ML17285A215
| ML17285A215 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/11/1989 |
| From: | Bosted C, Johnson P, Sorensen R, Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17285A212 | List: |
| References | |
| 50-397-88-40, NUDOCS 8901270347 | |
| Download: ML17285A215 (19) | |
See also: IR 05000397/1988040
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION
V
Report No:
Docket No:
Licensee:
50-397'/88-40
50-397
Washington Public Power Supply System
P. 0.
Box 968
Richland,
WA 99352
Facility Name:'ashington
Nuclear Project
No.
2
(WNP-2)
Inspection at:-
WNP-2 Site near Richland,
Inspection
Conducted:
November
14, - December
14,
1988
Inspectors:
.
C.
ste
,
en>or
es>
ent
nspector
rensen,
es> ent
nspector
ate
igne
~(i r
ate
cygne
s((~
g
Approved by:
R.
P.
merman
Nov.
30 - Dec.
Chief,
eac
r Projects
Branch
nson
Nov.
- Dec.
Chief,
eact r Projects
Section
3
ate.S
gne
'i(zq
ate
>gne
(~i
P.
H.
o
Reacto
nson,
C ief
rojects Section
3
Date Signed
Summary:
Ins ection
on November
14 - December
14,
1988
50-397/88-40
A~I:
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rements to sample th'e control room essential charcoal adsorber filter material after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation (paragraph 3), and (2) delay in declaring the SN-7 degraded voltage protective circuit inoperable when a delinquent surveillance test was identified (paragraph 8). Strengths were observed: ( 1) in the performance of operations personnel during shutdown and restart evolutions, (2) in the. scheduling and management of work activities during the forced outage (December 1-7), and (3) in management's timely decision to shut down the plant in response to identified leakage of a wetwell vacuum breaker (paragraph 9). Weaknesses were observed in Technical Specification interpretations associated with the two violations discussed above. DETAILS Persons Contacted ¹L. D.
- C
- C
¹J.
- K.
C.
- R
'J.
- A.
- D
- R.
- S
- J
- 'g
- R
M. Oxsen, Assistant Managing.Director for Operations Bouchey, Director, Licensing and Assurance McGilton, Manager, Safety and Assurance Powers, Plant Manager Baker, Assistant Plant Manager Cowan, Nuclear Safety. Assurance Manager Edwards, equality Control Manager Graybeal, Health Physics and Chemistry Manager Harmon, Maintenance Nanager Hosier, Licensing Manager Kobus, guality Assurance Manager Koenigs, Technical Manager McKay, Operations Manager Peters, Administrative Manager Shaeffer, Assistant Operations Manager Mebring, Assistant Maintenance Manager Wuestefeld, Assistant Technical-Manager The inspectors also interviewed various control room operators; shift supervisors and shift managers; and maintenance, engineering, quality assurance, and management personnel. ¹Attended the Exit Meeting on December 2, 1988.
- Attended the Exit Meeting on December 9, 1988.
Plant Status . At the start of the inspection period, the plant was operating at 100% power. The plant operated at this level until November 18. On that date, during a surveillance on the main steam isolation valves, an instrument and controls technician error caused recirculation pump RRC-P-1B to trip from fast to slow speed. This caused reactor power to decrease to approximately 80Ã power. This mistake occurred when the technician placed a digital voltmeter on incorrect terminals, and selected the ohms scale instead of the volts scale. This caused a low impedance to be placed across the terminals, causing actuation of the recirculation pump trip (RPT) relay. Reactor power was restored shortly thereafter when the pump was shifted to fast speed. On November 21, an entry into Technical Specification 3.0.3 was required when three time delay relays for degraded voltage on vital bus SM-7 were found not to have been calibrated within the required Technical Specification time interval (see paragraph 8 for additional details). Reactor power was reduced to begin a reactor shutdown at about 3:00 p.m., and an Unusual Event was declared to comply with the Emergency Plan Implementing Procedures. h'hile the shutdown was in progress, technicians and engineers completed the required testing, and the shutdown was stopped at 89/ power. After the plant exited the 3.0.3 action statement, power was restored to 1005. Power levels remained at this level until, November 30. On November 30, vacuum breaker valve CSP-V-9 failed to pass. a local leak rate test, which required that the plant be shut down to satisfy the Technical Specifications. At approximately 11:50 p.m., a normal reactor shutdown was started and another Unusual Event was declared (see paragraph 9 for additional details). The plant entered Mode 4 at approximately 11:30 am on December I and the Unusual Event was terminated. An unscheduled maintenance outage commenced with work being performed on the vacuum breaker valve 'and other items needing attention. After replacing the "Viton" seat in CSP-V-9 twice, and another seat in a similar valve, CSP-V-5, the plant made preparations to startup on December 2. On December 2, during valve stroke testing of main steam isolation valves, the outboard valve (MS-V-28A) f'r the "A" steam line would not fully.stroke and stuck at approximately 4 inches open. The valve would close, but it could not be fully opened (normal stroke is 15 inches). A decision was made after some troubleshooting to disassemble the valve. Once the valve piston was removed, galling was observed on the interior of the valve cylinder. Following maintenance on the valve and replacement of the valve piston, the plant was started up on December 8 and power was increased. At the end of the reporting period, reactor power was approximately 80Ã. Previousl Identified NRC Ins ection Items 92701, 92702) The inspectors reviewed records, interviewed personnel, and inspected plant conditions relative to licensee actions on previously identified inspection findings: a. (Closed) Unresolved Item 397/88-37-04): Control Room Ventilation Charcoa Fi ter On November 7, the inspector noticed that the control room ventilation "8" train charcoal filter operation had exceeded 720 hours without a sample of the filter material. This was not in compliance with.Technical Specification 4.7.2.d, which requires that a sample of the charcoal in the emergency filtration units be taken after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation. The "B" train had reached 720 hours of operation on October 24. The charcoal was sampled on November 8 after inspector discussions with management. At the time of sampling, the "B" train had been run for approximately 943 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.588115e-4 months <br />. The hours of operation on charcoal filter unit "B" as of 12:00 a.m., November 7, 1988, were recorded as 933.9 in Plant Procedures Manual (PPM) 7.0.0, "Shift and Daily Instrument Checks". The licensee's failure to take a sample of the charcoal filter medium after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation was identified as*an apparent violation of Technical Specification 4.7..2.d. (Enforcement Item 88-40-01). 4. 0 erational Safet Verification 71707 a. Plant Tours The following plant areas were toured by the inspectors during the course of the inspection: Reactor Building Control Room Diesel Generator Building Radwaste Building Service Water Buildings Technical Support Center Turbine Generator Building Yard Area and Perimeter b. The following items were observed during the tours: (1) 0 eratin Lo s and Records. Records were reviewed against Technical Specification and administrative control procedure requirements. t1onitorin Instrumentation. Process instruments were observed or corre ation etween c annels and for conformance with Technical Specification requirements. Shift Ma~nein . Control room and shift manning were observed ~or con ormance with 10 CFR 50.54. (k), Technical Specifica- tions, and administrative procedures. The attentiveness of the operators was observed in the execution of their duties and the control room was free of distractions such as non-work related radios and reading materials. (4) E ui ment Lineu s. Valves and electrical breakers were veri ie to e in the position or condition required by Technical Specifications and Administrative procedures for the applicable plant mode. This verification included routine control board indication reviews and conduct of partial system lineups. Technical Specification limiting conditions for operation were verified by direct observation. E ui ment Ta in . Selected equipment, for which tagging requests had been initiated, was observed to verify that tags were in place and the equipment was in the condition specified. General Plant E ui ment Conditions. Plant equipment was observed for indications of system leakage, improper lubrica- tion, or other conditions that would prevent the system from fulfillingits functional requirements. Annunciators were observed to ascertain their status and operability. During a tour on October 11, the inspector noted that several panel screws were missing from motor control center YiC-4A. This was brought to the attention of maintenance management and this matter was corrected by the end of the reporting period. (7) Fire Protection. Fire fighting equipment and controls were h hi 1Sp if' d administrative procedures. See paragraph 10 for- additional comments. (8) Plant Chemistr . Chemical analyses and trend results were reviewe or conformance with Technical Specifications and administrative control procedures. (9) Radiation Protection Controls. The inspectors periodically o served radio ogica protection practices to determine whether the licensee's program was being implemented in conformance with facility policies and procedures and in compliance with regulatory requirements. The inspectors also observed compliance with Radiation Exposure Permits, proper wearing of protective equipment and personnel monitoring devices, and personnel frisking practices. Radiation monitoring equipment was,frequently monitored to verify operability and adherence to calibration frequency. ( 10) Plant Housekee in . Plant conditions and material/equipment storage were o served to determine the general state of cleanliness and housekeeping. (11} ~Securit . The inspectors periodically observed security practices to ascertain that the licensee's implementation of the security plans was in accordance with site procedures, that the search equipment at the access control points was opera- tional, that the vital area portals were kept locked and alarmed, and that personnel allowed access to the protected area were =badged and monitored and the monitoring equipment was functional. No- violations or deviations were identified. 5. En ineered Safet Feature S stem Malkdown 71707, 7l710) Selected engineered safety feature systems (and systems important to safety) were walked down by the inspectors to confirm that the systems were aligned in accordance with plant procedures. During the walkdown of the systems, items such as hangers, supports, electrical power supplies, cabinets, and cables were, inspected to determine that they were operable and in a condition to perform their required functions. The inspectors also verified that the system valves were in the req'uired position and locked as appropriate. The local and remote position indication and controls were also confirmed to be in the required position and operable. Accessible portions of the following systems were walked down on the indicated dates: ~Sstem Diesel Generator Systems, Divisions 1, 2, and 3. Dates November 28 Hydrogen Recombiners Low Pressure Coolant Injection, (LPCI) Tra.ins "A", "B", and "C" Low Pressure Core Spray (LPCS) High Pressure Core -Spray (HPCS) Reactor Core Isolation Cooling (RCIC) Residual Heat Removal (RHR), Trains "A" and "B" r Scram Discharge Volume System Standby Liquid Control (SLC) System Standby Service h'ater System 125V DC Electrical Distribution, Divisions 1 and 2 250V DC Electrical Distribution No violations deviations were identified. 6. Surveillance Testin 61726 November 28 November 21 November 21 November 21 November 29 December 6 November 28 November 29 November 16 November 16, 21, 28 ( November 16, 21, 28 Surveillance tests required to be performed by the Technical Specifications (TS) were reviewed on a sampling basis to verify that: 1) the surveillance tests were correctly included on the facility schedule; 2) a technically adequate procedure existed for performance of the surveillance tests; 3) the surveillance tests had been performed at the frequency specified in the TS; and 4) test results satisfied acceptance criteria or were properly dispositioned. Portions of the following surveillance tests were observed by the inspectors on the dates shown: Procedure 7.4.3.7.6.3 Source Range Monitor (SRM) Channel B Functional Test Dates Performed December 6 8.3.118 Leak Rate Testing of Main Steam valve MS-V-28A December 6 7.4.3.7. 10.3 Loose Parts Monitor Channel Check December 9 Yo violations or deviations were identified. During the inspection period, the inspectors observed and reviewed documentation associated with maintenance and problem investigation activities to verify compliance with regulatory requirements and with administrative and maintenance procedures, required gA/gC involvement, proper use of safety tags, proper equipment alignment and use of jumpers, personnel qualifications, and proper retesting. The inspectors also verified that reportabi lity for these activities was correct. The inspectors witnessed portions of the following maintenance activities: Descri tion Troubleshoot Main/Arcing Contacts of SM-1 Feeder Breaker per AT 7616 Cleaning of floor drain receiver level switches per AT 6526 Repair Containment Supply/Purge (CSP) Valve V-9 per AT 7834 Repair Reactor Pressure Vessel (RPV) temperature recorder per AV 1879 Shim rails on 606 'verhead crane per AT 7086 Inspection of turbine lube oil storage tank per AT 7691 No violations or deviations were identified. 8. De raded Volta e Rela s for Vital Bus SM-7 Dates Performed November 16 November 21 December 1 December 1 December 1 December 6 On November 18, 1988, while conducting a safety system functional inspection (SSFI) of the LPCS, members of the licensee's staff initiated a nonconformance report (NCR) which noted that three 3-second time delay relays which are part of the degraded voltage protection for vital buses SM-7 and SM-8 had not been included in the channel functional test (CFT) performed pursuant to Section 4.3.3. 1 of the Technical Specifications (the NCR incorrectly referenced the channel calibration (CC) procedure). These relays had been included in the Scheduled Maintenance System (SMS), scheduled at alternate refueling intervals. Inspection by NRC represen- tatives determined that the sequence of the licensee's subsequent evaluation of this issue was as shown below (all times PST). Additional information on the event was also provided in the licensee's LER 88-36, issued on December 19, 1988. 11/18 5:00 p.m. (approx.) Technical staff personnel reviewing the NCR concluded that exclusion of the relays from the CFT had been previously reviewed with NRR. However, the issue was identified as 11/19 11/21 8:30 a.m. (approx.) 8:45 a.m. ll:00 a.m. ll:30 a.m. - 2:00 p.m. 2:00 p.m. 2:59 p.m. 3:10 p.m. reportable pursuant to 10 CFR 50.73 (this was done, in part, to ensure appropriate followup review) and maintenance personnel scheduled to work during the week-end were requested to check the maintenance history for these relays to determine their status. Maintenance record revi'ew indicated that the SMS testing had been performed for vital bus SM-8 during the 1987 refueling outage. However, the SMS testing scheduled for SM-7 during the 1988 refueling outage had been deferred because of conflicting plant conditions, and had not been performed since initial plant startup. This was not considered at that time as a potentially reportable issue, since the monthly CFT was not believed to include or require testing of the 3-second relays. After further discussion of the issue, the licensee's technical staff questioned whether the 3.-second relays had been inappropriately excluded from the channel calibration (CC), required'y Section 4.3.3. 1 of the Technical Speci-'ications to be performed at 18-month intervals. The SMS testing for the .3-second relays in SM-8 had been performed within 18 months (+ 25%), but the relays in SM-7 had not. The NRC Senior Resident Inspector (SRI) questioned whether the plant's status was consistent with Section 3.0.3 of the Technical Specifications (as of approximately 8:30 a.m.) because of an overdue surveillance test. After further review and discussion with the technical staff, licensee management concluded that the CC surveil- lance on SM-7 was delinquent. A licensee plan was initiated to request from NRR a 24-hour waiver of compliance while the 3-second relays were calibrated. NRC SRI again questioned whether plant status was consistent with Section 3.0.3 of the Technical Specifications. Procedure review and other actions in progress in prepa- ration for on-line calibration of the 3-second relays. Discussions ongoing with Region V and NRR regarding possible relief from Technical Specifications requirements which would necessitate plant shutdown. After further query from the NRC SRI, Plant Management declared the SM-7 degraded voltage protection inoperable (due to delinquent surveillance). Technical Specification 3.0.3 was entered, and an Unusual Event was declared. Plant operators began reducing power in preparation for shutdown as required by Section 3.0.3. Channel calibration was begun for SM-7 3-second degraded voltage relays. 5:55 6:00 p.m. Channel calibration was completed satisfactorily. SM-7 degraded voltage circuit was declared operable; power reduction was halted at 89K. 7:25 p.mo The licensee exited Technical Specification 3.0.3,and terminated the Unusual Event. Region V representatives initially questioned whether plant staff personnel had inappropriately delayed acting on the NCR initiated on November 18. Interviews with cognizant members of the licensee's staff and examination of pertinent documents led to the following conclusions: The NCR prepared on November 18 questioned the correctness of the CFT scope. Plant Technical personnel who reviewed the NCR at that time believed that this was not a problem. Following further questions by Region V and NRR (on November 23), the licensee submitted an amendment request to clarify the Technical Specification requirement. This amendment, issued by NRR on January 6, 1989, more clearly indicated that the 3-second relays are not required to be included in the monthly CFT. The adequacy of the CC was not questioned until early on Monday morning, November 21. After review of available information and discussion with Plant Technical personnel, Plant Management concluded at about 11:00 a.m. that the CC for SM-7. was incomplete and therefore delinquent. The NRC representatives concluded that the time period between 8:30 and 11:00 a.m. provided reasonable allowance for management evaluation of the condition before a conclusion regarding acceptability of the surveillance test status was reached. The licensee's delay (between ll:00 a.m. and 2:00 p.m.) in declaring the SM-7 degraded voltage protection circuit inoperable and entering Technical Specification 3.0.3 was inappropriate. NRC guidance in Generic Letter 87-09 apparently contributed to this delay, although this guidance was stated in the generic letter to be a basis for requesting a change t'o the Technical Specifications, not for excusing deviation from them. Technical Specification 4.0.3 states that failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Conditi'on for Operation. The Plant Procedures Manual, Section 1.3. 1, Standing Orders/Night Orders (implemented pursuant to the requirements of Technical Specification 6.8.1 and Regulatory Guide 1.33), also states in Attachment I, Standin 0 eratin Orders ( Item 6), that if any surveillance test or other condition indicates that a system is not operable, the Shift Manager is to log the condition and begin the action required by the Technical Specifications (in this case, Section 3.0.3; i.e., place the plant in at least STARTUP within the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />) in a timely manner. i -The fact that the Shift Manager did not take these actions until 2:00 p.m.,was a violation of regulatory requirements (Enforcement Item 88-40-02). However, the licensee apparently did not violate Technical Specification 3.0.3, since the CC was completed (and the degraded voltage feature was made operable) at 5:55 p.m., within 7 hours of the ll:00'.m. determination by Plant Management that the'C surveillance was delinquent. The importance of correct understanding of Technical Specifications requirements and generic letters and of appropriate response to the requirements was discussed with licensee management during an exit interview conducted on December 2, 1988. One violation was identified, as discussed above. , ~ Containment Pur e Valves and Vacuum Breakers During surveillance testing of the containment supply purge (CSP) valves for the week of November 27, the licensee identified a problem with position indication on one of the reactor building (RB) to wetwell (WW) vacuum breakers, CSP-V-10. Following testing and while cycling the vacuum breaker to repair the problem, technicians noticed leakage from under the disc of the upstream butterfly valve, CSP-V-9, a containment isolation valve. A local leak rate test (LLRT) of V-9 was subsequently conducted on November 30, which was unsuccessful due to excessive leakage through V-9. The plant was then shut. down, as required by the Technical Specifications, and an Unusual Event was declared. The licensee's investigation showed the cause of the leakage to be failure of the resilient seat material. This seat material attaches to the valve disc, and was found tom as a result of cycling the valve. The RB to WW vacuum breakers and associated butterfly valves are oriented in, such a way that condensation .from the suppression pool can collect behind the butterfly valve, causing corrosion at the bottom of the seating area. This corrosion may have been a contributing factor to tearing of the resilient seat when the valve was cycled. The resilient valve seats are made of a substance called Viton, which replaced the original Buna-N type material. The tom seat for V-9 was replaced; however, the replacement was subsequently noticed to have tears in it during the torquing of the seat retention ring. The second seat was also replaced and, as a precautionary measure, the butterfly valves associated with the other two RB to WW vacuum breakers were also visually inspected. One valve, CSP-V-6, was found to be acceptable, but another, CSP-V-5, was found with a tom seat. The seat for CSP-V-5 was replaced and passed a subsequent LLRT. Drywell containment purge supply and exhaust butterfly valves which have system configurations similar to the 'vacuum breaker butterfly valves were also leak tested satisfactorily. The licensee also leak tested WW purge exhaust butterfly valves CEP-3A and CEP-4A with satisfactory results. These valves are similar in configuration to the vacuum breaker butterfly valves and are also located in a horizontal pipe run. 10 During followup discussions with the inspector, the ~ licensee committed to perform leak testing of RB to WW vacuum breakers CSP-V-5, 6, and 9 and WW purge valves CEP-V-3A and 4A to ensure their integrity following each operation. This was a result of their particular location associated with the WW and their orientation in horizontal pipe- runs which could make them susceptible to corrosion and potential damage of the resilient seats whenever they"are operated. As of the end of the report period, the licensee was reviewing the adequacy of the Viton seats for this application. The inspector will follow the licensee's continuing actions under the routine inspection program. No violations or deviations were identified. Licensee Event Re ort (LER Followu 90712, 92700) LER 88-24, "Special Report - Reactor Containment Greater than 150 Degrees for f1ore Than Eight Hours", was reviewed by the inspector and was found to have several areas for which more information was needed. Accordingly, a letter was sent to the licensee requesting additional information on environmental qualification of the safety related equipment. This item will remain open pending receipt of the supplemental report. No violations of NRC requirements or deviations were identified. 'I Review of Periodic and S ecial Re orts (90713) Periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9.1 and 6.9.2 were reviewed by the inspector. This review included the following considerations: the report contained . the information required to be reported by NRC requirements; test results and/or supporting information were consistent with design predictions and performance specifications; and the reported information appeared valid. Within this scope, the following reports were reviewed by the inspectors. o Monthly'perating Report for October, 1988. No violations or deviations were identified. Exit Meetin 30703) Region V representatives met with licensee management representatives (denoted in paragraph 1) on December 2, 1988, to discuss their findings related to the November 21, 1988 event involving degraded voltage protection. The importance of proper and timely management response to Technical Specification requirements was discussed (paragraph 8). The resident inspectors met with licensee management representatives periodically during the report period to discuss inspection status, and ar, exit meeting was conducted with the indicated personnel (refer to paragraph 1) on December 9, 1988. The scope of the inspection and the inspector's findings, as noted in this report, were discussed and acknowledged by the licensee representatives. Prior to the December 8 plant restart, licensee management committed to perform a local leak rate test after each operation of vacuum breakers CSP-V-5, 6, and 9 and suppression pool purge valves CEP-V-3A and 4A (paragraph 9). The licensee did not identify as proprietary any'f the information reviewed by or discussed with the inspector during the inspection.