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{{Adams | |||
| number = ML20203H770 | |||
| issue date = 07/24/1986 | |||
| title = Insp Rept 50-458/86-20 on 860501-0615.Violation Noted: Failure to Follow Administrative Procedures for Issue of Temporary Change Notices & Failure to Implement Procedures to Maintain Safety Sys Drawing Configuration | |||
| author name = Chamberlain D, Jaudon J, Jones W | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000458 | |||
| license number = | |||
| contact person = | |||
| case reference number = TASK-1.G.1, TASK-TM | |||
| document report number = 50-458-86-20, NUDOCS 8608050087 | |||
| package number = ML20203H757 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 15 | |||
}} | |||
See also: [[see also::IR 05000458/1986020]] | |||
=Text= | |||
{{#Wiki_filter:. _ __ _ . _ . | |||
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APPENDIX B | |||
, U. S. NUCLEAR REGULATORY COMMISSION | |||
< | |||
REGION IV | |||
NRC Inspection Report: 50-458/86-20 License: NPF-47 | |||
' | |||
Docket: 50-458 | |||
Licensee: Gulf 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: May 1 through June 15, 1986 | |||
Inspectors: { | |||
D. D. diamberlain, Senior Resident Inspector | |||
' | |||
Date | |||
(pars. 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11 and 12) | |||
% , ~ | |||
W. B.FJones, Resident inspector l | |||
b'h$b | |||
Date | |||
(pars. 1,2,3,4,5 6,7,8 9, and 11) | |||
/ | |||
' | |||
Approved: _ _ /w MYl 72 Y/[ | |||
J./.Jfudon,Ghiel,ProjectSectionA Dafe / | |||
Eeactor Proj& cts Branch | |||
8608050087 860730 | |||
, PDR ADOCK 05000458 | |||
G PDR | |||
't | |||
. . , - - ,- r__- ,_-.r,_m..-,.m-m_ _,.,_..,m , _y _, ,. . - - , , _ _ _ , , . , . ,, . . - _ | |||
. | |||
-2- | |||
Inspection Summary | |||
Inspection Conducted Iby 1 through June 15, 1986 (Report 50-458/86-20) | |||
Areas Inspected: Routine, unannouncad inspection of licensee action on | |||
previous inspection findings, status of operating license conditions, | |||
Nuclear Review Board activities, startup test witness, safety system | |||
walkdown, operational safety verification, maintenance witness, surveillance | |||
witness, licensee plans for coping with strikes and status of IMI action | |||
item. | |||
Results: Within the ten areas inspected, two violations were idantified | |||
(failure to follow administrative procedures for issue of temporary | |||
change notices, paragraph 9, and failure to implement procedures to | |||
maintain safety system drawing configuration, paragraph 6). | |||
_- - __ - ____-_ | |||
- | |||
o | |||
-3- | |||
DETAILS | |||
1. P_ersons Contacted | |||
Principal Licensee Employees | |||
M. Arant, Technician, Ir.strumentation and Control (I&C) | |||
*R. J. Backen, Supervisor (Acting), Operations Quality | |||
' Assurance (QA) | |||
W. J. Beck, Supervisor, Reactor Engineering | |||
*W.E H. Cahill, Jr. , Senior Vice President, River Bend | |||
Nuclear Group | |||
*E. M. Cargill, Supervisor, Radiation Programs | |||
*T. C. Crouse, Manager, QA | |||
*J. R. Cummings, Procedure Coordinator, | |||
*P. E. Freehill, Superintendent, Startup and Test | |||
A. O. Fredieu, Assistant Operations Superviser | |||
P. F. Gillespie, Senior Compliance Analyst | |||
D. R. Gipson, Assistant Plant Manager, Operations | |||
*E. R. Grant, Supervisor, Nuclear Licensing | |||
*B. R. Hall, Supervisor, Plant Services, | |||
*R. W. Helmick, Director, Projects, | |||
*G. K. Henry, Supervisor, Electrical Engineering i | |||
K. C. Hodges, Supervisor, Quality Systems | |||
*R. J. King, Licensing Engineer | |||
*A. D. Kowalczuk, Assistant Plant Manager, Maintenance | |||
*W. H. Odell. Manager, Administration | |||
*T. F. Plunkett, Plant Manager | |||
*S. R. Radebaugh, Assistant Plant Manager, Services | |||
W. J. Reed, Director, Nuclear Licensing | |||
D. Reynerson, Director, Nuclear Plant Engineering (NUPE) | |||
N. Simpson, Technician I&C | |||
*M. H. Small, Acting Supervisor, Operations Quality Control (QC) | |||
R. B. Stafford, Director, Operations QA | |||
*K. E. Suhrke, Manager, Projects | |||
^ P. c. Tomlinson, Director, Quality Services | |||
D. Williamson, Operations Supervisor | |||
The NRC senior resident inspector (SRI) and resident inspector (RI) also | |||
interviewed additional licensee personnel during the inspection period. | |||
* Denotes those persons that attended the exit interview conducted on | |||
June 19, 1986. NRC Region IV Section Chief, J. P. Jaudon, NRC resident | |||
inspector (RI), W. B. Jones and Nuclear Reactor Regulation (NRR) | |||
licensing Project Manager, S. Stern also attended the exit interview. | |||
. | |||
-4- | |||
2. Licensee Action on Previous Inspection Findings | |||
a. (Closed) Violation (458/8569-01): Failure of design document control | |||
program. | |||
i s | |||
This violation was a failure to post approved design changes against | |||
the effected design documents and a failure to distribute design | |||
change documents to document control stations. NUPE revised | |||
procedures NUPE-AA-54 and 59 to provide more control for the posting | |||
and routing of design change documents. A 100 percent audit of | |||
s design change files was conducted, and all noted discrepancies were | |||
- | |||
?' corrected. Training was conducted on the revised procedural | |||
requirements, and-subsequent quality assurance surveillances revealed | |||
no recurrence of.the problem. The SRI reviewed the revised | |||
- | |||
. , | |||
procedu'res and,the"other corrective actions. | |||
This, viol'ation is_ closed. | |||
~ | |||
~ | |||
_ b. (Closed) Violation (458/8569-02): Improper use of a field change | |||
, , notice. | |||
' | |||
NUPE issued procedure NUPE-AA-64, " Control and Approval of Field | |||
Change Notices (FCN's)" which provides detailed instructions and | |||
' | |||
restrictions for the use of FCNs. The licensee had conducted an | |||
audit of the design change files and the discrepancies noted had been | |||
corrected. Training of NUPE personnel responsible for completing | |||
FCN's had been completed. | |||
This violation is closed. | |||
i | |||
c. (Closed) Violation (458/8604-01): Failure to control temporary | |||
circuit alterations administratively. . | |||
The licensee actions in response to this violation included: a | |||
complete inspection of control room panels for unauthorized lifted | |||
leads or jumpers; QC hold points included in electrical maintenance | |||
work requests to inspect for proper restoration; implementation of a | |||
main control room cabinet access and work monitoring program; the | |||
change of control room panel locks; a maintenance procedure revision | |||
serialized tagging of any lifted lead or jumper for accountability | |||
and the temporary alterations program was suspended and replaced by | |||
design modification request procedures. The SRI has monitored | |||
licensee actions relative to temporary alterations, and the | |||
additional controls appear to be effective. | |||
! Ihis violation is closed. | |||
d. (Closed) Violation (458/8581-01): Failure to maintain a controlled | |||
copy of a temporary change notice (TCN) in front of the affected | |||
. controlled procedure in the control room Station Operating | |||
Manual (50M) | |||
_ . _ _ _ -- _- | |||
_ | |||
- _ _ . . . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . __ | |||
. | |||
, | |||
-5- | |||
' | |||
' | |||
<The licensee :took immediate corrective actions by performing a | |||
~ departmental review of all station operating procedures (SOPS), | |||
, , abnormal' operating procedures (A0Ps), and emergency operating | |||
- | |||
procedures (EOPs). . During the review,.the licensee identified | |||
several SOP's with duplicate copies of the same TCN and a few TCN's | |||
filed with the wrong S0P. These conditions were immediately | |||
corrected.' | |||
, | |||
m Responsibility for maintaining and ensuring that updates to the Main | |||
' Control- Room procedure manual are properly filed has been reassigned | |||
to Station Document Control (SDC). Periodic reviews of the Main | |||
Control Room SOMs are being conducted by the SDC in accordance with | |||
administration procedure ADM-005, " Station Document Control," | |||
Section 6.6. | |||
' | |||
_ | |||
This violation is closed. | |||
3. Status of Operating License Conditions | |||
Facility Operating License NPF-47 for River Bend Station was issued on | |||
November 20, 1985, and Attachment 1 to this license contains items which | |||
must be completed to the satisfaction of NRC Region IV. The following | |||
status is provided for the Attachment 1 license conditions: | |||
a. (Closed) License Condition 1.a.: Verify the station electric | |||
distribution voltage analyses are in accordance with the guidelines | |||
of Branch Technical Position PBS-1, Position 4, prior to completion | |||
of the initial test program. | |||
GSU has completed special situation test 1-SST-6, ." Bus Load Test," | |||
and the results were provided to Stone and Webster (S&W) for | |||
comparison to analytical model results. Memorandum S-CRB-9031 dated | |||
June 4, 1986, summarizes the results of that comparison and indicates | |||
that the test versus analytical results are acceptable with no test | |||
voltage drops more than 3 percent lower than the analytical values as | |||
recommended by Branch Technical Position PSB-1. | |||
This license condition is closed. | |||
b. (Closed) License Condition 1.b: Evaluate and complete modifications | |||
on battery powered lighting systems used in areas of the plant | |||
outside the main control room required for safe shutdown and | |||
personnel evacuation prior to completion of the initial test program. | |||
The RI reviewed the licensee's emergency lighting plan as detailed on | |||
Stone and Webster Lighting Plan drawings 12210-EE-65 thru 79. | |||
Emergency lighting stations were selected from the drawings and | |||
verified to be installed and operational for areas identified in | |||
Final Safety Analysis Report (FSAR) Table 9.5-2, " Illumination Level | |||
and Type of Fixtures used in Plant Areas Necessary for Safe Shutdown | |||
and Evacuation of Personnel." In addition, areas previously | |||
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-6- | |||
, | |||
identified as being deficient in illumination were selectively | |||
observed to meet the requirements of Table 9.5-2. As a result of | |||
this walkdown, one area was identified which did not meet the minimum | |||
illumination requirements. The area identified was an egress | |||
stairwell located on the east side of the turbine building between | |||
the elevation 95'0" and 123'6". The licensee initiated modification | |||
request (MR) 86-162 and maintenance work request (MWR) 41124 to | |||
install the emergency light. This work was completed on June 15, | |||
1986. | |||
This license condition is closed. | |||
4. Nuclear Review Board Activities | |||
The RI reviewed the Nuclear Review Board (NRB) minutes for the period | |||
February 1985 thru December 1985, to assess the overall effectiveness of | |||
the licensee's implementation of the off-site review committee. These | |||
minutes were evaluated against the NRB responsibilities outlined in | |||
Section 6.5.3 of the Technical Specifications (TS) and the NRB Manual. | |||
The RI noted during the above review, that the NRB has chartered four | |||
subcommittees to assist the NRB in fulfilling their responsibilities. | |||
These subcommittees are: | |||
o "Unreviewed Safety Questions Committee" (USQC); | |||
o " Quality Assurance Program Audit Committee" (QAPAC); | |||
o " Quality Concern Subcommittee;" and | |||
o "NRB/FRC Committee" | |||
The USQC was established to assist the NRB in meeting its responsibilities | |||
for reviewing proposed changes to the plant and its documentation to | |||
ensure that changes are not made which constitute an unreviewed safety | |||
question. Specifically, the USQC will review: | |||
o all safety evaluations for changes to procedures, equipment, systems | |||
or experiments which were determined not to involve unreviewed safety | |||
questions; | |||
o selected procedures, equipment, systems, tests and experiments which | |||
did not receive an evaluation to verify that they were properly | |||
classified and did not require a safety evaluation; | |||
o changes which were determined to be unreviewed safety questions and | |||
the associated changes to licensing documents; | |||
o. proposed changes to the Operating License or TS; and | |||
o violations of codes, regulations, orders, TS license requirements, | |||
procedures and instructions having nuclear safety significance. | |||
- | |||
. | |||
-7- | |||
The QAPAC was chartered to advise the NRB on the effectiveness of the | |||
' | |||
Quality Assurance Program. This is accomplished thru the QAPAC's | |||
participation in and review of audits performed by the QA audit group for | |||
the areas required by 6.5.3.8 of the TS. The requirement of the QAPAC to | |||
audit the Operational Quality Assurance Program every 24 months for | |||
compliance with 10 CFR part 50, Appendix B, is being fulfilled thru the | |||
Joint Utility Audit Group (JUAG). | |||
The Quality Concern Subcommittee receives and coordinates quality concern | |||
inquiries dealing with the operation of the quality assurance department | |||
and management of the River Bend Nuclear Group. This committee insures | |||
that proper action is taken by all departments to allegations or inquiries | |||
which are received. | |||
The NRB established the NRB/FRC Committee to monitor the FRC's activities. | |||
This Committee assesses the FRC's fulfillment of its responsibilities by: | |||
o reading all FRC minutes and other reports issued by the FRC; | |||
o occasional attendance at FRC meetings; | |||
o occasional verification of an FRC review item; and | |||
o semi-annual meetings to discuss FRC activities. | |||
The above subcommittees chartered by the NRB appear to be adequate to meet | |||
the function and responsibilities established in the Technical | |||
Specifications for the licensee's off-site review committee. The NRB | |||
meeting' minutes reviewed by the RI for the period February 1985 thru | |||
December 1985 demonstrate that the NRB members are cognizant of their | |||
responsibilities and have established programs to fulfill these | |||
responsibilities. The NRC inspectors will evaluate the effectiveness of | |||
the NRB during future inspections. | |||
No violations or deviations were identified in this area of the | |||
inspection. | |||
5. Startup Test Witness | |||
During this inspection period, the SRI and RI witnessed startup testing | |||
conducted under the startup testing program. The NRC inspectors observed | |||
that: personnel conducting the test were cognizant of the test acceptance | |||
criteria, precautions and prerequisites prior to beginning the test; the | |||
test was conducted in accordance with an approved procedure; the test | |||
procedure was being used and signed off by the personnel conducting the | |||
test; and data were being collected and recorded as required. The NRC | |||
inspectors witnessed the following startup tests: | |||
' | |||
. | |||
-8- | |||
o 1-ST-27 Turbine Trip and Generator Load Reject | |||
o 1-ST-28 Shutdown from Outside the Control Room | |||
o 1-ST-25B Main Steam Isolation Valve (MSIV) Full | |||
Closure | |||
o 1-ST-19 Core Performance | |||
The following observations were made during the performance of the above | |||
startup tests: | |||
o Test 1-ST-27 Turbine Trip and Generator Load Reject: | |||
The SRI and RI observed the performance of Section 6.3, "High Power | |||
Generator Load Rejection" to startup test 1-ST-27, " Turbine Trip and | |||
Generator Load Rejection" on May 29, 1986. The reactor was at | |||
approximately 96 percent thermal power when a generator load | |||
rejection was initiated by tripping a generator differential relay. | |||
This caused both generator output breakers to open and a turbine | |||
control valve fast closure (TCVFC) to occur. The reactor scrammed, | |||
as expected, when the TCVFC signal was initiated. The peak reactor | |||
pressure reached during the transient was 1106 psig. The NRC | |||
inspectors noted that the bypass valves opened along with nine safety | |||
relief valves (SRV) to reduce and control reactor pressure. | |||
Following the initial SRV blowdown, only 1 SRV was observed to | |||
reopen, which is consistent with the test acceptance criteria. The | |||
licensee is presently evaluating SRVs B21*F041D and B21*F041F which | |||
apparently opened briefly in their safety mode. The results of this | |||
evaluatior, will be reviewed by the NRC inspectors during their review | |||
of 1-ST-27 test results. The licensee has collected test data to | |||
evaluate for conformance to acceptance criteria. | |||
No violations or deviations were identified in this area of | |||
inspection. | |||
o 1-ST-28 Shutdown from Outside the Control Room: | |||
The SRI and RI observed the performance of Section 6.3, " Cold | |||
Shutdown from Outside the Control Roem", on May 30, 1986. Following | |||
the reactor scram initiated during 1-ST-27, reactor pressure was | |||
reduced to 120 psig from the main control room (MCR). The licensee | |||
had previously demonstrated the ability to scram and maintain the | |||
reactor in hot shutdown from outside the control room on February 15, | |||
1986. With reactor pressure at 120 psig, control of the Division I | |||
residual heat removal (RHR) system was transferred from the MCR to | |||
the remote shutdown panel (RSP). When the nucleer control operators | |||
(NCO) began to realign the RHR system from the low pressure coolant | |||
injection (LPCI) mode to the RHR mode, the suppression pool suction | |||
valve 1E12*F004A indication failed in the intermediate position. | |||
- _ | |||
. . _ . ._. . - - __ | |||
. _ _ _ . _ . . _ _ . . . __ _ _ . _ _ _ _ | |||
. | |||
-9- | |||
l | |||
After verifying locally that 1E12*F004A was closed, the NCO attempted | |||
to open 1E12*F006A valve as required to establish shut down cooling. | |||
This attempt failed however because of the interlock that prevents | |||
4 -the F006A valve from opening if the F004A is not in the closed | |||
. (indicated) position. Control of the division II RHR system was then | |||
transferred to the division II RSP and the system aligned to the shut -r | |||
down cooling mode without incident. The NCO established a cooldown | |||
rate of approximately 80 F per hour with the heat transfer path to | |||
the' standby, service water system. The licensee has collected the | |||
test data to evaluate for conformance to acceptance criteria. | |||
, | |||
' | |||
% violation or deviations were observed in this area of inspection. | |||
* ' | |||
, | |||
. | |||
, | |||
; | |||
- | |||
,o' ' | |||
.T,est 1-ST-25B Main Steam Isolation Valve (MSIV) Full Closure: | |||
. | |||
2 | |||
, | |||
Th'e RI witnessed the pefformance of startup test 1-ST-25B, "MSIV Full | |||
Closure", on-June 8,1986 with reactor power at 100 percent and rated | |||
: ., . core f, low at 96 percent. The licensee initiated the test at | |||
1733 hours.by simulating a loss of condenser vacuum which results in | |||
the MSIVs closing when in the run mode. Upon initiation of the MSIV | |||
full closure, the reactor tripped and the SRVs opened momentarily to | |||
control pressure. Following the initial opening of the SRVs, only | |||
' | |||
, 1 SRV was observed to reopen. The high pressure core spray (HPCS) | |||
and reactor core isolation cooling (RCIC) systems initiated on | |||
reactor vessel water level reaching the level 2 setpoint. The | |||
reactor vessel water level recovered quickly because of the | |||
subsequent swell of vessel water and the continuous feed from the | |||
reactor feed water pumps. The NC0 secured the HPCS injection valve | |||
prior to the system injecting based on the rising vessel water level. , | |||
Approximately 3 minutes into the transmit, the reactor feed pumps | |||
tripped on vessel high level. The reactor feed water B pump was | |||
subsequently restarted and vessel level maintained within the normal | |||
band. Control room conduct during the test was observed to be well | |||
coordinated and efficient. The licensee has collected the test data | |||
to evaluate for conformance to acceptance criteria. | |||
" | |||
No violations or deviations were identified in this area of | |||
inspection, | |||
i | |||
! o Test 1-ST-19 Core Performance | |||
The SRI witnessed the running of traversing incore probe (TIP) traces | |||
in preparation for computer calculation of reactor power / thermal | |||
limits. The required data was extracted and inserted into startup | |||
; test procedure 1-ST-19, " Core Performance" by the licensee for test | |||
; condition (TC) 6 verification of core thermal limits and core thermal | |||
i power. A preliminary review of the test data by the SRI revealed | |||
i- | |||
that core thermal limits were well within the TS limits and core | |||
thermal power was approximately 99.3 percent of rated thermal power. | |||
1 | |||
. | |||
i | |||
. _ _ _ _ , . _ . , , , , _ , m . _ , _ _ _ _ . _ . _ . . .,.-,,...,_,_...,,_,_,,,.,,m._,,,_,,_._,_, - | |||
,m.__~_. .m-- - _- | |||
l | |||
* - | |||
* . ,. | |||
-10- | |||
L | |||
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* | |||
, | |||
No violations or deviations were identified in this area of | |||
. inspection. | |||
, | |||
6., Safety System Walkdown. | |||
l'Duringthisinspebtionperiod,theSRIandRIperformedawalkdownofthe | |||
' | |||
"C",RHR system to verify proper system alignment for operability as | |||
required by TS for Operational Conditions 1, 2, 3, 4, and 5. It was | |||
observed that: | |||
, | |||
o sy' stem valves were properly aligned; | |||
o abnormal control-room instrumentation readings or alarms were | |||
- | |||
present; | |||
o no leakage from major components was present; | |||
o the "C" RHR pump upper and lower bearing oil reservoirs were properly | |||
filled; and | |||
o accessible hangers and supports were intact. | |||
No anomalies were noted that would have affected "C" RHR system | |||
operability for low pressure coolant injection (LPCI). However, certain | |||
discrepancies were noted when comparing the system condition / lineup with | |||
the engineering piping and instrument diagrams (PIDs). It was noted that | |||
pipe caps were missing from five vent and drain locations where PID-27-7C | |||
indicated caps were installed; valve E12*MOVF064C which was open showed | |||
closed on PID-27-7C; and valve E12*VF063C was not locked as shown on | |||
PID-4-3C. Also, several valves were locked closed although the drawings | |||
did not indicate locks installed. The failure to implement procedures to | |||
maintain a safety system condition / lineup as shown on design output | |||
drawings or to modify the design output drawings to reflect the required | |||
system condition was identified by the SRI as an apparent violation | |||
(458/8620-02). Subsequent discussions with licensee management revealed | |||
that minimum flow valve E12*MOVF064C is open for standby operation and | |||
condensate fill and flush valve E12*VF063C (not a major flow path valve) | |||
had been identified as unlocked during an operations review of all valves | |||
on the locked valve list. It had been noted that the valve was unlocked | |||
and inaccessible without a scaffold or ladder. No action had been taken | |||
to gain access to the valve to install a lock. The licensee took | |||
immediate action to obtain ladders, and the valve was locked as shown on | |||
PID-4-3C when the SRI identified that the valve was not locked. It was | |||
also noted that the system operating procedure valve lineup did not | |||
indicate that valve E12*VF063C was to be locked closed as show on the PID. | |||
In response to the identified violation, the licensee should address how | |||
they will assure that PID drawing, Stone and Webster flow diagrams, system | |||
operating procedure valve lineups, locked valve lists and actual system | |||
configuration are consistent for an identified system operational | |||
condition and they should identify procedural controls which allow | |||
deviation from drawing requirements. For example, procedures may allow | |||
- | |||
, | |||
s a-- | |||
.. | |||
, | |||
. | |||
-11- | |||
the locking of more valves than shown on the drawings at the discretion of | |||
the operations staff. This issue should also be addressed for other | |||
drawings / documents used routinely by operations and/or maintenance for | |||
performance of work activities. | |||
7. Operational Safety Verification | |||
The SRI and RI observed operational activities throughout the inspection | |||
period and closely monitored operational events. Control room activities | |||
and conduct were observed to be well controlled and efficient. Proper | |||
control room staffing was maintained and access to the control room | |||
operational area was controlled. The licensee was adhering to limiting | |||
conditions for operation (LCO) as they occurred. Operators were | |||
questioned regarding lit annunciators and they understood why the | |||
annunciators were lit in all cases. Selected shift turnover meetings were | |||
observed, and all necessary information concerning plant status was | |||
apparently being covered in these meetings. A walkdown of the "C" RHR | |||
system was conducted, and the valves were observed to be in the proper | |||
position for standby operation. Several plant tours were conducted and | |||
overall plant cleanliness was good. During these plant tours, radiation | |||
protection area postings were observed to be accurate. | |||
During this inspection period the licensee completed a 100 hour | |||
verification run at full power operation, and all planned startup testing | |||
was completed. | |||
No violations or deviations were identified in this area of inspection. | |||
8. Maintenance Witness | |||
During this inspection period, the RI observed portions of selected | |||
corrective maintenance activities to verify that maintenance activities | |||
are being conducted in accordance with approved procedures, TS and | |||
appropriate industrial codes and standards. The RI verified through | |||
direct observation and review of records that: | |||
o maintenance activities did not violate LCOs; | |||
o redundant components were available; | |||
o required administrative approvals and tagouts were obtained before | |||
initiating work; | |||
o procedures were adequate to control the work; | |||
o radiological controls were properly implemented where applicable; | |||
o QC hold points were established and observed; and | |||
o replacement parts and materials used were properly certified. | |||
, o | |||
-12- | |||
The_following two corrective maintenance activities were observed: | |||
o Control Rod Drive (CRO) Cooling Water Check Valve: | |||
On May 7, 1986, the licensee experienced a failure of control | |||
rod 24-33 to insert or withdraw during control rod manipulations. | |||
Trouble shooting of the hydraulic control unit (HCU) revealed that | |||
CRD cooling water check valve C11-V138 was not reseating when drive | |||
water was applied to HCU 24-33. This condition allowed the drive | |||
water to flow back thru the cooling water line and thus the necessary | |||
lift was not being provided to the CRD drive piston to insert the rod | |||
or retract the collet finger to allow withdraw of the control rod. | |||
This condition would not have prevented the control rod from | |||
inserting during a reactor scram. | |||
~The. licensee initiated prompt maintenance work request (MWR) 39099 to | |||
clean', inspect and replace if necessary the ball checks to CRD | |||
cooling water check valve C11-V138 on HCU 24-33. The RI verified | |||
prior to initiating work that the MWR had been properly initiated; QC | |||
notification points were established, the job plan was appropriate to | |||
control the work; and a job briefing had been performed as evidenced | |||
by maintenance personnel signatures on the job briefing sheet. In | |||
addition, the RI verified the requirements of T.S. 3.1.3.1 for an | |||
inoperable control = rod were being complied with. | |||
Prior to in'itiating work, the nuclear equipment operator (NE0) | |||
obtained the shift supervisors permission to isolate HCU 24-33. The | |||
, ' HCU was then isolated and tagged out using clearance number 86-192, | |||
and.the ball was removed from the check valve. Inspection of the | |||
ball revealed surface scratches and abrasions. Subsequent flushings | |||
of the cooling water line produced a 1/2" x 1/16" round metal sliver. | |||
' | |||
A new ball was then installed in the check valve and the system | |||
. verified operable using surveillance test procedure STP-052-0101, | |||
" Control Rod Movement Operability Check." LC0 86-394, which was | |||
initiated to track the action requirements of T.S. 3.1.3.1 was closed | |||
based on the acceptable performance of STP-052-0101. | |||
No violations or deviations were identified in this area of | |||
inspection, | |||
o Division II Emergency Diesel Generator: | |||
During the performance of surveillance test STP-309-0202," Diesel | |||
Generator Division II Operability Test," on June 5,1986, the " ready | |||
to load" light did not illuminate after the diesel generator achieved | |||
rated frequency and voltage. The licensee noted this condition and | |||
immediately shut down the diesel. The failure of the " ready to load | |||
light" to illuminate indicates that the standby generator breaker | |||
1 ENS *ACB27 would not have closed onto standby buss IENS*SW618. | |||
Prompt MWR 41549 was then initiated to trouble shoot and restore the | |||
ready to load circuit to operable status. The RI reviewed the MWR | |||
. * | |||
. | |||
-13- | |||
prior to the initiation of work and verified that a deficiency tag | |||
had been placed and the applicable LC0 initiated. The appropriate | |||
hold points were placed in the procedure and a quality control | |||
representative was present during the performance of this MWR. The | |||
RI verified that the lifted leads were identified in the Lifted Lead | |||
and Jumper Log as tag, numbers 86-3101-001 thru 006 and that the | |||
, restored leads were independently verified. | |||
The licensee identified-that relays EGS*UVRA-UVRB needed the pick up | |||
voltages adjusted to within the setpoint limits. This work was | |||
subsequently performed under MWR 41549, and the diesel restored to | |||
operable statuu at 2041 hrs after the successful completion of | |||
surveillance STP-309-0202. LC0 86-472 was then cancelled at | |||
2045 hours. | |||
No violations'or deviations were identified in this area of | |||
inspection. | |||
9. Surveillance Witness | |||
The SRI and RI witnessed surveillance testing conducted by the licensee | |||
during this inspection period and the following observations were made: | |||
o Surveillance Test STP-051-4210: | |||
The SRI witnessed a portion of the instrumentation | |||
surveillance STP-051-4210, "RPS/RHR Reactor Vessel Steam Dome | |||
Pressure - High, Monthly Chfunct, 18 Month Chcal, and 18 Month LSFT | |||
(821-N078B,B21-6798)" conducted on May 28, 1986. The portion of the | |||
test observed was known to cause a half scram signal and the | |||
technicians were cautious in verifying the other division was not | |||
tripped prior to test performance. They also limited the time that | |||
the half scram was allowed to be in by close coordination with the | |||
operations staff. During the performance of this test it was noted | |||
that there were more than one copy of certain pages of the procedure | |||
in the official work copy. This created some confusion during test | |||
performance. The SRI discussed this with instrumentation maintenance | |||
supervision and it was discovered that there were seven open TCNs | |||
against this procedure and the preparer of the last TCN (No. 86-0581) | |||
had failed to use copies of previous permanent TCN pages to markup | |||
for the new TCN as required by administrative procedure. This | |||
resulted in the official work copy of the procedure containing three | |||
page eights and three page elevens with a different TCN number on | |||
each page. This failure to follow administrative procedures for | |||
issue of TCNs was identified by the SRI as an apparent violation | |||
(458/8620-01). | |||
The SRI also discussed the status of incorporation of TCNs in | |||
procedure revisions in response to a previous NRC violation and it | |||
was found that there are a total of 38 procedures with more than | |||
three open TCNs out of a total procedure population of | |||
. - - .. - .-- - . | |||
,, . | |||
-14- | |||
approximately 3500. The SRI requested a date from licensee | |||
management for revision of these 38 procedures to incorporate the | |||
'open TCNs and management stated that these procedures would be issued | |||
* | |||
by September 1, 1986. | |||
o Surveillance Test STP-309-0202: | |||
The RI observed the performance of surveillance STP-309-0202, " Diesel | |||
Generator Division II Operability Test," Revision 5 on June 5, 1986, | |||
.with the plant in operational condition 1. This surveillance is | |||
' | |||
. designed to demonstrate the operability of the Division II diesel | |||
, | |||
generator and satisfie's T.S. Sections 4.8.1.1.2.a.1 through | |||
i 4.8.1.-l.2.a.7;and 4.8.1.1.2.c.1 and c.2. | |||
. | |||
- | |||
) Prior to' initiating the test, communications were established between | |||
' | |||
the NE0 at the remote diesel panel and the NC0 in the Main Control | |||
i Room. LC0 86-472 had previously been initiated in accordance with | |||
! T.S. 3.8.1.1.b because of the preplanned preventive maintenance (PM) | |||
' | |||
which had been performed on this diesel generator prior to beginning | |||
'STP-309-0202. Upon initiation of the STP, the diesel generator was | |||
observed to attain rated frequency and voltage within the required 10 | |||
seconds, however the " ready to load" light did not illuminate at | |||
either the remote panel or in the main control room. The diesel | |||
generator was shutdown within 30 seconds of its starting and prompt | |||
MWR 41549 initiated. The failure of the " ready to load" light to | |||
illuminate indicates that the standby generator breaker IENS*ACB27 | |||
would not have closed onto standby bus IENS*SW618. After verifying | |||
the diesel generator had failed its surveillance test, the licensee | |||
initiated the actions required by T.S. 3.8.1.1.b for an inoperable | |||
diesel generator, because of any cause other than the performance of | |||
preplanned preventive maintenance. | |||
MWR 41549 was completed at 2015 hours on June 5, 1986, and | |||
STP-309-0202 was satisfactorily performed. LC0 86-472 was then | |||
: closed based on the diesel generator satisfying the operability test. | |||
No violations or deviations were identified in area of the inspection. | |||
i 10. Licensee Plans for Coping With Strikes | |||
, | |||
The SRI reviewed licensee plans for coping with strikes during this | |||
inspection period. It was found that the licensee had addressed such | |||
issues as personnel / training requirements, security support requirements, | |||
> | |||
offsite support, etc. This area will be reviewed further during future | |||
NRC inspections as required. | |||
No violations or deviations were identified in this area of inspection. | |||
: | |||
1 | |||
! | |||
! | |||
. . . . | |||
-15- | |||
11. Status of TMI Action Item | |||
(Closed) NUREG-0737, Item I.G.1: Trair;ing during initial startup test | |||
. phase. | |||
The licensee has completed the initial startup test program training | |||
requirements as described in NUREG-0737, Item I.G.1 and FSAR | |||
Section 14.2.3 Training During Initial Startup Test Phase. The startup | |||
test training program was established to assure that personnel from each | |||
of the six operating shift crews: | |||
o observed a reactor scram; | |||
o observed a pressure regulator transient; | |||
o observed a water level setpoint transient; | |||
o operated the operation of RCIC system; and | |||
o observed a turbine trip or load rejection. | |||
The startup test program was balanced, as much as practical, between the | |||
six shif ts to assure that each shift was exposed to the above, off-normal | |||
events. This has resulted in each shift having at least four individuals | |||
who have observed actual plant responses for each of these events. Based | |||
on the shift experience that now exists for each shift crew, no further | |||
licensee action regarding NUREG-0737, Item I.G.1 is necessary. | |||
This TMI action item is closed. | |||
12. Exit and Inspection Interview | |||
An exit interview was conducted on June 19, 1986, with licensee | |||
representatives (identified in paragraph 1). During this interview, the | |||
SRI reviewed the scope and findings of the inspection. | |||
}} |
Latest revision as of 11:14, 7 December 2021
ML20203H770 | |
Person / Time | |
---|---|
Site: | River Bend |
Issue date: | 07/24/1986 |
From: | Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
To: | |
Shared Package | |
ML20203H757 | List: |
References | |
TASK-1.G.1, TASK-TM 50-458-86-20, NUDOCS 8608050087 | |
Download: ML20203H770 (15) | |
See also: IR 05000458/1986020
Text
. _ __ _ . _ .
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,
APPENDIX B
, U. S. NUCLEAR REGULATORY COMMISSION
<
REGION IV
NRC Inspection Report: 50-458/86-20 License: NPF-47
'
Docket: 50-458
Licensee: Gulf 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: May 1 through June 15, 1986
Inspectors: {
D. D. diamberlain, Senior Resident Inspector
'
Date
(pars. 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11 and 12)
% , ~
W. B.FJones, Resident inspector l
b'h$b
Date
(pars. 1,2,3,4,5 6,7,8 9, and 11)
/
'
Approved: _ _ /w MYl 72 Y/[
J./.Jfudon,Ghiel,ProjectSectionA Dafe /
Eeactor Proj& cts Branch
8608050087 860730
, PDR ADOCK 05000458
G PDR
't
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Inspection Summary
Inspection Conducted Iby 1 through June 15, 1986 (Report 50-458/86-20)
Areas Inspected: Routine, unannouncad inspection of licensee action on
previous inspection findings, status of operating license conditions,
Nuclear Review Board activities, startup test witness, safety system
walkdown, operational safety verification, maintenance witness, surveillance
witness, licensee plans for coping with strikes and status of IMI action
item.
Results: Within the ten areas inspected, two violations were idantified
(failure to follow administrative procedures for issue of temporary
change notices, paragraph 9, and failure to implement procedures to
maintain safety system drawing configuration, paragraph 6).
_- - __ - ____-_
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o
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DETAILS
1. P_ersons Contacted
Principal Licensee Employees
M. Arant, Technician, Ir.strumentation and Control (I&C)
- R. J. Backen, Supervisor (Acting), Operations Quality
' Assurance (QA)
W. J. Beck, Supervisor, Reactor Engineering
- W.E H. Cahill, Jr. , Senior Vice President, River Bend
Nuclear Group
- E. M. Cargill, Supervisor, Radiation Programs
- T. C. Crouse, Manager, QA
- J. R. Cummings, Procedure Coordinator,
- P. E. Freehill, Superintendent, Startup and Test
A. O. Fredieu, Assistant Operations Superviser
P. F. Gillespie, Senior Compliance Analyst
D. R. Gipson, Assistant Plant Manager, Operations
- E. R. Grant, Supervisor, Nuclear Licensing
- B. R. Hall, Supervisor, Plant Services,
- R. W. Helmick, Director, Projects,
- G. K. Henry, Supervisor, Electrical Engineering i
K. C. Hodges, Supervisor, Quality Systems
- R. J. King, Licensing Engineer
- A. D. Kowalczuk, Assistant Plant Manager, Maintenance
- W. H. Odell. Manager, Administration
- T. F. Plunkett, Plant Manager
- S. R. Radebaugh, Assistant Plant Manager, Services
W. J. Reed, Director, Nuclear Licensing
D. Reynerson, Director, Nuclear Plant Engineering (NUPE)
N. Simpson, Technician I&C
- M. H. Small, Acting Supervisor, Operations Quality Control (QC)
R. B. Stafford, Director, Operations QA
- K. E. Suhrke, Manager, Projects
^ P. c. Tomlinson, Director, Quality Services
D. Williamson, Operations Supervisor
The NRC senior resident inspector (SRI) and resident inspector (RI) also
interviewed additional licensee personnel during the inspection period.
- Denotes those persons that attended the exit interview conducted on
June 19, 1986. NRC Region IV Section Chief, J. P. Jaudon, NRC resident
inspector (RI), W. B. Jones and Nuclear Reactor Regulation (NRR)
licensing Project Manager, S. Stern also attended the exit interview.
.
-4-
2. Licensee Action on Previous Inspection Findings
a. (Closed) Violation (458/8569-01): Failure of design document control
program.
i s
This violation was a failure to post approved design changes against
the effected design documents and a failure to distribute design
change documents to document control stations. NUPE revised
procedures NUPE-AA-54 and 59 to provide more control for the posting
and routing of design change documents. A 100 percent audit of
s design change files was conducted, and all noted discrepancies were
-
?' corrected. Training was conducted on the revised procedural
requirements, and-subsequent quality assurance surveillances revealed
no recurrence of.the problem. The SRI reviewed the revised
-
. ,
procedu'res and,the"other corrective actions.
This, viol'ation is_ closed.
~
~
_ b. (Closed) Violation (458/8569-02): Improper use of a field change
, , notice.
'
NUPE issued procedure NUPE-AA-64, " Control and Approval of Field
Change Notices (FCN's)" which provides detailed instructions and
'
restrictions for the use of FCNs. The licensee had conducted an
audit of the design change files and the discrepancies noted had been
corrected. Training of NUPE personnel responsible for completing
FCN's had been completed.
This violation is closed.
i
c. (Closed) Violation (458/8604-01): Failure to control temporary
circuit alterations administratively. .
The licensee actions in response to this violation included: a
complete inspection of control room panels for unauthorized lifted
leads or jumpers; QC hold points included in electrical maintenance
work requests to inspect for proper restoration; implementation of a
main control room cabinet access and work monitoring program; the
change of control room panel locks; a maintenance procedure revision
serialized tagging of any lifted lead or jumper for accountability
and the temporary alterations program was suspended and replaced by
design modification request procedures. The SRI has monitored
licensee actions relative to temporary alterations, and the
additional controls appear to be effective.
! Ihis violation is closed.
d. (Closed) Violation (458/8581-01): Failure to maintain a controlled
copy of a temporary change notice (TCN) in front of the affected
. controlled procedure in the control room Station Operating
Manual (50M)
_ . _ _ _ -- _-
_
- _ _ . . . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . __
.
,
-5-
'
'
<The licensee :took immediate corrective actions by performing a
~ departmental review of all station operating procedures (SOPS),
, , abnormal' operating procedures (A0Ps), and emergency operating
-
procedures (EOPs). . During the review,.the licensee identified
several SOP's with duplicate copies of the same TCN and a few TCN's
filed with the wrong S0P. These conditions were immediately
corrected.'
,
m Responsibility for maintaining and ensuring that updates to the Main
' Control- Room procedure manual are properly filed has been reassigned
to Station Document Control (SDC). Periodic reviews of the Main
Control Room SOMs are being conducted by the SDC in accordance with
administration procedure ADM-005, " Station Document Control,"
Section 6.6.
'
_
This violation is closed.
3. Status of Operating License Conditions
Facility Operating License NPF-47 for River Bend Station was issued on
November 20, 1985, and Attachment 1 to this license contains items which
must be completed to the satisfaction of NRC Region IV. The following
status is provided for the Attachment 1 license conditions:
a. (Closed) License Condition 1.a.: Verify the station electric
distribution voltage analyses are in accordance with the guidelines
of Branch Technical Position PBS-1, Position 4, prior to completion
of the initial test program.
GSU has completed special situation test 1-SST-6, ." Bus Load Test,"
and the results were provided to Stone and Webster (S&W) for
comparison to analytical model results. Memorandum S-CRB-9031 dated
June 4, 1986, summarizes the results of that comparison and indicates
that the test versus analytical results are acceptable with no test
voltage drops more than 3 percent lower than the analytical values as
recommended by Branch Technical Position PSB-1.
This license condition is closed.
b. (Closed) License Condition 1.b: Evaluate and complete modifications
on battery powered lighting systems used in areas of the plant
outside the main control room required for safe shutdown and
personnel evacuation prior to completion of the initial test program.
The RI reviewed the licensee's emergency lighting plan as detailed on
Stone and Webster Lighting Plan drawings 12210-EE-65 thru 79.
Emergency lighting stations were selected from the drawings and
verified to be installed and operational for areas identified in
Final Safety Analysis Report (FSAR) Table 9.5-2, " Illumination Level
and Type of Fixtures used in Plant Areas Necessary for Safe Shutdown
and Evacuation of Personnel." In addition, areas previously
__
,
, -
.
,
. . . .
-
.. , -
. .
,
-6-
,
identified as being deficient in illumination were selectively
observed to meet the requirements of Table 9.5-2. As a result of
this walkdown, one area was identified which did not meet the minimum
illumination requirements. The area identified was an egress
stairwell located on the east side of the turbine building between
the elevation 95'0" and 123'6". The licensee initiated modification
request (MR)86-162 and maintenance work request (MWR) 41124 to
install the emergency light. This work was completed on June 15,
1986.
This license condition is closed.
4. Nuclear Review Board Activities
The RI reviewed the Nuclear Review Board (NRB) minutes for the period
February 1985 thru December 1985, to assess the overall effectiveness of
the licensee's implementation of the off-site review committee. These
minutes were evaluated against the NRB responsibilities outlined in
Section 6.5.3 of the Technical Specifications (TS) and the NRB Manual.
The RI noted during the above review, that the NRB has chartered four
subcommittees to assist the NRB in fulfilling their responsibilities.
These subcommittees are:
o "Unreviewed Safety Questions Committee" (USQC);
o " Quality Assurance Program Audit Committee" (QAPAC);
o " Quality Concern Subcommittee;" and
o "NRB/FRC Committee"
The USQC was established to assist the NRB in meeting its responsibilities
for reviewing proposed changes to the plant and its documentation to
ensure that changes are not made which constitute an unreviewed safety
question. Specifically, the USQC will review:
o all safety evaluations for changes to procedures, equipment, systems
or experiments which were determined not to involve unreviewed safety
questions;
o selected procedures, equipment, systems, tests and experiments which
did not receive an evaluation to verify that they were properly
classified and did not require a safety evaluation;
o changes which were determined to be unreviewed safety questions and
the associated changes to licensing documents;
o. proposed changes to the Operating License or TS; and
o violations of codes, regulations, orders, TS license requirements,
procedures and instructions having nuclear safety significance.
-
.
-7-
The QAPAC was chartered to advise the NRB on the effectiveness of the
'
Quality Assurance Program. This is accomplished thru the QAPAC's
participation in and review of audits performed by the QA audit group for
the areas required by 6.5.3.8 of the TS. The requirement of the QAPAC to
audit the Operational Quality Assurance Program every 24 months for
compliance with 10 CFR part 50, Appendix B, is being fulfilled thru the
Joint Utility Audit Group (JUAG).
The Quality Concern Subcommittee receives and coordinates quality concern
inquiries dealing with the operation of the quality assurance department
and management of the River Bend Nuclear Group. This committee insures
that proper action is taken by all departments to allegations or inquiries
which are received.
The NRB established the NRB/FRC Committee to monitor the FRC's activities.
This Committee assesses the FRC's fulfillment of its responsibilities by:
o reading all FRC minutes and other reports issued by the FRC;
o occasional attendance at FRC meetings;
o occasional verification of an FRC review item; and
o semi-annual meetings to discuss FRC activities.
The above subcommittees chartered by the NRB appear to be adequate to meet
the function and responsibilities established in the Technical
Specifications for the licensee's off-site review committee. The NRB
meeting' minutes reviewed by the RI for the period February 1985 thru
December 1985 demonstrate that the NRB members are cognizant of their
responsibilities and have established programs to fulfill these
responsibilities. The NRC inspectors will evaluate the effectiveness of
the NRB during future inspections.
No violations or deviations were identified in this area of the
inspection.
5. Startup Test Witness
During this inspection period, the SRI and RI witnessed startup testing
conducted under the startup testing program. The NRC inspectors observed
that: personnel conducting the test were cognizant of the test acceptance
criteria, precautions and prerequisites prior to beginning the test; the
test was conducted in accordance with an approved procedure; the test
procedure was being used and signed off by the personnel conducting the
test; and data were being collected and recorded as required. The NRC
inspectors witnessed the following startup tests:
'
.
-8-
o 1-ST-27 Turbine Trip and Generator Load Reject
o 1-ST-28 Shutdown from Outside the Control Room
o 1-ST-25B Main Steam Isolation Valve (MSIV) Full
Closure
o 1-ST-19 Core Performance
The following observations were made during the performance of the above
startup tests:
o Test 1-ST-27 Turbine Trip and Generator Load Reject:
The SRI and RI observed the performance of Section 6.3, "High Power
Generator Load Rejection" to startup test 1-ST-27, " Turbine Trip and
Generator Load Rejection" on May 29, 1986. The reactor was at
approximately 96 percent thermal power when a generator load
rejection was initiated by tripping a generator differential relay.
This caused both generator output breakers to open and a turbine
control valve fast closure (TCVFC) to occur. The reactor scrammed,
as expected, when the TCVFC signal was initiated. The peak reactor
pressure reached during the transient was 1106 psig. The NRC
inspectors noted that the bypass valves opened along with nine safety
relief valves (SRV) to reduce and control reactor pressure.
Following the initial SRV blowdown, only 1 SRV was observed to
reopen, which is consistent with the test acceptance criteria. The
licensee is presently evaluating SRVs B21*F041D and B21*F041F which
apparently opened briefly in their safety mode. The results of this
evaluatior, will be reviewed by the NRC inspectors during their review
of 1-ST-27 test results. The licensee has collected test data to
evaluate for conformance to acceptance criteria.
No violations or deviations were identified in this area of
inspection.
o 1-ST-28 Shutdown from Outside the Control Room:
The SRI and RI observed the performance of Section 6.3, " Cold
Shutdown from Outside the Control Roem", on May 30, 1986. Following
the reactor scram initiated during 1-ST-27, reactor pressure was
reduced to 120 psig from the main control room (MCR). The licensee
had previously demonstrated the ability to scram and maintain the
reactor in hot shutdown from outside the control room on February 15,
1986. With reactor pressure at 120 psig, control of the Division I
residual heat removal (RHR) system was transferred from the MCR to
the remote shutdown panel (RSP). When the nucleer control operators
(NCO) began to realign the RHR system from the low pressure coolant
injection (LPCI) mode to the RHR mode, the suppression pool suction
valve 1E12*F004A indication failed in the intermediate position.
- _
. . _ . ._. . - - __
. _ _ _ . _ . . _ _ . . . __ _ _ . _ _ _ _
.
-9-
l
After verifying locally that 1E12*F004A was closed, the NCO attempted
to open 1E12*F006A valve as required to establish shut down cooling.
This attempt failed however because of the interlock that prevents
4 -the F006A valve from opening if the F004A is not in the closed
. (indicated) position. Control of the division II RHR system was then
transferred to the division II RSP and the system aligned to the shut -r
down cooling mode without incident. The NCO established a cooldown
rate of approximately 80 F per hour with the heat transfer path to
the' standby, service water system. The licensee has collected the
test data to evaluate for conformance to acceptance criteria.
,
'
% violation or deviations were observed in this area of inspection.
- '
,
.
,
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,o' '
.T,est 1-ST-25B Main Steam Isolation Valve (MSIV) Full Closure:
.
2
,
Th'e RI witnessed the pefformance of startup test 1-ST-25B, "MSIV Full
Closure", on-June 8,1986 with reactor power at 100 percent and rated
- ., . core f, low at 96 percent. The licensee initiated the test at
1733 hours0.0201 days <br />0.481 hours <br />0.00287 weeks <br />6.594065e-4 months <br />.by simulating a loss of condenser vacuum which results in
the MSIVs closing when in the run mode. Upon initiation of the MSIV
full closure, the reactor tripped and the SRVs opened momentarily to
control pressure. Following the initial opening of the SRVs, only
'
, 1 SRV was observed to reopen. The high pressure core spray (HPCS)
and reactor core isolation cooling (RCIC) systems initiated on
reactor vessel water level reaching the level 2 setpoint. The
reactor vessel water level recovered quickly because of the
subsequent swell of vessel water and the continuous feed from the
reactor feed water pumps. The NC0 secured the HPCS injection valve
prior to the system injecting based on the rising vessel water level. ,
Approximately 3 minutes into the transmit, the reactor feed pumps
tripped on vessel high level. The reactor feed water B pump was
subsequently restarted and vessel level maintained within the normal
band. Control room conduct during the test was observed to be well
coordinated and efficient. The licensee has collected the test data
to evaluate for conformance to acceptance criteria.
"
No violations or deviations were identified in this area of
inspection,
i
! o Test 1-ST-19 Core Performance
The SRI witnessed the running of traversing incore probe (TIP) traces
in preparation for computer calculation of reactor power / thermal
limits. The required data was extracted and inserted into startup
- test procedure 1-ST-19, " Core Performance" by the licensee for test
- condition (TC) 6 verification of core thermal limits and core thermal
i power. A preliminary review of the test data by the SRI revealed
i-
that core thermal limits were well within the TS limits and core
thermal power was approximately 99.3 percent of rated thermal power.
1
.
i
. _ _ _ _ , . _ . , , , , _ , m . _ , _ _ _ _ . _ . _ . . .,.-,,...,_,_...,,_,_,,,.,,m._,,,_,,_._,_, -
,m.__~_. .m-- - _-
l
- -
- . ,.
-10-
L
~
,
No violations or deviations were identified in this area of
. inspection.
,
6., Safety System Walkdown.
l'Duringthisinspebtionperiod,theSRIandRIperformedawalkdownofthe
'
"C",RHR system to verify proper system alignment for operability as
required by TS for Operational Conditions 1, 2, 3, 4, and 5. It was
observed that:
,
o sy' stem valves were properly aligned;
o abnormal control-room instrumentation readings or alarms were
-
present;
o no leakage from major components was present;
o the "C" RHR pump upper and lower bearing oil reservoirs were properly
filled; and
o accessible hangers and supports were intact.
No anomalies were noted that would have affected "C" RHR system
operability for low pressure coolant injection (LPCI). However, certain
discrepancies were noted when comparing the system condition / lineup with
the engineering piping and instrument diagrams (PIDs). It was noted that
pipe caps were missing from five vent and drain locations where PID-27-7C
indicated caps were installed; valve E12*MOVF064C which was open showed
closed on PID-27-7C; and valve E12*VF063C was not locked as shown on
PID-4-3C. Also, several valves were locked closed although the drawings
did not indicate locks installed. The failure to implement procedures to
maintain a safety system condition / lineup as shown on design output
drawings or to modify the design output drawings to reflect the required
system condition was identified by the SRI as an apparent violation
(458/8620-02). Subsequent discussions with licensee management revealed
that minimum flow valve E12*MOVF064C is open for standby operation and
condensate fill and flush valve E12*VF063C (not a major flow path valve)
had been identified as unlocked during an operations review of all valves
on the locked valve list. It had been noted that the valve was unlocked
and inaccessible without a scaffold or ladder. No action had been taken
to gain access to the valve to install a lock. The licensee took
immediate action to obtain ladders, and the valve was locked as shown on
PID-4-3C when the SRI identified that the valve was not locked. It was
also noted that the system operating procedure valve lineup did not
indicate that valve E12*VF063C was to be locked closed as show on the PID.
In response to the identified violation, the licensee should address how
they will assure that PID drawing, Stone and Webster flow diagrams, system
operating procedure valve lineups, locked valve lists and actual system
configuration are consistent for an identified system operational
condition and they should identify procedural controls which allow
deviation from drawing requirements. For example, procedures may allow
-
,
s a--
..
,
.
-11-
the locking of more valves than shown on the drawings at the discretion of
the operations staff. This issue should also be addressed for other
drawings / documents used routinely by operations and/or maintenance for
performance of work activities.
7. Operational Safety Verification
The SRI and RI observed operational activities throughout the inspection
period and closely monitored operational events. Control room activities
and conduct were observed to be well controlled and efficient. Proper
control room staffing was maintained and access to the control room
operational area was controlled. The licensee was adhering to limiting
conditions for operation (LCO) as they occurred. Operators were
questioned regarding lit annunciators and they understood why the
annunciators were lit in all cases. Selected shift turnover meetings were
observed, and all necessary information concerning plant status was
apparently being covered in these meetings. A walkdown of the "C" RHR
system was conducted, and the valves were observed to be in the proper
position for standby operation. Several plant tours were conducted and
overall plant cleanliness was good. During these plant tours, radiation
protection area postings were observed to be accurate.
During this inspection period the licensee completed a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />
verification run at full power operation, and all planned startup testing
was completed.
No violations or deviations were identified in this area of inspection.
8. Maintenance Witness
During this inspection period, the RI observed portions of selected
corrective maintenance activities to verify that maintenance activities
are being conducted in accordance with approved procedures, TS and
appropriate industrial codes and standards. The RI verified through
direct observation and review of records that:
o maintenance activities did not violate LCOs;
o redundant components were available;
o required administrative approvals and tagouts were obtained before
initiating work;
o procedures were adequate to control the work;
o radiological controls were properly implemented where applicable;
o QC hold points were established and observed; and
o replacement parts and materials used were properly certified.
, o
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The_following two corrective maintenance activities were observed:
o Control Rod Drive (CRO) Cooling Water Check Valve:
On May 7, 1986, the licensee experienced a failure of control
rod 24-33 to insert or withdraw during control rod manipulations.
Trouble shooting of the hydraulic control unit (HCU) revealed that
CRD cooling water check valve C11-V138 was not reseating when drive
water was applied to HCU 24-33. This condition allowed the drive
water to flow back thru the cooling water line and thus the necessary
lift was not being provided to the CRD drive piston to insert the rod
or retract the collet finger to allow withdraw of the control rod.
This condition would not have prevented the control rod from
inserting during a reactor scram.
~The. licensee initiated prompt maintenance work request (MWR) 39099 to
clean', inspect and replace if necessary the ball checks to CRD
cooling water check valve C11-V138 on HCU 24-33. The RI verified
prior to initiating work that the MWR had been properly initiated; QC
notification points were established, the job plan was appropriate to
control the work; and a job briefing had been performed as evidenced
by maintenance personnel signatures on the job briefing sheet. In
addition, the RI verified the requirements of T.S. 3.1.3.1 for an
inoperable control = rod were being complied with.
Prior to in'itiating work, the nuclear equipment operator (NE0)
obtained the shift supervisors permission to isolate HCU 24-33. The
, ' HCU was then isolated and tagged out using clearance number 86-192,
and.the ball was removed from the check valve. Inspection of the
ball revealed surface scratches and abrasions. Subsequent flushings
of the cooling water line produced a 1/2" x 1/16" round metal sliver.
'
A new ball was then installed in the check valve and the system
. verified operable using surveillance test procedure STP-052-0101,
" Control Rod Movement Operability Check." LC0 86-394, which was
initiated to track the action requirements of T.S. 3.1.3.1 was closed
based on the acceptable performance of STP-052-0101.
No violations or deviations were identified in this area of
inspection,
o Division II Emergency Diesel Generator:
During the performance of surveillance test STP-309-0202," Diesel
Generator Division II Operability Test," on June 5,1986, the " ready
to load" light did not illuminate after the diesel generator achieved
rated frequency and voltage. The licensee noted this condition and
immediately shut down the diesel. The failure of the " ready to load
light" to illuminate indicates that the standby generator breaker
1 ENS *ACB27 would not have closed onto standby buss IENS*SW618.
Prompt MWR 41549 was then initiated to trouble shoot and restore the
ready to load circuit to operable status. The RI reviewed the MWR
. *
.
-13-
prior to the initiation of work and verified that a deficiency tag
had been placed and the applicable LC0 initiated. The appropriate
hold points were placed in the procedure and a quality control
representative was present during the performance of this MWR. The
RI verified that the lifted leads were identified in the Lifted Lead
and Jumper Log as tag, numbers 86-3101-001 thru 006 and that the
, restored leads were independently verified.
The licensee identified-that relays EGS*UVRA-UVRB needed the pick up
voltages adjusted to within the setpoint limits. This work was
subsequently performed under MWR 41549, and the diesel restored to
operable statuu at 2041 hrs after the successful completion of
surveillance STP-309-0202. LC0 86-472 was then cancelled at
2045 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.781225e-4 months <br />.
No violations'or deviations were identified in this area of
inspection.
9. Surveillance Witness
The SRI and RI witnessed surveillance testing conducted by the licensee
during this inspection period and the following observations were made:
o Surveillance Test STP-051-4210:
The SRI witnessed a portion of the instrumentation
surveillance STP-051-4210, "RPS/RHR Reactor Vessel Steam Dome
Pressure - High, Monthly Chfunct, 18 Month Chcal, and 18 Month LSFT
(821-N078B,B21-6798)" conducted on May 28, 1986. The portion of the
test observed was known to cause a half scram signal and the
technicians were cautious in verifying the other division was not
tripped prior to test performance. They also limited the time that
the half scram was allowed to be in by close coordination with the
operations staff. During the performance of this test it was noted
that there were more than one copy of certain pages of the procedure
in the official work copy. This created some confusion during test
performance. The SRI discussed this with instrumentation maintenance
supervision and it was discovered that there were seven open TCNs
against this procedure and the preparer of the last TCN (No. 86-0581)
had failed to use copies of previous permanent TCN pages to markup
for the new TCN as required by administrative procedure. This
resulted in the official work copy of the procedure containing three
page eights and three page elevens with a different TCN number on
each page. This failure to follow administrative procedures for
issue of TCNs was identified by the SRI as an apparent violation
(458/8620-01).
The SRI also discussed the status of incorporation of TCNs in
procedure revisions in response to a previous NRC violation and it
was found that there are a total of 38 procedures with more than
three open TCNs out of a total procedure population of
. - - .. - .-- - .
,, .
-14-
approximately 3500. The SRI requested a date from licensee
management for revision of these 38 procedures to incorporate the
'open TCNs and management stated that these procedures would be issued
by September 1, 1986.
o Surveillance Test STP-309-0202:
The RI observed the performance of surveillance STP-309-0202, " Diesel
Generator Division II Operability Test," Revision 5 on June 5, 1986,
.with the plant in operational condition 1. This surveillance is
'
. designed to demonstrate the operability of the Division II diesel
,
generator and satisfie's T.S. Sections 4.8.1.1.2.a.1 through
i 4.8.1.-l.2.a.7;and 4.8.1.1.2.c.1 and c.2.
.
-
) Prior to' initiating the test, communications were established between
'
the NE0 at the remote diesel panel and the NC0 in the Main Control
i Room. LC0 86-472 had previously been initiated in accordance with
! T.S. 3.8.1.1.b because of the preplanned preventive maintenance (PM)
'
which had been performed on this diesel generator prior to beginning
'STP-309-0202. Upon initiation of the STP, the diesel generator was
observed to attain rated frequency and voltage within the required 10
seconds, however the " ready to load" light did not illuminate at
either the remote panel or in the main control room. The diesel
generator was shutdown within 30 seconds of its starting and prompt
MWR 41549 initiated. The failure of the " ready to load" light to
illuminate indicates that the standby generator breaker IENS*ACB27
would not have closed onto standby bus IENS*SW618. After verifying
the diesel generator had failed its surveillance test, the licensee
initiated the actions required by T.S. 3.8.1.1.b for an inoperable
diesel generator, because of any cause other than the performance of
preplanned preventive maintenance.
MWR 41549 was completed at 2015 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.667075e-4 months <br /> on June 5, 1986, and
STP-309-0202 was satisfactorily performed. LC0 86-472 was then
- closed based on the diesel generator satisfying the operability test.
No violations or deviations were identified in area of the inspection.
i 10. Licensee Plans for Coping With Strikes
,
The SRI reviewed licensee plans for coping with strikes during this
inspection period. It was found that the licensee had addressed such
issues as personnel / training requirements, security support requirements,
>
offsite support, etc. This area will be reviewed further during future
NRC inspections as required.
No violations or deviations were identified in this area of inspection.
1
!
!
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-15-
11. Status of TMI Action Item
(Closed) NUREG-0737, Item I.G.1: Trair;ing during initial startup test
. phase.
The licensee has completed the initial startup test program training
requirements as described in NUREG-0737, Item I.G.1 and FSAR
Section 14.2.3 Training During Initial Startup Test Phase. The startup
test training program was established to assure that personnel from each
of the six operating shift crews:
o observed a reactor scram;
o observed a pressure regulator transient;
o observed a water level setpoint transient;
o operated the operation of RCIC system; and
o observed a turbine trip or load rejection.
The startup test program was balanced, as much as practical, between the
six shif ts to assure that each shift was exposed to the above, off-normal
events. This has resulted in each shift having at least four individuals
who have observed actual plant responses for each of these events. Based
on the shift experience that now exists for each shift crew, no further
licensee action regarding NUREG-0737, Item I.G.1 is necessary.
This TMI action item is closed.
12. Exit and Inspection Interview
An exit interview was conducted on June 19, 1986, with licensee
representatives (identified in paragraph 1). During this interview, the
SRI reviewed the scope and findings of the inspection.