.g). This report is not
required if the measured level of
radioactivity was not the result of plant
effluents; however, in such an event, the
condition shall be reported and described
in the Annual Radiological Environmental
Monitoring Report.
�pfl gec v<
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION IV
S Y., 1141L it
-., 611 RYAN PLAZA DRIVE, SUITE 1000
ARLINGTON,TEXAS 76011
SAFETY EVALUATION BY THE NUCLEAR REGULATORY COMMISSION
RELATED TO AMENDMENT NO. 42 TO FACILITY OPERATING LICENSE DPR-34
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET 50-267
INTRODUCTION
The NRC revised the regulations for reactor plant reporting requirements by
changes to 10 CFR Part 50.72, concerning immediate notification requirements,
and by incorporating a new Part 50.73, concerning the Licensee Event Report
system. These revisions became effective on January 1, 1984, and were the
subject of a letter to all Power Reactor Licensees (Generic Letter 83-43)
dated December 19, 1983. This letter requested all licensees to propose
revisions to their plant's Technical Specifications (TS) to incorporate the
guidance contained therein for conformance with the revised regulations.
Public Service Company of Colorado (PSC) responded to that request by
application dated March 6, 1984 (P-84068) for the Fort St. Vrain Nuclear
Generating Station (FSV).
EVALUATION
The March 6, 1984 application was reviewed against the staff positions con-
tained in Generic Letter 83-43 and found to be in conformance with that
guidance. Since no definition presently exists in the FSV TS and since the
term "Reportable Event" is defined in the revised regulations (10 CFR 50.73) ,
no new definition was added. Also, all Reportable Events will be reviewed
by the Nuclear Facility Safety Committee thereby exceeding the proposed
guidance making this requirement more conservative than the guidance.
Since the application incorporates the NRC-provided guidance in order to
ensure compliance with recently revised regulations and since these
changes are administrative in nature, we find the changes to be acceptable.
ENVIRONMENTAL CONSIDERATION
We have determined that the amendment does not authorize a change in effluent
types or total amounts nor an increase in power level and will not result in
any significant environmental impact. Having made this determination, we have
further concluded that the amendment involves an action which is insignificant
from the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4) ,
that an environmental impact statement or negative declaration and environmental
impact appraisal need not be prepared in connection with the issuance of this
amendment.
•
- 2 -
CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public
will not be endangered by operation in the proposed manner, and (2) such
activities will be conducted in compliance with the Commission's regulations
and the issuance of this amendment will not be inimical to the common
defense and security or to the health and safety of the public.
Dated: June 4, 1984
The following NRC personnel have contributed to this Safety Evaluation:
Philip C. Wagner
Hello