ML20140E298
| ML20140E298 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 01/22/1986 |
| From: | Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20140E254 | List: |
| References | |
| 50-458-85-81, NUDOCS 8602030221 | |
| Download: ML20140E298 (9) | |
See also: IR 05000458/1985081
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APPENDIX C
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report: 50-458/85-81
License: NPF-47
Docket:
50-458
Licensee: Culf States Utilities Company (GSU)
P. O. Box 2951
Beaumont, Texas
77704
Facility Name: River Bend Station (RBS)
Inspection At: River Bend Station, St. Francisville, Louisiana
Inspection Conducted: December 1 through December 31, 1985
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Inspectors:
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D. D. Chamberlain, Senior' Resident Inspector
Date
(pars. 1, 2, 3, 4, 5, 6, 7, and 8)
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W. B. Jones, Resi,Jent Inspector
Date
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(pars. 2, 3, 4, 5, 6, 7, and 8,M
Approved:
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J .'
J udon, Chief, Project Section A
Da'te
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P2act r Prt ects Branch
0602030221 060129
ADOCK 05000450
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Inspection Summary
Inspection Conducted December 1 through December 31, 1985 (Report 50-458/85-81)
Areas Inspected: Routine, unannounced inspection of licensee event report (LER)
review, startup test procedure review, startup test witnessing, startup test
program quality assurance review, operational safety verification, and site tours.
The inspection involved 208 inspection-hours onsite by two NRC inspectors.
Results: Within the areas inspected, two violations were issued in the areas of
startup test witnessing and operational safety verification (failure of document
control program and inadequate retest following modifications or repair,
paragraphs 4.A and 6, respectively).
In addition one deviation was issued in
the area of startup test procedure review (failure to require verification of
proper diesel generator load sequencing, paragraph 3).
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DETAILS
1.
Persons Contacted
Principal Licensee Employees
- R. E. Bailey, Supervisor, Quality Concern
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L. Ballard, Projects Supervisor
- C, Banks, Security
- W. H. Cahill, Jr., Senior Vice President, River Bend Nuclear Group
E. M. Cargill, Superintendent, Radiological Programs
- R. P. Carter, Security
- T. C. Crouse, Manager, Quality Assurance (QA)
- D. L. Davenport, Supervisor, Plant Security
- J. C. Deddens, Vice President River Bend Nuclear Group
- Jan Evans, Stenographer
- C. E. Foster, Assistant Plant Security Supervisor
P. E. Freehill, Superintendent', Startup and Test
A. D. Fredieu, Assistant Operations Supervisor
D. R. Gipson, Assistant Plant Manager, Operations
- P. D. Graham, Assistant Plant Manager, Services
- E. R. Grant, Supervisor, Nuclear Licensing
R. W. Helmick, Director, Projects
B. D. lley, Licensing Engineer
- K. C. Hodges Supervisor, Quality Systems
R. Jackson, Shif t Supervisor, Operations
D. Jernigan, Engineer, Startup and Test
G. R. Kimell, Supervisor, Operations QA
- R. King, Engineer, Licensing
A. D. Kowalczuk, Assistant Plant Manager, Maintenance
T. Lacy, Shif t Supervisor, Operations
- W. H. Odell, Manager, Administrative
- T. L. Plunkett, Plant Manager
W. J. Reed, Director Nuclear Licensug
D. Reynerson, Director, Nuclear Plant Engineering
- F. L. Richter, Operations, QA
- C, G. Sprangers, Engineer, QA
R. B. Stafford, Director, Quality Services
- K. E. Suhrke, Manager, Projects
- P. F. Tomlinson, Director Operation QA
C. Warren, Shif t Supervisor, Operations
Stone and Webster
- B. R. Hall, Assistant Superintendent, Field Quality Control
R. L. Spence, Superintendent, Field Quality Control
The NRC senior resident inspector (SRI) and resident inspector (RI) also
interviewed additional licensee, Stone and Webster (S&W), and other
contractor personnel during the inspection period.
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- Denotes those persons that attended the exit interview conducted on
January 10, 1985. NRC resident inspector, W. B. Jones, and NRC security
inspector, R. A. Caldwell, also attended the exit interview.
2.
Licensee Event Report Review
(Closed) License Event Report (LER) 458/85-08:
The NRC inspector reviewed the corrective actions taken by the licensee to
avoid future recurrence of the RPV level transient caused by opening RHR
'.'A" suppression pool valve (1E12*M0VF004A) before RHR "A" shutdown cooling
suction valve (1E12*liOVF006A) was completely closed.
The following corrective actions are complete:
Station Operating Procedures (SOPS) contain the necessary cautions for
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the operator.
A caution statement has been added to Surveillance Test Procedure
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(STP) STP-309-601 instructing the operator to ensure that the shutdown
cooling suction valve F006A is fully shut prior to opening suppression
pool suction valve F004A.
All plant operators have been informed of the incident as well as the
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above procedural change.
A yellow caution sign having the same warning as that added to
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STP-309-601 was mounted on the control panel near the RHR system
suction valve switches as an aid to the operators.
STP-309-601 has been revised to restore isolation valve (1E12*liOVF008
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and 1E12*t10VF009) motor breakers to the closed position prior to
restoration of valves F006A and F004A.
Operations has completed their review of STPs involving the RHR system
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having the potential to cause a level transient. This review revealed
tnat appropriate precautions or detailed instructions exist that would
preclude the simultaneous opening of the F004A and F006A valves.
In addition, the licensee is presently performing an engineering evaluation
to determine if an interlock between the F004A and F006A valves is feasible
to prevent the F004A valve from opening before the F006A valve has closed.
The R1 will monitor this engineering evaluation as an open item
(458/8581-04).
This LER is closed.
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3.
Startup Test Procedure Review
The SRI reviewed STP 1-ST-31, " Loss of Offsite Power," Revision 0, and a
draft Revision 1.
Several comments were generated during this review and
the corments were discussed with startup test personnel. One comment
revealed that the procedure failed to implement a FSAR Chapter 14 commit-
ment to verify proper sequencing of diesel generator loads during a loss of
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offsite power. This failure to implement a FSAR commitment was identified
by the SRI as a deviation (458/8581-03).
The licensee took inmediate
corrective action to add the verification of proper load sequencing in
Revision 1 to 1-ST-31 and to initiate a 100% review of other STPs to assure
implementation of regulatory requirements and commitments.
This review has
been completed by plant staff compliance with an overview by operations QA.
All questions generated from the review are being addressed by startup and
no acceptance criteria changes have resulted nor are any anticipated. All
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questions will be resolved prior to performance of the respective startup
test. All required actions are being taken to respond to the identified
deviation and this deviation is closed with no further response from the
licensee required.
Except for the one deviation noted, procedure 1-ST-31 appeared to meet all
applicable regulatory requirements and commitments.
4.
Startup Test Witness
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During this inspection period the SRI and RI witnessed startup testing
activities conducted under the startup testing program. The NRC inspectors
observed that:
(a) personnel conducting the test were cognizant of the
test acceptance criteria, precautions and prerequisites prior to beginning
the test; (b) the test was being conducted in accordance with an approved
procedure and the test procedure was being used and signed off by the
personnel conducting the test; and (c) data was being collected and
recorded as required. The NRC inspectors witnessed the following startup
tests:
1-ST-14
Reactor Core Isolation Cooling (RCIC)
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1-ST-26
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1-ST-27
Turbine Trip and Generator Load
Rejection
1-ST-31
The following observations were made during the performances of the above
startup tests.
a.
1-ST-14 Reactor Core Isolation Cooling:
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The NRC inspectors witnessed the performance of the RCIC system run in
the condensate storage tank (CST) to reactor vessel mode and RCIC pump
cold start at rated pressure on December 2 and 5, 1985, respectively.
Prior to beginning the RCIC system run on December 2, 1985, the NRC
inspectors reviewed S0P-0035, " Reactor Core Isolation Cooling," Revi-
sion 0, located in the main control roca, which was being used in
conjunction with ST-14. The review revealed that "laters" located in
the procedure had been removed by Temporary Change Notice (TCN)
85-1287, however, this TCN was not posted at the beginning of SOP-0035
as required by Administrative Procedure ADM-003, " Development, Control
and Use of Procedures," Revision 7.
Further review of S0P-0035
revealed that TCN-1287 also carrected a mistake in the electrical
line-up and added eight valves to the valve line-up sheet. This
matter was brought to the attention of the nuclear control operator
(NCO) stationed at the 601 control panel. On Decerbtr 3, 1985, the
NRC inspectors again reviewed 50P-0035 and noted thu'. TCN-1287 was not
posted in front of the procedure.
This document control problem was
identified by the NRC inspectors as an apparent violation (458/8581-01).
The NRC inspectors then brought this condition to the attention
of the shift supervisor who immediately initiated actions to review
all the procedures in the control room for similar problems.
b.
1-ST-26 Safety Relief Valves:
The SRI and RI witnessed the performance of 1-ST-26, " Safety Relief
Valves," during this inspection period. This startup test has been
completed and the established acceptance criteria appears to have been
met. The completed test package will be reviewed during a future NRC
inspection.
No violations or deviations were identified in this area of
inspection,
c.
1-ST-27 Turbine Trip and Generator Load Rejection:
The SRI observed a turbine trip performed on December 23, 1985, for
startup test 1-ST-27 " Turbine Trip and Generator Load Rejection."
Reactor power was at approximately 9% when the trip was initiated and
when the turbine stop valves closed, both turbine bypass valves opened
to control reactor pressure. No reactor scram occurred and reactor
pressure was controlled as required.
No violations or deviations were identified in this area of
inspection.
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d.
1-ST-31 Loss of Offsite Power:
The SRI observed the performance of startup test 1-ST-31, " Loss of
Offsite Power," on December 6,1985.
Reactor power was above 10% for
the test and when the loss of offsite power was initiated the
Division I, II and III diesel generators started and carried
safety-related electrical loads as required. The loss of offsite
power event was terminated as required by the procedure and no safety
relief valves lifted and reactor water level did not drop to the
point of initiating the high pressure core spray pump during the
event. After termination of the event, the reactor core isolation
cooling system was used to restore reactor water level to normal. An
initial review of the test data revealed that all control room
chillers did not start as required by the procedure.
Licensee
investigation of the failure of all chillers to start revealed that
one standby diesel loaded approximately 3 seconds earlier than the
other diesel which allowed chilled water flow to be established on
one division and the other division chillers did not start due to
chilled water flow already being established.
It appears that the
chillers performed as' designed and the licensee initiated a review to
determine if all chillers are needed during a loss of offsite power
event. The preliminary indication, based on heat load calculations
performed by S&W engineering, is that only one chiller is required
during the first 20 minutes of a loss of offsite power event and the
operator can start other chillers after 20 minutes if they are
needed. The licensee will evaluate the present design of the chiller
start logic and initiate changes if required.
The SRI will evaluate
the final disposition during the required review of the final test
package.
No violations or deviations were identified in this area of
inspection.
5.
Startup Test Program Quality Assurance (QA) Review
During this inspection period the RI began a review of the 1kensee's QA
audit and surveillance program for operational activities conducted under
the startup and power ascension test program. This initial review revealed
no problems and the program review will be completed during a future NRC
inspection.
No violations or deviations were identified in this area of ir.spection.
6.
Operational Safety Verification
The SRI and RI observed operational activities throughout the inspection
period and closely monitored operational events.
Significant operational
activities observed included several attempted turbine rolls which were
stopped due to high vibration, a successful completion of turbine roll and
initial synchronization of the main generator on December 3,1985. The
main generator was on-line for a short time (approximately 30 minutes) and
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a reverse power trip was received.
The problem was traced to a wiring
problem, which was corrected, and the generator was again synchronized on
December 4,1985, and maintained on-line with a generator load of
approximately 60 megawatts. The highest power level reached during this
inspection period was approximately 200 megawatts electrical. Also, on
December 31, 1985, a turbine trip occurred with reactor power at
approximately 20%.
Subsequent analysis of the trip revealed that a power
to load unbalance signal was received which caused the turbine control
valves to close and resulted in a turbine trip and subsequent reactor
scram on high pressure. The power to load unbalance signal apparently
occurred due to the combination of a failed pressure transmitter and an
electrical grid upset from loss of a 500 kilovolt line.
During this event
and other operational activities, operational staff actions were observed
to be well coordinated and efficient and the plant responded as expected.
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In addition to observing operational activities, the SRI reviewed licensee
generated condition reports during the inspection period. A review of two
condition reports (CRs 85-0559 and 85-0561) and the subsequent followup by
the SRI revealed that inadequate retest was performed following modifica-
tions/ repairs made on two separate occasions.
These two occasions were the
change out of wiring to the back-up scram valves with no documented
continuity checks or functional test performed and a wiring change to the
shutdown cooling suction valve (E12*F008) with no valve stroking performed
upon completion of the wiring modifications.
In both instances, no docu-
mented engineering-approved alternative testing was provided, and the func-
tion of the components was compromised by the modifications / repairs.
Subsequent review of this situation by the licensee revealed that both
conditions were complicated by either a design error or by unclear design
instructions. This design control problem is being documented by the
licensee QA organization.
A review by the SRI of the significance of the compromise of the function
of the components revealed that the back-up scram valves are mentioned in
the River Bend FSAR but they apparently are not taken credit for in any
accident analysis and the shutdown cooling suction valve was locked closed
per a commitment to the NRC and shutdown cooling operation was not compro-
mised while the wiring error was in place. However, the failure to perform
adequate retest on a safety-related/important to safety component repre-
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sents a serious breakdown in the QA program and this inadequate retest
following a modification or repair was identified by the SRI as an apparent
violation (458/8581-02).
7.
Site Tours
The SRI and RI toured areas of the site during the inspection period to
observe general work practices and gain knowledge of the facility.
No violations or deviations were identified in this area of inspection.
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8.
Exit Interview
An exit interview was conducted on January 10, 1986, with licensee
representatives (identified in paragraph 1). During this interview, the
SRI reviewed the scope and findings of the inspection.