HomeMy WebLinkAbout851193.tiff rJC`EOP RECV(,r UNITED STATES
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'"4 NUCLEAR REGULATORY COMMISSION
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i; REGION IV
Ill RYAN PLAZA DRIVE,SUITE 1000
°cs� N°‘� ARLINGTON,TEXAS 78011
AUG 2 8 1985
Docket: 50-267
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Public Service Company of Colorado }
1
ATTN: M. Holmes, Licensing Manager SEP 3 1985 I
P. 0. Box 840
Denver, Colorado 80201-0840 GzeeLer. cow.
Gentlemen:
As we discussed during the meeting on Technical Specifications (TSs) at the
Fort St. Vrain Station on July 26, 1985, and during the telephone conference
on August 13, 1985, it is our intention to include the review and approval of
some of the applications for proposed changes to the TSs as part of the Upgrade
Program. Attached as Enclosures 1 and 2 are the lists of applications which
we intend to review as part of the Upgrade Program and those we intend to
review separately. It is our present opinion that the applications included
in Enclosure 1 are not completely acceptable and would require some
modification; the comparable requirements in the Upgrade Program draft,
however, appear to be more acceptable. Therefore, it appears that a more
efficient use of your and our resources can be achieved by incorporating
certain reviews into the Upgrade Program.
I would appreciate your review of the enclosures to ensure the accuracy and
completeness of the pending applications and any comments you may have on our
approach.
Sincerely,
yLae C
Philip C. Wagner
Senior Project Manager
Enclosures:
As stated
cc:
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
851193
-2-
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. O. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. O. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 19}
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
Enclosure 1
ACTIONS TO BE INCLUDED IN THE TECHNICAL
SPECIFICATION UPGRADE PROGRAM
1. Electrical Technical Specifications (TAC 51955)
- The original application (P-83415) dated 12/30/83, which was
submitted in response to Safety Evaluations dated 8/17/82 and
10/12/82 related to degraded grid and onsite power systems
undervoltage protection, was superseded by a rewrite of the
electrical systems section of the TSs in response to a 10/12/83
enforcement conference commitment. The 6/26/84 (P-84187)
reapplication did not contain all of the requested upgrades
and was further revised by a 2/6/85 (P-85041) application.
(This application also responds to Generic Letter 84-15
(TAC 55861) for diesel generator reliability.)
- Since the latest submittal is not completely acceptable and since
recent NRC positions of diesel generator testing appear to be more
acceptable to PSC, the needed revisions will be included in the
Upgrade Program and no further action will be taken on the submitted
applications.
- The recently completed modifications to the AC and DC electrical
systems are being reviewed for acceptability as a separate issue.
2. NUREG-0737 Technical Specifications (TAC 54535)
- The application dated 7/31/84 (P-84242) was submitted in response to
Generic Letters 83-36 and 37. Review of the Upgrade Program draft
indicates that the requirements contained therein more closely follow
the guidance contained in the Generic Letters than the proposed
application. Therefore, inclusion of this review in the Upgrade
Program seems to be appropriate.
3. Circulator Operability Requirements (TAC 56927)
- The application dated 2/6/85 (P-85042) would require numerous
changes to be found acceptable. The requirements proposed in the
Upgrade Program are being revised as a result of the meetings
conducted from July 22 to 26, 1985. Therefore, efficiency would
indicate that his application should be included in the Upgrade
Program.
4. Restart Requirements
- The tendon wire surveillance requirements approved in the July 8,
1985 Safety Evaluation will be included in the Upgrade Program as
committed to by PSC.
- The revised interim Control Rod Drive Mechanism and Reactivity
Control Technical Specifications approved in Safety Evaluations
dated July 12, 1985 and July 23, 1985, will be included in the
Upgrade Program as committed to by PSC.
Enclosure 2
ACTIONS BEING REVIEWED INDEPENDENTLY
1 Instrument Setpoints (TAC 47416)
- Problems discovered in the methods used to establish instrument
setpoints resulted in a 10/1/80 applications which was discussed at
a meeting in Denver on 10/27/83 during which PSC agreed to a
resubmittal . The reapplication provided by letter dated 6/21/85
(P-85214) is being reviewed by NRR under TIA 83-15.
- The recent application contains some accident reanalyses which will
require review in addition to the instrumentation review. Present
plans call for completing our review in a time frame to allow
incorporation into the Upgrade Program, however, due to the
magnitude of this review effort, a contingency plan to use existing
values in the Upgrade at the time of issuance should be developed.
2. L.C.O. 4.1.9 (TAC 52634)
- Following the discovery of nonconservative assumptions in establishing
these power-to-flow requirements 10/11/83, PSC submitted an LER com-
mitting to follow conservative guidance and provide a corrected require-
ment. An application dated 12/15/83 (P-83403) was reviewed and
determined to be unacceptable. Meeting between us, our consultant at
ORNL and PSC (see summary dated 9/6/84) have not resulted in a revised
application. ORNL guidance on 7/17/85 and PSC submittal 8/1/85
(P-85271) indicate progress is being made.
- This issue should be resolved in the near future with the result
then included in the Upgrade Program. It may be appropriate for
another meeting among all involved parties to reach final resolution
prior to the commitment of including the new requirements in the
10/15/85 Upgrade submittal .
3. ISI/IST Requirements (TAC 53417)
- The 12/30/83 (P-83416) application was submitted in response to
commitments to include appropriate requirements in the TSs. The
preliminary results of our review were discussed during site meetings
on July 22 to 26, 1985. The results of our review and the meeting
discussions will be transmitted to PSC in the near future.
- We have determined that these inspection and testing requirements
should not be delayed until completion of the Upgrade Program and we
will review the resubmittal PSC committed to submit within 90 days
of their receipt of our evaluation as a separate issue.
4. Circulator Overspeed Trip Setpoint (TAC 55052)
- The application dated 5/10/84 (P-84137) contained insufficient justifi-
cation to allow a compete review. A response to the 10/25/84 questions
was submitted by letter dated 12/27/84 (P-84537). A request for NRR
review assistance was made by memo (Denise to Eisenhut) dated 1/28/85.
-2-
- Since this change will have little effect on the Uprade Program, the
review is being handled separately.
5. Organizational Changes (TAC 56743)
- The 1/14/85 (P-85009) application was determined to be obsolete
prior to the expiration of the notice period and a revised set of
requirements was submitted by application dated 6/10/85 (P-85155).
The revised application, however, includes a reference to the
Fuel Surveillance Program which was submitted by letter dated 5/3/85
(P-85151).
- Prior to the approval of the revised application, it will be
necessary for NRC to accept the Fuel Surveillance Program as
incorporating all of the past commitments and agreements on this
subject. This review should be completed in the near future so the
organizational changes can be approved and issued to reestablish
compliance with the TS.
6. Xenon Stability Testing (TAC 57788)
- In order to complete the initial rise-to-power testing program item
on xenon stability, it is necessary to allow a temporary change
to the TS requirement on control rod position. A special test
exception type change was proposed on 5/22/85 (P-85179).
- In order to allow completion of the test program during Cycle 4
(see P-85063 dated 3/5/85 for the PSC commitment to complete these
tests) this change needs to be processed as a separate issue.
UNITED STATES WELD GOIivTY Cpbmk7lSe!n,-;
DEPARTMENT OF THE INTERIOR Q ^ `! -'7a17'
BUREAU OF RECLAMATION
LOWER MISSOURI REGION SEP 3 1985
UPPER COLORADO REGION
'GREELEY, COLO.
NOTICE OF INITIATION OF INVESTIGATIONS
Name of Investigation: High Mountain Aquifer Study, Colorado
Type of Investigation: Special
Location of Investigation: Scattered throughout the high mountain areas of
the State of Colorado are 71 potential glacial
• aquifer sites. For the most part these sites
are in close proximity to the Continental
Divide.
Date Investigation Initiated: Investigation will be initiated in fiscal
year 1985. The study will culminate in January of 1990 with a Special
Report which will present the study findings.
1. Scope of Investigation: The primary purpose of this study is to
determine the possibility of using the glacial aquifers as storage reser-
voirs. If these basins can be used as storage reservoirs, they would
collect water from snowmelt and rainfall and release it when downstream
demands are high. Two advantages anticipated from using glacial aquifers
are a reduction in evaporation losses and a savings of capital cost for
constructing surface storage reservoirs. Environmental and social impacts
of aquifer storage may also prove to be fewer than those associated with
surface storage reservoirs.
This study will evaluate the technical aspects of using these glacial
aquifers as storage reservoirs. It is not intended to reach full-scale
development of any high mountain aquifer. Hydrologic studies, engineering
cost estimates, and economic and environmental evaluations will be made to
determine if these sites compare favorably with other sources of water
supplies and storage. The study will also look at the legal constraints
involved and their effect on this type of development. The resolving of
state water rights issues will not be a part of this study effort.
2. Problem: The water supplies in Colorado for irrigation, municipal , and
industrial uses are becoming more scarce and expensive to develop. The use
of high mountain glacial aquifers as storage reservoirs may prove to be a
more economical , efficient, and environmentally acceptable way to develop
and manage Colorado' s water resources.
3. Prospective Solutions: The 71 potential sites are mostly found above
the 8500 foot elevation and have an average elevation of just over 9000
feet. They vary in size from about 100 to 2,000 acres. As a storage
facility, the basins would capture water at a time of high flow and release
water when a dependable water supply is needed downstream. These glacial
aquifers may also be developed for replacement and/or compensatory storage
for the movement of water from one tributary basin to another. These water
transfers could be entirely on one side of the Continental Divide, or
involve a transmountain diversion.
4. Participation: The Bureau of Reclamation and the CORY Company, Inc. , a
private entity, signed a contract on August 21, 1984, to share the cost of
this investigation. Other interested entities include the State of
Colorado and the Colorado delegation, the Colorado Water Conservation
Board, the Colorado State Engineer, the Colorado River Water Conservation
District, and the West Divide Water Conservancy District. The Forest
Service and the Fish and Wildlife Service have been notified of upcoming
study activities. Meetings have been held with 10 of the major water orga-
nizations in the State to advise them of the proposed investigations and to
exchange information.
5. Indication of interest: Expressions of interest should be sent to:
East Slope West Slope
Bureau of Reclamation Bureau of Reclamation
Lower Missouri Region Upper Colorado Region
PO Box 25247 PO Box 2553
Building 20, Denver Federal Center 125 S. State Street
Denver, CO 80225-0247 Salt Lake City, UT 84147
(Attention: William J. Steele (Attention: Harl Noble
Regional Planning Officer Regional Planning Officer
303-236-0489) 801-524-5517)
Prepared: August 1 , 1985 Transmitted: August 30, 1985
. % a
, 27/-
Regional Plann�hg Officer Regional Planning 0 ficer
Lower Missouri egion Upper Colorado Region
ACTINCaRe iol Director `� ds!?alte
g � eg onal Directo
Lower Missouri Region Upper Colorado Region
SSINS No. : 6835
IN 85-75
W'E�LDD COUNTY COMMISS!us
UNITED STATES
NUCLEAR REGULATORY COMMISSION o E 4EIn c1L c
fa
OFFICE OF INSPECTION AND ENFORCEM� � �!
WASHINGTON, D.C. 20555 SEP 5 1985
' Li
August 30, 1985
IE INFORMATION NOTICE NO. 85-75: IMPROPERLY INSTALLED INSTRUMENTATION,
INADEQUATE QUALITY CONTROL AND INADEQUATE
POSTMODIFICATION TESTING
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is to alert addressees of two recent instances of
improper system modifications, inadequate quality control and inadequate post-
modification testing following installation of environmentally qualified
equipment. Recipients are expected to review the information for applicability
to their facilities and consider actions, if appropriate, to preclude similar
problems occurring at their facilities. However, suggestions contained in this
information notice do not constitute NRC requirements; therefore, no specific
action or written response is required.
Description of Circumstances:
LaSalle Unit 2
On June 10, 1985, at 11: 30 a.m. , the licensee informed the NRC Resident
Inspector that for approximately 5 days LaSalle Unit 2 had been without the
capability of automatic actuation of emergency core cooling (ECCS) and that
for approximately 3 days during this period the plant had been without secondary
containment integrity. The major cause of this condition was improper installation
(the variable and reference legs were reversed) of the two reactor vessel level
actuation switches which control Division I automatic depressurization system
(ADS) , low pressure core spray (LPCS), and reactor core isolation cooling
(RCIC).
Unit 2 was shut down in February 1985 for an outage that included installation
of environmentally qualified electrical equipment. LaSalle has three divisions
of ECCS equipment. In March 1985, ECCS Division III was taken out of service
for maintenance. On June 5, 1985, ECCS Division II was taken out of service
for modifications. On June 3, 1985, secondary containment was declared inoperable
for maintenance on the reactor building ventilation system. The result of
these scheduled actions was that two of three ECCS divisions and secondary
containment were inoperable, leaving ECCS Division I available for use.
Subsequently, it was discovered that the variable and reference legs to the
8508270286
IN 85-75
August 30, 1985
Page 2 of 3
reactor vessel level actuation switches for ECCS Division I had been accidentally
reversed since June 3, 1985; thus leaving the plant with no ECCS automatic
actuation and no secondary containment.
The cause of the piping reversal was initially the result of incorrect design
drawings which were released to the contractor on April 1, 1985. The licensee's
site personnel recognized the error on April 4, 1985, and issued a Field Change
Request to correct it. However, the isometric drawings being used at the
location of the modification activities were not corrected. Therefore, the
contractor proceeded to connect piping in the reverse order from the correct
configuration. Figure 1 shows the correct configuration and Figure 2 shows the
reversal . This error was not identified by the Quality Control (QC) Program
because the contractor' s QC did not assign inspection hold points for either
the electrical or mechanical piping connections for any of the 22 instruments
replaced by the modification. Consequently, the installation adequacy was not
verified against the design drawings, which did include the field change and,
therefore, which could reasonably be expected to have revealed the error in the
two instruments that were piped backwards.
Subsequent postmodification testing failed to detect the error because (as
shown in Figure 3) the test shut the instrument block isolation valves and
injected a test pressure source through the installed test connections
downstream from the instrument. This test method isolated the portion of the
piping where the reversal occurred from the test because it was upstream of the
shut valves.
The error was found as a result of a fortuitous observation by an instrument
technician who was performing an unrelated test. If this technician had not
noticed the error, there was a significant possibility that the plant would
have operated with one division of ECCS unavailable.
The safety significance of these events was reduced because the plant was in a
cold shutdown condition. However, no ECCS equipment was available for automatic
operation in the event of low reactor vessel level . In addition, secondary
containment was allowed to be relaxed because the licensee believed ECCS
Division I was operable. Primary containment also was open. Consequently,
had a leak occurred, no ECCS systems would have functioned automatically and
secondary containment would not have been available either. Technical
specifications required the operability of some ECCS equipment during the time
that the plant was shutdown, and upon loss of ECCS, secondary containment
integrity was subsequently required.
Trojan
On July 20, 1985, the Trojan Nuclear Power Plant tripped from 100% power
because of a turbine trip that was caused by the loss of the unit auxiliary
transformer. All systems functioned normally except that low suction pressure
caused one auxiliary feedwater pump to trip and then the other auxiliary
feedwater pump to trip after restart of the first auxiliary feedwater pump.
IN 85-75
August 30, 1985
Page 3 of 3
The cause of the trips of the auxiliary feedwater pumps can be traced back to
improper postmodification adjustment and inadequate postmodification testing
following retrofit of environmentally qualified controllers for the auxiliary
feedwater system. The auxiliary feedwater pump trips on low suction pressure
were caused by excessive combined flow from the two auxiliary feedwater pumps
that draw from a single header from the condensate storage tank. The flow
control valves were open farther than required after new environmentally
qualified controllers had been installed during a recent refueling outage.
When the flow control valves were adjusted following the modification of the
controllers, only one auxiliary feedwater pump was run at a time and used to
adjust the control valve limit switch settings. Consequently, when both pumps
were started following the reactor trip on July 20, 1985, the combined flow was
excessive.
Discussion:
Information Notice 85-23, "Inadequate Surveillance and Postmaintenance and
Postmodification System Testing," described a series of events occurring at
McGuire in November of 1984, where improper system modifications and inadequate
postmodification testing also were involved.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
dward ordan, Director
Divisi of Emergency Preparedness
and ngineering Response
Office of Inspection and Enforcement
Technical Contacts: Eric Weiss, IE
(301) 492-9005
M. Jordan, SRI, LaSalle
(815) 357-8611
Robert Dodds, Region V
(415) 943-3720
Attachments:
1. Figures Illustrating LaSalle Level Instrument Problems
2. List of Recently Issued IE Information Notices
nttachnent 1
IN 85-75
August 3C, 1985
REFERENCE REFERENCE
REACTOR REACTOR
PRESSURE PRESSURE
VESSEL VESSEL VARIABLE
VARIABLE
A A A A
INSTRUMENT INSTRUMENT BLOCK
_ - BLOCK i---
r I
I I
I• • I♦ A
A A
LOW LIS HIGH LOW LIS HIGH
Figure 1 Figure 2
REFERENCE
REACTOR
PRESSURE
VESSEL VARIABLE
V
A A
INSTRUMENT
BLOCK
r--
I
I CLOSED CLOSED
---1 OPEN
CLOSED CLOSED
LOW LIS HIGH
(OPEN)
U
I
TEST
DEVICE
Figure 3
Attachment 2
IN 85-75
August 30, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-74 Station Battery Problems 8/29/85 All power reactor
facilities holding
an OL or CP
84-70 Reliance On Water Level 8/26/85 All power reactor
Sup. 1 Instrumentation With A facilities holding
Common Reference Leg an OL or CP
85-73 Emergency Diesel Generator 8/23/85 All power reactor
Control Circuit Logic Design facilities holding
Error an OL or CP
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35. 24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
OL = Operating License
CP = Construction Permit
SSINS No. : 6835
IN 85-74
UNITED STATES WELD CrIPIY r,11MlSS!P FTIS
NUCLEAR REGULATORY COMMISSION -�
OFFICE OF INSPECTION AND ENFORCEMENT DE FIT
WASHINGTON, DC 20555
SEP 51985 ,`
August 29, 1985
GRELLEY, CO'._O.
IE INFORMATION NOTICE NO. 85-74: STATION BATTERY PROBLEMS
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This notice describes problems that have occurred with lead-acid station
batteries at several nuclear power plants. These problems were discovered as a
result of inspections by the NRC Performance Appraisal Team (PAT). It is
expected that recipients will review the information for applicability to their
facilities and consider actions, if appropriate, to preclude a similar problem
occurring at their facilities. However, suggestions contained in this information
notice do not constitute NRC requirements; therefore, no specific action or
written response is required.
Description of Circumstances:
All four recent inspections by the PAT identified problems with lead-acid
station batteries. Although all of the problems are not listed here, a summary
of a few of the more significant problems with lead-acid station batteries that
were identified as a result of these inspections is provided below:
Cooper PAT Inspection (November 1984):
The battery rated-load discharge test was performed at a discharge rate
significantly less than the manufacturer' s recommended rated-load discharge
rate for the 8-hour period of the test. The licensee had no records of battery
charging following the completion of battery discharge test and consequently
the time and date the batteries were returned to service could not be determined.
The licensee failed to correct specific gravity measurements for electrolyte
temperature and level . The licensee had no written procedures for conducting
charges of the station batteries.
McGuire PAT Inspection (February 1985):
Three cells were placed on single-cell chargers for about 2 years, thus raising
questions regarding the operability of the battery and electrical independence
and separation of the Class 1E dc power systems. The cells on single cell
8508270031 - -
IN 85-74
August 29, 1985
Page 2 of 3
charge were at voltages higher than specified in the vendor manual . Although
one cell in the battery was jumpered out, the float voltage for the entire
battery was not reduced; consequently each cell was floated at a voltage higher
than specified in the battery vendor manual . The battery performance discharge
test was performed improperly because the test was stopped before reaching the
minimum specified voltage.
Susquehanna PAT Inspection (February 1985):
There were no station procedures for maintaining station batteries in accordance
with the battery vendor' s manual or IEEE Std 450-1975, "IEEE Recommended
Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries
for Generating Stations and Substations" (which is endorsed by Regulatory Guide
1. 129, "Maintenance, Testing and Replacement of Large Lead Storage Batteries
for Nuclear Power Plants"). Intercell resistance values were not compared with
previous values to determine when corrective action was required. The licensee
did not always conduct equalizing charges when required; nor did the licensee
have procedures for monitoring the progress of an equalizing charge or determining
when the charge should be terminated. Surveillance procedures for the 60-month
rated-capacity discharge test did not conform to IEEE Std 450-1975 because the
test was terminated at the end of 8 hours, instead of when the terminal voltage
fell to the minimum specified value (usually 1.75 volts per cell). The licensee' s
procedures did not require that the average specific gravity be calculated and
compared to the technical specification acceptance criteria.
San Onofre PAT Inspection (March 1985):
During the first 2 years of operation, the battery capacity tests required by
the Final Safety Analysis Report (FSAR) and IEEE Std 450-1980 were not performed
on Units 2 and 3. The total battery float voltage was not adjusted to account
for two jumpered out cells. The two jumpered out cells did not receive the
manufacturer' s specified surveillances, maintenance, or charges. The pilot
cells were not being changed on a yearly basis, as recommended by the vendor' s
technical instructions. The station engineer responsible for the technical
aspects of battery operation, maintenance, and surveillance did not receive
surveillance results and data sheets on a routine basis.
Discussion:
Recent IE inspections of operating facilities indicate that several widespread
deficiencies may exist in the operation and maintenance of station batteries.
These deficiencies are attributable to a variety of causes, including licensee
error, inadequate knowledge of batteries by maintenance technicians and
supervisors, and inadequate procedural guidance. The results of these
inspections suggest a general lack of appreciation amongst licensee personnel
for proper maintenance and surveillance of station batteries. Although batteries
contain no moving parts, considerable care and attention to detail is required
to maintain them operable. Too often, licensees may be treating these vital
engineered safety features (ESF) power supplies as "passive" components and not
providing them the necessary management and technical attention.
IN 85-74
August 29, 1985
Page 3 of 3
The following reference materials provide guidance as to the individual
requirements for a facility' s station batteries.
1. IEEE 450-1975, and 1980
2. Regulatory Guide 1.129, Rev. 1, "Maintenance, Testing and Replacement
of Large Lead Storage Batteries for Nuclear Power Plants" (This
regulatory guide endorses IEEE Std 450-1975 with certain exceptions. )
3. Facility Technical Specifications
4. Final Safety Analysis Report (FSAR)
5. Station Battery Vendor Technical Manual (The vendors of station
batteries periodically update their manuals to include revised
guidance. )
Other recent problems with station batteries were described in IE Information
Notice 84-83: VARIOUS BATTERY PROBLEMS, November 14, 1984.
No specific action or written response is required by this information notice.
If you have questions about this matter, please contact the Regional Adminis-
trator of the appropriate NRC regional office or this office.
s/I-'---
ar Jordanor
Divisi of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
CONTACTS: E. W. Weiss, IE
(301) 492-9005
L. J. Callan, IE
(301) 492-9497
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 85-74
August 29, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
84-70 Reliance On Water Level 8/26/85 All power reactor
Sup. 1 Instrumentation With A facilities holding
Common Reference Leg an OL or CP
85-73 Emergency Diesel Generator 8/23/85 All power reactor
Control Circuit Logic Design facilities holding
Error an OL or CP
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35.24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
OL = Operating License
CP = Construction Permit
SSINS NO. : 6835
IN 84-70 SUPP. 1
/NEIO COOFTY Prt"t•r., ,.'
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMEN ! SEP 3 ,19
WASHINGTON, D. C. 20555
August 26, 1985 GRtE LL
IE INFORMATION NOTICE NO. 84-70 SUPPLEMENT 1: RELIANCE ON WATER LEVEL
INSTRUMENTATION WITH A COMMON
REFERENCE LEG
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is provided to alert licensees and applicants of the
potential for degradation of safety.associated with operator reliance on level
instruments that share a common reference leg. In this regard, this notice
supplements and reemphasizes the information contained in IE Information Notice
84-70, Reliance On Water Level Instrumentation With A Common Reference Leg. This
notice serves to alert licensees and applicants to the need for operators to
recognize normal and abnormal water level instrument behavior under various plant
conditions. Recipients are expected to review the information for applicability
to their facilities and consider actions, if appropriate, to preclude similar
problems occurring at their facilities. However, suggestions contained in this
information notice do not constitute NRC requirements; therefore, no specific
action or written response is required.
Description of Circumstances:
On February 13, 1985, while performing a reactor startup at TVA' s Browns Ferry
Nuclear Plant, a half scram was received on low reactor water level . A few
minutes before the half scram, the operators had noticed that two of the three
narrow-range General Electric measurement and control (GEMAC) water level
instruments were reading approximately 40 inches of reactor vessel water level .
The other narrow-range instrument was indicating approximately 10 inches of water
level . Two wide-range YARWAY instruments were observed by the operators to be
indicating approximately 40 inches. At the time of the half scram, reactor
pressure was approximately 40 psig and reactor coolant temperature was
approximately 286°F. Although four of the instruments observed by the operators
indicated nearly normal reactor water level (33 ± 5 inches) , actual reactor water
level was approximately 10 inches.
8508230052
IN 84-70 SUPP. 1
August 26, 1985
Page 2 of 3
The operators incorrectly concluded that the narrow-range instrument indicating
10 inches was erroneous since four other level instruments were indicating high.
The two GEMAC instruments that indicated 40 inches share a common reference leg.
This reference leg had lost some of its water inventory, causing all level
instrumentation that tapped off that leg to erroneously indicate high. The
GEMAC instrument that was reading 10 inches tapped off a different reference
column than the aforementioned instruments. The two wide-range YARWAY
instruments each have separate reference columns not shared by any of the narrow-
range GEMAC instruments. At approximately 40 psig reactor pressure and with
actual reactor water level at 33 ± 5 inches, the YARWAY instruments should
indicate ?60 inches, a normal YARWAY level indication. A YARWAY level indication
of 40 inches should have alerted the operators that water level was abnormally
low. The operators did not check the shutdown vessel flooding range level
indication which was available in the control room. This instrument would have
confirmed actual low water level conditions since it is calibrated for cold plant
conditions.
Discussion:
The cause of this event was a partial loss of water inventory from a reference
leg that is common to several water level instruments, including instrument
channels required by the Technical Specifications. From a reactor safety
perspective, this event highlights the need for operators to be cognizant of
level instruments that share a common reference leg and also to be aware of
level instrument behavior subject to various plant conditions.
A problem in a reference leg, such as that experienced at Browns Ferry, not only
affects level indication, but may also affect the operability of reactor pro-
tection instrument channels required by the Technical Specifications. In this
event, the most critical technical specification instruments affected were two
level switches, one in each reactor protection trip system train. These switches
were inoperable since they were common to the faulty reference leg. Had the
operators realized earlier what instruments and switches were affected by the
faulty reference leg, proper corrective action may have been taken to shut down
the plant in a timely manner in accordance with the Technical Specifications.
It is important that operators understand level instrument response to various
plant conditions. One way to achieve this understanding is through training
to emphasize level instrument system design, temperature and/or pressure compen-
sation, instrument calibration, and the purpose of the instruments (i . e. ,
process monitor vs. control ). Although the YARWAY instruments are designed to
provide wide-range accident level indication and are calibrated to be most
accurate at normal operating reactor pressure, the operators at Browns Ferry
could still have used these instruments as additional level response indication.
Had the operators realized that a YARWAY level of 40 inches was abnormally low
for the existing low reactor pressure, they might have been alerted earlier to
the fact that actual vessel water level was low.
Licensees and applicants may wish to review their system descriptions, operating
procedures, and operator training programs to ensure that a common reference leg
shared by multiple level instruments is adequately addressed. Operator awareness
of the effects a malfunction in a common reference leg can have on the level
instruments and recognition of proper water level indication subject to various
plant conditions can enhance plant safety.
IN 84-70 SUPP. 1
August 26, 1985
Page 3 of 3
Although Information Notice 84-70 described an event at a pressurized water
reactor and this supplement describes an event at a boiling water reactor,
problems associated with reliance on water level instrumentation with a common
reference leg can occur at either type of reactor.
No specific action or written response is required to this information notice.
If you need additional information about this matter, please contact the Regional
Administrator of the appropriate NRC regional office or one of the technical
contacts listed below.
lrd . Jordan, Director
Divisi of Emergency Preparedness
and gineering Response
Office of Inspection and Enforcement
Technical Contacts: Eric W. Weiss, IE
(301) 492-9005
P. D. Wagner, Region II
(404) 221-2688
Attachment:
List of Recently Issued IE Information Notices
Attachment 1
IN 84-70 Supp. 1
August 26, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-73 Emergency Diesel Generator 8/23/85 All power reactor
Control Circuit Logic Design facilities holding
Error an OL or CP
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35. 24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
85-66 Discrepancies Between 8/7/85 All power reactor
As-Built Construction facilities holding
Drawings And Equipment an OL or CP
Installations
OL = Operating License
CP = Construction Permit
Enclosure
EVALUATION OF TECHNICAL SPECIFICATION CHANGES
FOR INSERVICE INSPECTION AND TESTING PROPOSED
ON DECEMBER 30, 1983
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN NUCLEAR GENERATING STATION
DOCKET 50-267
INTRODUCTION
By letter dated December 30, 1983 (Reference 1) the Public Service Company
of Colorado (PSC) proposed changes to the Fort St. Vrain Technical
Specifications concerning inservice inspection and testing requirements
(ISIT). These proposed changes are a continuation of the ISIT upgrade
program initiated in response to a commitment in Section 3.3 of the Safety
Evaluation Report of January 20, 1972 (Reference 2) to review the in-service
inspection program five years following the start of commercial operation.
The NRC and PSC agreed to a staged and prioritized upgrade program with those
systems of highest priority identified as Category I (Reference 3). Our
reviews and revisions of essentially all the Category I systems were issued
on March 8, 1983 (Reference 4) as License Amendment No. 33 together with a
supporting Safety Evaluation Report. Reference 1 addresses a single Category
I surveillance interval and certain Category II and III changes but does not
complete the upgrade program. We used the ASME Boiler and Pressure Vessel
- 2 -
Code, Section XI , Division 2, "Rules for Inspection and Testing of
Components of Gas Cooled Reactors," as guidance in performing our review.
This was in addition to our usual custom review practice for Fort St. Vrain
that is based upon operating history and experience, consideration of safety
importance based on safety analyses, and the unique design features of the
facility.
GENERAL COMMENTS
1) The NRC has not yet formally adopted Division 2, Section XI of the ASME
Code. This is due to both a lack of resources for the NRC to participate
in the final stages in the Code's development and to a technical concern.
The technical concern pertains to Articles 1GB-1221, "Exemptions Based on
Component Function," and 1223,"Exemptions Based on Component Size,"
which would exempt from examinations those connections to the primary
system for which failure would not result in a rate of depressurization
greater than that used for the design basis accident. While we believe
exemptions for connections of the size of instrument lines are acceptable,
as is permitted for LWRs, substantially larger sizes, and certainly
those approaching the reference cross sectional area for a design basis
accident should not be exempted from review.
- 3 -
In spite of the fact that the Division 2 Code has not been adopted by
the NRC it is useful for guidance purposes in many areas of gas reactor
ISIT. In the present review, changes in many surveillance requirements
for safety related water systems were proposed. For this reason sections
of the ASME Code dealing with pumps (Article 1GP) and valves (Article
IGV) were particularly relevant to our review. In general , these sections
were modeled after corresponding sections of the ASME Code for LWRs and
are judged in many cases directly pertinent to inspection and testing of
the safety related water systems of Fort St. Vrain.
We have in the past and continue to recommend that Public Service of
Colorado commit to referencing applicable sections of the Division 2
ASME Code. By reference to the Code both the presentation and the review
of the inspection and testing program is expedited and many details of
procedures practice are clarified.
2) As stated previously the inservice inspection upgrade program is not yet
complete. In accordance with Reference 3, the following systems remain
outstanding:
- 4 -
TABLE 1
CATEGORY II SYSTEMS
CONTROL AND ORIFICE ASSEMBLY (12)
NITROGEN SYSTEM (25)
EMERGENCY FEED AND CONDENSATE SYSTEM (31)
PURIFICATION COOLING WATER SYSTEM (47)
CATEGORY III SYSTEMS
FUEL STORAGE FACILITY AUXILIARY SYSTEM (14)
CONTROL COMPLEX HVAC (75)
FIRE PUMP HOUSE HVAC (75)
AUXILIARY BOILER FUEL OIL SYSTEM (84)
ESSENTIAL ELECTRIC POWER SYSTEM (92)
CATEGORY IV SYSTEMS
FEED AND CONDENSATE SYSTEMS (31, 32, 33)
CIRCULATING WATER SYSTEM (41)
DECAY HEAT REMOVAL HX (42)
FIRE PROTECTION SYSTEM (45)
CO2 SYSTEM (51)
TURBINE STEAM SYSTEMS (52, 53, 54, 55)
- 5 -
TABLE 1 (Continued)
RADIOACTIVE LIQUID WASTE STORAGE SYSTEM (62)
RADIOACTIVE GAS WASTE SYSTEM (63)
FUEL STORAGE WELL HVAC (73)
PCRV ENVIRONMENT MONITORING (73)
CONTROL COMPLEX ENVIRONMENT MONITORING (75)
AUXILIARY BOILER (84)
PRIMARY AND AUXILIARY ELECTRIC POWER SYSTEMS (92)
COOLANT MEASUREMENT DISPLAY SYSTEM (93)
OVERALL PLANT CONTROL SYSTEM (93)
PCRV INSTRUMENTS AND DATA ACQUISITION SYSTEM (93)
FAST GAS AND IODINE SAMPLING SYSTEM (93)
CONTROL ROD AND ORIFICING CONTROL SYSTEM (93)
SG T/C AND S/G INSTRUMENTATION SYSTEM (93)
ANALYTICAL INSTRUMENTATION SYSTEM (93)
The review of the surveillance requirements for these systems is expected
to be accomplished mainly in the comprehensive Technical Specification
Upgrade Program recently initiated by PSC (Reference 5) . This program
is organizing the Technical Specifications in the general format of
NRC's PWR Standard Technical Specifications.
- 6 -
In Reference 3 PSC provided a listing, divided into the four priority
categories, of all the systems to be reviewed in the context of the 1972
commitment regarding upgraded ISIT requirements given in Reference 2.
Before we can consider PSC is in compliance with its 1972 commitment PSC
should review this listing, make modifications if needed, and then certify
to its completeness.
EVALUATION
The proposed changes submitted and reviewed by this amendment action are as
follows:
TABLE 2
CATEGORY I SYSTEM
SR 5.3.9 SAFETY VALVES SURVEILLANCE
CATEGORY II SYSTEMS
SR 5.2.7 WATER TURBINE DRIVE SURVEILLANCE
SR 5.2.8 BEARING WATER PUMP AND MAKEUP PUMP SURVEILLANCE
SR 5.2.9 HELIUM CIRCULATOR BEARING WATER ACCUMULATORS SURVEILLANCE
SR 5.2.10 (b,d) FIRE WATER SYSTEM/FIRE SUPPRESSION WATER SYSTEM SURVEILLANCE
SR 5.2. 16 (g) PCRV CLOSURE LEAKAGE SURVEILLANCE REQUIREMENTS
SR 5.2.21 ACM TRANSFER SWITCHES, VALVES AND INSTRUMENTS
SR 5.2.24 REACTOR AUXILIARY COOLING WATER SYSTEMS SURVEILLANCE
SR 5.3.4 SAFE SHUTDOWN COOLING VALVES SURVEILLANCE
SR 5.4.4 PCRV COOLING WATER SYSTEM TEMPERATURE INSTRUMENTS SURVEILLANCE
SR 5.4.5 PCRV COOLING WATER SYSTEM FLOW INSTRUMENTS SURVEILLANCE
SR 5.5.3 REACTOR BUILDING EXHAUST SYSTEM SURVEILLANCE
- 7 -
TABLE 2 (Continued)
CATEGORY III SYSTEM
SR 5.7.2 FUEL STORAGE FACILITY SURVEILLANCE
Our evaluation of each of these proposed changes is stated below.
1. SR 5.2.7 - Water Turbine Drive Surveillance
a) The proposed change would extend the annual test interval for one
circulator and the associated water supply valving in each loop to
the next scheduled plant shutdown if the test was not performed
during the previous year. We find this proposed change acceptable
provided that the surveillance interval does not exceed 18 months
on the basis that (1) operating experience has illustrated
sufficiently satisfactory performance of this system such that no
significant hazard is created by an extension of this test
interval , (2) potential hazards from an additional shutdown and
startup transients would be avoided and (3) decay heat can still be
remaved via steam driven circulators or the Liner Cooling System.
The provision to not have the surveillance interval exceed 18
months is consistent with NRC Standard Technical Specifications
for LWRs and ensures a minimum surveillance interval .
- 8 -
b) The proposed change would extend the annual test interval for
safety valves in the water turbine supply lines to the next
scheduled plant shutdown if the test was not performed during the
previous year. We find this proposed change acceptable, provided
that the surveillance interval does not exceed 18 months , based on
the same reasons as given in comment 1 (a) above.
c) The phrase "every three months" has been changed to read
"quarterly" for the functional testing of both turbine water
removal pumps and the turbine removal tank overflow to the reactor
building sump. We recommend that the standard technical
specification terminology of 92 days be used rather than that
proposed.
- 9 -
1U.1 Draft Item 4.5. 1. 1b.1 - Helium Circulator
- Power Operation and Low Power
The turbine water removal pumps would be tested only once per refueling
cycle, an extension of the surveillance interval inconsistent with both
the ASME Code and the current 92 day interval . No mention is made of
surveillance of turbine water drain tank overflow. Our concerns in
this regard are: (1) Why should the test interval for the turbine water
removal pumps be extended to once per refueling cycle? and (2) Why is
surveillance of the turbine water drain tank overflow not specified?
2. SR 5.2.8 - Bearing Water Pump and Makeup Pump Surveillance
The bearing water makeup pumps have been added to this surveillance.
We recommend that circulating bearing water pumps be added to the title
for consistency.
a) The phrase "every three months" has been changed to read
"quarterly" for the operation of the Normal Makeup Pump in the
recycle mode. We recommend that the terminology of STS of 92 days
be used rather than that proposed.
1U indicates that this is an item from the Fort St. Vrain Technical
Specification Upgrade program that corresponds to the Technical
Specification under review immediately preceding.
- 10 -
b) Thg phrase "every three months" has been changed to read
"quarterly" for the functional testing of the Emergency Makeup
Pump. We recommend that the terminology of STS of 92 days be used
rather than that proposed.
d) The proposed surveillance for the bearing pumps would provide
for a functional test of the pumps and associated instruments and
controls at each scheduled plant shutdown, or at the next
scheduled plant shutdown if less than a year has elapsed from the
previous test. This schedule would not disrupt normal plant
operation and provides a test not previously required. We find
this surveillance acceptable as it is consistent with current
practice in the application of the ASME Code to current plants,
provided the surveillance interval does not exceed 18 months.
2U. Draft Item 4.5.1. 1.b2 - Helium Circulator
- Power Operation and Low Power
The bearing water makeup pump would be tested once per refueling cycle
and no mention of the emergency makeup pump is made. This is incon-
sistent with SR 5.2.8 and the ASME Code. In addition, no mention of
bearing water pump surveillance (SR 5.2.8d) is made. Therefore, infor-
mation should be provided on: (1) Why is the bearing water makeup pump
surveillance inconsistent with SR 5.2.8 and the ASME Code? and (2) Why
is there no surveillance requirement for the bearing water pump?
- 11 -
3. SR 5.2.3 - Helium Circulator Bearing Water Accumulators
The proposed test interval for testing of the helium circulator bearing
water accumulators, instruments and controls would be extended from
monthly to quarterly. The licensee justifies this change on a review
of prior test results which shows satisfactory performance. Based on
this justification we the find the proposed change acceptable. We
recommend that the quarterly interval be stated as 92 days.
3U. Draft Item 4.5. 1.1.a - Helium Circulator
- Power Operation and Low Power
This Draft Item requires functional testing of the bearing water
accumulators system at 31 day intervals but does not identify testing
of the instrumentation and controls as an explicit requirement. In
addition, annual calibration of the instruments is not indicated as is
done in SR 5.2.9. Therefore, information should be provided addressing:
(1) Why does the proposed surveillance of the accumulators omit required
testing of the instrumentation and controls? and (2) Why is annual
calibration not indicated?
4. SR 5.2. 10 - Fire Water System/Fire Suppression Water System Surveillance
b) A reduction by 5 percent in the flow and head testing requirements
for the firewater pumps has been proposed to account for pump
degradation. Degradation to this degree is acceptable under the
- 12 -
ASME Code and the pump performance continues to exceed the minimum
performance requirements by a sufficient margin. We find the
proposed changes acceptable.
d) The fire suppression water system pressure is changed to read "275
feet water gauge" from "125 psig." This is an acceptable change.
4U. Draft Item 4.5.4
Why does the system functional test in Part f.5 not provide for minimum
flow and head measurements? To be acceptable this Draft Item should be
consistent with the requirements of the ASME Code.
5. SR 5.2.16 - PCRV Closure Leakage Surveillance Requirements
g) The proposed change would require once during each refueling cycle
a leakage test and for each helium purification cooler well , a
calibration of the well pressure monitoring instruments and a
functional test of the instruments and controls used to automatically
isolate the purification system. The addition of this surveillance
requirement verifies the operability of instruments used to monitor
containment integrity. To make a judgement on the acceptability of
the proposed leakage test a description of how this requirement
meets the intent of the ASME Code, Article 1GB, "Examination and
Inspection," should be provided.
- 13 -
5U. Draft Item 4.6.4.5.a.2 - PCRV Integrity
Draft Item 4.6. 1.3 - Interspace Pressurization
It is not clear that these two surveillances together contain the
essential requirements of SR 5.2.16(g) . If they do not, then SR 5.2. 16(g)
should be included explicitly. Alternatively, (1) each Draft Item could
be modified to clarify that it pertains to "well" closures as well as
"penetration" closures, or (2) a definition could be added that clarifies
that a "well" closure is also a penetration closure.
6. SR 5.2.21 - ACM Transfer Switches, Valves, and Instrument Surveillance
The surveillance requirement has been retitled from the previous title
of SR 5.2.21 - Handvalve and Transfer Switch Surveillance.
a) for those valves and transfer switches that must be manually
positioned for actuation of the Alternate Cooling Method (ACM)
mode of operation, the licensee proposes to change the
surveillance interval to annually or at the next scheduled plant
shutdown if such a test was not performed during the previous
year. While we understand that full operation of these valves and
switches is not possible during plant operation, we nevertheless
believe that it is necessary to demonstrate operability of these
components more frequently. Thus we do not find this proposed
change acceptable and recommend that the original surveillance
interval for an operability check of this equipment be maintained
(4 to 8 months) and a full functional test be performed at annual
or at refueling intervals not to exceed 18 months.
- 14 -
b) A new surveillance requirement for calibration at each refueling
interval has been proposed for local indicators for the helium
purification dryer inlet temperature, for the helium purification
pumpdown line pressure and for the reactor plant cooling water
surge tank cover gas pressure. From the information provided it
is not clear that: (1) these proposed surveillances are sufficient
to assure operational readiness of these components, and (2) the
components to be given surveillance provide a complete set to assure
operational readiness of the systems they serve. Therefore, infor-
mation needs to be provided addressing the above.
6U. Draft Item 4.7.8.c.2 - ACM Diesel Generator
This Item pertains to the surveillance of the ACM
transfer switches. The proposed 18 month surveillance interval for the ACM
switches is longer than the surveillance intervals proposed for these
same switches in SR 5.2.21 and is not considered acceptable for the
same reason and discussed in comment 6(a) above. In addition, it
should be explained where the remaining surveillances have been
located in the Upgrade Technical Specifications.
7. SR 5.2.24 - Reactor Auxiliary Cooling Water Systems Surveillance
The title of this surveillance requirement has been changed from
Circulation Water Makeup System Surveillance. We find this
acceptable.
- 15 -
b) The surveillance interval for functionally testing each
circulating water pump is proposed to be extended to monthly from
weekly. As the monthly interval is in accordance with the ASME
Code and surveillance requirements have been added to the proposed
change regarding instrument calibration, pump performance
capability and mechanical condition, we find the proposed change
acceptable.
d) The proposed surveillance requirement would be a new requirement
pertaining to the integrity of the circulating water makeup pond
embankments. The proposed addition is consistent with LWR service
water requirements and is considered acceptable.
e) The proposed surveillance requirement would be a new requirement
pertaining to the testing of each service water pump and the
associated instruments. We have reviewed these requirements and
have found them in partial accord with the ASME Code. We would
find them acceptable if the licensee either conforms to the detailed
requirements of the ASME Code or provides an acceptable alternative.
f) The proposed surveillance requirement would be a new requirement
pertaining to the testing of each reactor plant cooling water pump
and the associated instruments. We have reviewed these
- 16 -
requirements and have found them in partial accord with the ASME
Code. We would find them acceptable if the licensee either conforms
to the detailed requirements of the ASME Code or provides an
acceptable alternative.
g) The proposed surveillance requirement would be a new requirement
pertaining to the testing of each purification cooling water pump
and the associated instruments. We have reviewed these require-
ments and have found them in partial accord with the ASME Code.
We would find them acceptable if the licensee either conforms to
the detailed requirements in accordance with the ASME Code or
provides an acceptable alternative.
h) The proposed surveillance requirement would be a new requirement
pertaining to the testing and calibration of instruments and valves
used for automatic insolation of portions of the reactor plant
cooling water system. We find that an interval of each refueling
cycle not to exceed 18 months for a full stroke test of each valve
is acceptable for those valves that cannot be tested during plant
operation. However, the interval for a functional check of those
valves and instruments capable of being tested by a partial stroke
should be performed semi-annually in accordance with the precedent
of SR 5.2.21 or quarterly in accordance with the guidance of the
ASME Code.
- 17 -
7U. Draft Items
The Draft does not recognize the Reactor Auxiliary Cooling Water System
as a comprehensive system and appears to make no provisions for some of
the surveillances described in SR 5.2.24. For example, in Section 4.7.4,
"Service Water System" no specific surveillances are required although
an operability demonstration is made by reference to surveillance testing
of the diesel generators (4.8. 1. 1.2) . However, 4.8.1.1.2 has no explicit
requirements pertaining to pumps, valves and instruments of this system.
Specific surveillance requirements for safety systems is required by the
ASME Code and should be added for the Reactor Auxiliary Cooling Water
System.
8. SR 5.3.4 - Safe Shutdown Cooling Valves Surveillance
The licensee proposes to test valves used for safe shutdown cooling on
an annual basis or following scheduled plant shutdown. This is not
acceptable except for cases where it is not physically possible to
perform a more frequent surveillance. For those valves that can be
tested during reactor operation, are required to initiate and function
during safe shutdown cooling, and which are not called upon to operate
during normal plant operation, the licensee should provide testing
requirements and intervals in conformance with the ASME Code.
- 18 -
8U. Draft Item 4.5.2.1.a - Steam Generator
The only pertinent surveillance for valves of the safe shutdown cooling
system is given by the above Draft Item which requires testing at 18
month intervals of the flow through the emergency feedwater header and
the emergency condensate header to the steam generators. This is not
satisfactory because it does not meet the objectives of the ASME Code
which requires shorter test intervals and explicit testing procedures.
The problem exists with all other Draft surveillance requirements in
Section 4.5, "Safe Shutdown Cooling" as these also refer to Item 4.5.2. 1a.
9. SR 5.3.9 - Safety Valves Surveillance
a) This proposed surveillance would require verification of safety
valve setpoints at five year intervals for the steam generator
superheater, reheater and steam/water dump tank. The requirement
is satisfactory provided 1) that a schedule for additional testing
is developed for any valve in a system that fails to function on a
regular test and 2) that an acceptable test procedure is developed
or referenced. Conformance to the ASME Code (Subsection 1GV) for
Class C valve testing would meet the above requirements and simplify
the development of an acceptable technical specification requirement.
- 19 -
b) The licensee proposed that all other Class I safety valves not
covered by other surveillance requirements shall be setpoint
tested at 10 year intervals. This is unacceptable. The licensee
should conform with the ASME Code in this matter.
9U. Draft Item 4.5.2.1.b - Steam Generator
This Item provides for setpoint testing of superheater and reheater
safety valves at five year intervals but does not identify a test
procedure or provide guidance in the event of testing failures.
Conformance with the ASME Code would resolve this problem.
Draft Item 4.7. 1.2
This Item provides a similar requirement for the steam/water pump system
safety valve setpoint. Again, conformance with the ASME Code would
resolve this problem.
10. SR 5.4.4 - PCRV Cooling Water System Temperature Instrument Surveillance
a) The proposed surveillance requirement clarifies the monthly
monitoring of the PCRV cooling system water inlet temperature,
individual tube water outlet temperatures , and the associated
outlet temperature alarms. We find the proposed clarification
acceptable.
- 20 -
b) The proposed surveillance requirement clarifies requirements for
annual calibration of the temperature monitoring scanner, the inlet
and outlet header temperature indicators , and the outlet subheader
temperature indicators. We find the proposed clarification
acceptable.
c) The proposed surveillance requirement would extend calibration
of the inlet header and tube outlet thermocouples from an annual
interval to a five year interval . We do not find this surveillance
interval extension acceptable since no justification has been
provided.
10U. Draft Item 3/4.6.3 - PCRV Liner Cooling System Temperatures
No surveillance requirements comparable to SR 5.4.4 are given in this
Draft Item. Consideration should be given to providing similar require-
ments in the upgraded technical specifications.
11. SR 5.4.5 - PCRV Cooling Water System Flow Instruments Surveillance
The proposed surveillance would extend annual calibration of the flow
scanner instruments and alarms and the six subheader flowmeters to the
next scheduled plant shutdown if they were not calibrated during the
previous year. We find this extension acceptable up to a surveillance
interval not exceeding 18 months since the potentials for additional
plant transients are reduced and since the LWR-STS, in general , specifies
surveillance intervals not to exceed 18 months when utilizing intervals
of shutdown or refueling.
- 21 -
11U. Draft Lem 3/4.6.2 - PCRV Liner Cooling System
No surveillance requirements comparable to SR 5.4.5 are given. We
believe that similar requirements should be considered in the upgraded
technical specifications.
12. SR 5.5.3 - Reactor Building Exhaust System Surveillance
e) The proposed surveillance requirement would verify at weekly
intervals that the total pressure drop across the HEPA filter and
charcoal absorber banks to be less than six inches of water at
filter design flow ± 10 percent. Subsequent to this proposal the
Draft Upgrade Technical Specifications have been issued and it is
our opinion that appropriate portions of Draft Item 4.6.5.2.c.3,
"Reactor Building Exhaust System," which generally specifies the
testing requirements given in the Standard Technical Specifications,
should be utilized in lieu of PSC's proposal .
f) The proposed surveillance requirement would verify annually the
performance capability and mechanical condition of each exhaust
fan or at the next scheduled shutdown if such verification was not
performed during the previous year. We find this surveillance
requirement acceptable provided that the surveillance interval
does not exceed 18 months , since the LWR-STS, in general ,
specifies surveillance intervals not to exceed 18 months when
utilizing intervals of shutdown or refueling.
- 22 -
g) The proposed surveillance would require calibration of the
instrumentation associated with the filters and fans at annual
intervals or at the next scheduled shutdown if calibration was not
performed during the previous year. We find these surveillance
requirements are acceptable up to a surveillance interval not
exceeding 18 months since the potentials for additional plant
transients are reduced and since the LWR-STS, in general , specifies
surveillance intervals not to exceed 18 months when utilizing
intervals of shutdown or refueling.
12U. Draft Item 4.6.5.2 - Reactor Building Exhaust System
The surveillance requirements described herein are generally consistent
with LWR Standard Technical Specifications. However, additional infor-
mation should be provided addressing: (1) Why there is no provision for
vibration monitoring of the fans and for calibration of the exhaust
system instrumentation and (2) Why should not the surveillance frequency
be at 720 hour intervals rather than semi-annually?
13. SR 5.7.2a - Fuel Storage Facility Surveillance
The proposed surveillance would require an annual functional test of
the emergency ventilation system. This surveillance, together with
parts a and b of SR 5.7.2c, have been substantially revised in Draft
Item 4.9.3, "Fuel Storage Well ." We believe action on this item should
be deferred until discussions on comment 13U. below are completed.
- 23 -
13U. Draft Item 4.9.3 - Fuel Storage Well
The material in this Draft Item appears to represent an improved and
better developed surveillance than that described in SR 5.7.2 and should
be considered for inclusion in SR 5.7.2a.
- 24 -
REFERENCES
1. 0. R. Lee (PSC) letter to J. T. Collins, "Proposed Technical
Specification Changes - Inservice Inspection and Testing
Requirements," No. P-83416, December 30, 1983.
2. Safety Evaluation by the Division of Reactor Licensing, U.S. Atomic
Energy Commission in the Matter of Public Service Company of Colorado,
Fort St. Vrain Nuclear Generating Station, Docket No. 50-267, January 30,
1972.
3. J. K. Fuller (PSC) letter to S. A. Varga, "Fort St. Vrain Inservice
Inspection and Testing Program," No. P-79289, November 30, 1979.
4. P. C. Wagner (NRC) letter and enclosure to 0. R. Lee (PSC) , "Fort St.
Vrain Nuclear Generating Station, Amendment No. 33 to Facility
Operating License DPR-34," March 8, 1983.
5 0. R. Lee (PSC) letter to E. H. Johnson (NRC) , "Technical Specification
Upgrade Program," December 20, 1984.
SSINS No. : 6835
IN 85-73
UNITED STATESNU GELD
OFFICECOFAR REGULATORY COMMISSION
INSPECTIONANDENFORCEME i n POtro Y� p�l`'��0,T.C
WASHINGTON, D. C. 20555 '( \ ; "iL,
:') 111
August 23, 1985 SEP 31985 '' '
GH _
IE INFORMATION NOTICE NO. 85-73: EMERGENCY DIESEL GENERATOR C0NTPOL-EIRCUTT
LOGIC DESIGN ERROR
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is to alert recipients of a potentially significant
emergency diesel generator (EDG) control logic error that could prevent trans-
fer to the emergency bus while the EDG is in the "maintenance shutdown" mode.
It is expected that recipients will review the information for applicability to
their facilities and consider actions, if appropriate, to preclude a similar
problem occurring at their facilities. However, suggestions contained in this
information notice do not constitute NRC requirements; therefore, no specific
action or written response is required.
Description Of Circumstances:
According to the design, the EDGs at Rancho Seco Nuclear Power Generating
Station enter the maintenance shutdown control mode whenever they are normally
shut down from the control room or the remote EDG control panel . On June 1,
1985, the plant was shut down for refueling, an EDG was in the maintenance
shutdown control mode after being secured from an operational condition (idling
at 600 rpm with the output breaker open) , when an emergency bus was de-energized
for planned work on a parallel bus. This created an undervoltage condition
equivalent to a loss of offsite power (LOOP) on the emergency bus. The diesel
generator sped up to the design speed but the EDG output breaker continuously
cycled open and closed, thereby rendering the EDG set inoperable.
Investigation by the licensee indicates that the cycling of the EDG output
breaker was the result of a design error in the EDG control circuit logic.
According to the licensee, the design 'deficiency affects proper response of
the EDG set when it is operating in the maintenance shutdown control mode.
Normal surveillance testing would not discover the control circuit design
error because surveillance is not done in the maintenance shutdown control mode.
The June 1, 1985 event at Rancho Seco represents the first time in the life of
the plant that an undervoltage signal occurred with an EDG in the maintenance
shutdown control mode.
8508210321
IN 85-73
August 23, 1985
Page 2 of 2
When an EDG is secured from operation, the control circuit logic places it in
the maintenance shutdown control mode. In this mode, the control logic opens
its output breaker and reduces its speed from 900 to 600 rpm. The EDG then
idles at 600 rpm for 15 minutes before coasting down to rest.
If a LOOP should occur while an EDG is in the maintenance shutdown control
mode, the undervoltage signal causes it to speed back up to 900 rpm and to
close its output breaker. This would cause the undervoltage signal to drop out.
However, the maintenance shutdown control mode does not drop out for 30 seconds
after the receipt of the undervoltage signal because of the control circuit
design error. Thus, the maintenance shutdown control logic senses that the EDG
output breaker is closed, opens the breaker, and resets the 15-minute timer for
the maintenance shutdown control mode. As soon as the EDG output breaker opens,
the undervoltage signal recurs and the EDG output breaker closes in response to
the LOOP. The EDG output breaker continues to cycle open and closed as this
process repeats itself. At Rancho Seco, this control circuit logic design
error has been corrected by installing a relay to de-energize the maintenance
shutdown control logic immediately upon receipt of an undervoltage signal .
The Rancho Seco plant utilizes General Motors (GM) Model 20-465-E4 diesel
generators with a 2750 kw nameplate rating. According to the licensee, the
design error was in the interface provided by the Architect-Engineer (Bechtel)
to the shutdown control logic provided by GM. Bechtel has advised the NRC
that the Rancho Seco diesel generator control logic is unique and other plants
designed by them are not affected.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
'PdwardL. Jordan, Director
Division, of Emergency Preparedness
and En4ineering Response
Office of Inspection and Enforcement
Technical Contact: W. Swenson, NRR
(301) 492-7876
R. Singh, IE
(301) 492-8985
Attachment: List of Recently Issued Information Notices
Attachment 1
IN 85-73
August 23, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-72 Uncontrolled Leakage Of 8/22/85 All power reactor
Reactor Coolant Outside facilities holding
Containment an OL or CP
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35. 23 And 10 CFR 35.24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
85-66 Discrepancies Between 8/7/85 All power reactor
As-Built Construction facilities holding
Drawings And Equipment an OL or CP
Installations
85-65 Crack Growth In Steam 7/31/85 All PWR facilities
Generator Girth Welds holding an OL or CP
OL = Operating License
CP = Construction Permit
Hello