IR 05000272/1991023
| ML18096A300 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 09/24/1991 |
| From: | Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18096A297 | List: |
| References | |
| 50-272-91-23, 50-311-91-23, 50-354-91-16, NUDOCS 9110080140 | |
| Download: ML18096A300 (99) | |
Text
- Report No License No Licensee:.
Facilities:
Dates:
Inspectors:
Approved:
U.S. NUCLEAR REGULATORY COMMISSION
. REGION I 50.,;272/91-23 50-311/91-23 50-354/91-16 DPR-70 DPR-75
. NPF-57 Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Nuclear Generating Station Hope Creek Nuclear Generating Station July 31, 1991 - September 10, 1991 T. P. Johnson, Senior Resident Inspector S. M~ Pindale; Resident Inspector S. T. Barr, Resident Inspector H. K. Lathrop, Resident Inspector A. E.,,J...ppez-dbtrg_;;React. *~gineer
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Inspection Summary(
Inspection 50-272/91-23; 50-311/91-23; 50-354/91-16 on July 31, 1991 - September 10, 1991 Areas Inspected: Resident safety inspection of the following areas: operations, radiological controls, maintenance and surveillance testing, emergency preparedness, security, engineering technical support, safety assessment/quality verification, and licensee event report Results: An executive summary follows.
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EXECUTivE SUMMARY
Salem Inspection Reports 50-272/91-23; 50-311/91-23 Hope Creek Inspection Report 50-354/91-16 July 31, 1991 - September 10, 1991 OPERATIONS (Modules 71707, 92701, 93702)
Salem: The Salem units were operated in a safe manner. Radiation monitoring system actuations were reported, and licensee actions were appropriate. A Unit 1 safeguards equipment cabinet (SEC) failure and associated ESF actuations were appropriately reswnded to by the licensee. The licensee has plans to replace the SECs, as 28 SEC failures have occurred in the past four years. Following discussion with the NRC, the reporting requirements for the capture of any endangered or threatened sea turtles was satisfactorily
- modifie IIope Creek: The Hope Creek unit was operated in a safe manner. Good operator response was observed during a feedwater pump control failure, even though the individuals were only recently qualified in their positions.
RADIOWGICAL CONTROLS (Modules 71707, 93702)
Salem: Periodic inspector observation of station workers and Radiation Protection personnel implementation of radiological controls and protection program requirements did not identify any deficiencies. The material condition of the post accident sampling system was good.
. Chemistry technicians were observed as being proficient and effective during sampling and analysis efforts.*
- Hope Creek: Periodic inspector observation of station workers and Radiation Protection personnel implementation of radiological controls and protection program requirements.did not identify any deficiencies. Chemistry, training, and emergency preparedness personnel failed to adequately follow procedures associated with post accident sampling system (PASS)
operations. Consequently, deficient conditions involving the operability of the PASS were not documented nor corrected in a timely manner.
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J MAINTENANCE/SURVEILLANCE (Modules 61726, 62703)
Salem: Routine observations did not identify any deficiencies. A Unit 2 reactor Moderator Temperature Coefficient test was well planned and effectively conducted. A steam generator low pressure protection channel was identified by the licensee io be inoperable for a 26 day.
period due to. personnel error (inadequate self-verification. during testing). The licensee remains to determine if the oondition was unanalyi:ed. An engineered safeguards feature actuation occurred during testing of vital bus undervoltage relays.
. Hope Creek: Routine observations did not identify any deficiencies. A High Pressure Coolant Injection (HPCI) system actuation occurred during surveillance testing. There was no injection to the reactor vessel. The licensee has not yet determined the root cause of the initiation. After extensive investigation, the licensee was unable to determine a definite root cause of the "D" emergency diesel generator test failure in Mayi 1991, but has enhanced surveillance procedures in an effort to prevent recurrenc EMERGENCY PREPAREDNFBS (Modules 71707, 93702)
Hurricane preparations by the licensee were proactive and conservative. A Hope Creek.
emergency drill with full onsite participation accountability appeared to fulfill the drill objectiv~s and.provided a meaningful training opportunit *
SECURITY (Modules 71707, 93702)
Routine observation of protected area access and egress showed good control by the license ENGINEERING/TECHNICAL SUPPORT (Module 71707)
Salem: Review of the management of engineering work activities determined that they were performed in accordance with applicable procedures and properly prioritized and execute The licensee used prudent engineering practices and a conservative safety approach in the replacement of a reactor coolant system temperature detector and the restoration of the 13 loop cold leg temperature channel. An SEC failure resulted in the initiation of a Unit 1
- Technical Specification required shutdown.
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Hope Creek: Review of the management of engineering work activities determined that they were performed in accordance with applicable procedures and properly prioritized and executed. The licensee continued the_investigation into issues surrounding Filtration, Recirculation and Ventilation System (FRVS) operability. Some degradation of
- environmentally qualified (EQ) components*had occurred and the licensee concluded that the vent fan heaters would not have been able to perform as designed for the.full period of performance. A 1 OCFR2 l report was submitted by the heater control panel vendor to :the NRC. A reactor building ventilation backdraft isolation damper investigation in September 1990 noted a number of EQ, document and spare parts issues. Licensee actions to promptly address these issues were appropriat *
. SAFETY ASSF.SSMENT/QUALITY VERIFICATION (Modules 40500, 71707, 90712, 90713, 92700}
Salem: Significant Event Response Team (SERT) reports documenting two events which occurred during the last report period were reviewed by the resident staff. Although a weakness was identified in one of the reports, the inspectors concluded that the SERT process had been effectively utilized by the licenSee for the assessment of the two event Hope Creek: A comprehensive and independent scram review was thorough and effective in identifying common causal factors.
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- SUMMARY OF OPERATIONS Salem Units 1and2 Both Salem Units remained at power throughout the report period. As of September 10, 1991, Unit 1 had been on-line for 78 continuous days and Unit 2 for 11.2 Hope Creek.
The unit maintained operations thr011ghout the reporting period, with weekly power reductions to support main turbine control valve surveillance testing. * OPERA TIO NS Inspection Activities
. The inspectors verified that the facilities were operated safely and in conformance with regulatory requirements. Public Service Electric and Gas (PSE&G) Company management control was evaluated by direct observation* of activities, tours of the facilities, interviews and discussions with personnel, independent verification of safety system status and Technical Specification compliance, and review of facility records. These inspection activities were conducted in accordance with NRC inspection procedures 71707 and 93702. The inspectors performed normal and back-shift inspections, including deep back-shift (9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />) inspections as follows:
Unit Inspection Hours Dates Salem 5:00 p.m. - 6:00 /2/91 Salem 5:30 a.*m. - 7:30 /19/91 Salem 8:00 a.m. - 12:00 noon.
9/7/91 Hope Creek 5:30 a.m. - 7:30 /19/91
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2 Inspection Findings and Significant Plant Events 2. Salem Unit 1 Engineered Safeguards Feature (FSF) Actuation At 7:08 p.m. on August 15, 1991, an ESF actuation occu-rred when the lA safeguard equipment cabinet (SEC) spuriously actuated. The SEC starts and stops equipment due to accident and/or loss of power signals from the solid state protection system (SSPS). The -
partial actuation started the No. l 1 safety-injection (SI) pump, the No. 11 residual* heat removal (RHR) pump and the No. 11 auxiliary feedwater (AFW) pump. Selected containment fans tripped as designe Operators were dispatched to the SEC and noted a "MODE OP" light indicating the SEC had actuated. An operator observed that a system failure indicator was also lit. The licensee declared the SEC inoperable, entered Technical Specification (TS) 3.3.2.1 and commenced a Unit 1 shutdown from 100% power. The SEC was reset, and equipment was resfored to normal. Six minutes later at 7:14 p.m., a second SEC spurious actuation occurred affecting
. the same equipment. In addition, the IA emergency diesel generator (EDG) started but did..
not load since the vital bus remained powered from offsite. Safety equipment was again returned to normal.
Licensee troubleshooting determined that two circuit boards failed. The faulty chassis that contained these circuit boards was replaced with a spare chassis, and the SEC was tested satisfactorily. The licensee declared the lA SEC operable, and the unit shutdown was terminated at 30% at 12:14 a.m. on August 16, 199 The inspector reviewed the incident report, the troubleshooting surveillance test Sl.MD-FT.SEC-OOOl(Q), control room logs, previous SEC failures, and LER 91-27. The inspector also discussed the event with licensed operators, system engineers and plant management personnel. The inspector determined that licensee response. to the event was conservative and -
appropriate. Subsequent to this SEC failure, another failure occurred on September 5, 1991 (discussed in Section 7. l.B of this report). At the end of this report period, 28 SEC failures had b_een identified (18 on Unit 1 and 10 on Unit 2) since July 1987. (Previous recent failures were discussed in NRC Inspections 272/91-09, 90-24, 90-22, 90-13, 90-11, 90-04).
The inspector noted that the licensee intends to replace the SEC with an upgraded system during the next refueling outage for each unit. The inspector verified that Design Change Package 25C-2267 was scheduled for the upcoming Unit 2 outage beginning in January 199 Unit 1 Engineered Safeguards Feature (ESF) Actuation - Valve Failure On August 23j 199~, the Unit I steam generator (SG) blowdown valve No. 12GB4 failed closed due to a ruptured diaphragm on its actuator. The valve failed following a stroke tes The licensee characterized this inadvertent valve closure as an ESF actuation (since the valve
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provides for containment isolation) and properly reported this event to the NRC in accordance with 10CFR50. 72. The inspector verified that 12GB4 was subsequently isolated, repaired and satisfactorily reteste C. *
Radiation Monitor Engineered Safety Feature (ESF) Actuations The following ESF actuations occurred and were reported by the licensee during the period:
Unit Radiation Monitor Date Time
2R41C Aug.28, 1991 6:03 R41C Sep. 1,1991 1:54 Rl2A Sep.10,199 :15 R12B Sep.10, 1991 9:13 'These events continue to be indicative of the degraded radiation monitor system. Systems responded as designed causing a containment ventilation isolation or a control room ventilation start. As stated in previous LERs and management meetings, licensee actioris include short term and long term equipment upgrades. The inspector reviewed licensee actions regarding these events. The licensee intends to submit an LER for these events. No unacceptable conditions were noted. Change in Reporting Requirements For Capture of Endangered Species at Salem Over the course of the 1991 summer, a large increase in the number of captured endangered or threatened sea turtles occurred at the Salem Nuclear Generating Station. The mechanism for the capture of sea turtles is their impingement upon the circulating water system intake screens. The two types of turtles which have been taken this summer are the loggerhead turtle, a threatened species, and the Kemp's ridley turtle, an endangered species. As of the end of the inspection period, 23 loggerhead turtles had been captured, all but one alive; and one live Kemp's ridley had been captured. In accordance with a National Marine Fisheries Service approved procedure, the liCensee holds the captured turtles for a short time to determine their state of health. Subsequently, the turtles are tagged and released at a remote part of the Delaware Ba Prior to this year, an average of approximately 3.5 sea turtles per summer had been captured at Artificial Island, with a previous high of ten in 1988. *The PSE&G environmental engineering staff has attributed the large increase of captured turtles this year to the*
especially hot and dry weather, which caused the salt line in the Delaware River to migrate north and produced an abundant food supply for the turtles, thus drawing a larger number of sea turtles to the Artificial Island vicinity.
. For each sea turtle taken at Artificial Island, PSE&G is required to notify and provide data
. on the individual turtle to the National Marine Fisheries Service (NMFS). An informal consultation in accordance with Section 7(a) of the Endangered Species Act was conducted between PSE&G, NRC, NMFS and the Environmental Protection Agency in 1981 to study thejmpingement of sea turtles at Artifid.al Island. This informal consultation concluded that operation of the nuclear p0wer plants on Artificial Island would not jeopardize continued existence of these sea turtles and established the requirement for PSE&G environmental licensing to make a report to NMFS for each sea* turtle taken at either Salem or Hope Cree As a result of this NMFS reporting requirement, Salem Station had been reporting each turtle capture to the NRC as a four hour report in accordance with l0CFR50.72(b)(2)(vi), which requires a licensee to report any event "related to the... protection of the environment, for which*... notification to other government agencies has been or will be. made."
Due to the burden placed on the Salem operating crews by the reporting of an unusually high number of turtle captures, PSE&G Licensing.discussed the 10CFR50. 72 reporting requirement with the NRC. Following discussions between PSE&G, the NRC resident staff, Region I and NRR, it was determined that the individual captures of endangered or threatened sea turtles did not have to be reported in accordance with 10CFR50.72(b)(2)(vi).
In the view of the NRC, the intent of this paragraph is to report to the NRC conditions that*
are directly harmful to the environment (such as inadvertent radiological or chemical releases) for which a press release or off-site notification to other government agencies has been or will be made. Consequently, in August, PSE&G initiated a change to their reporting procedures and ceased. reporting turtle captures to the NRC Operations Center. The licensee is still required by a Technical Specification, Appendix B, requirement to inform. the NRC resident within 24 hour~ of a sea tu.rtle captur. Hope Creek Feedwater Control Failure On August 3, 1991, with the unit at 100% power, the "C" reactor feed pump (RFP)
suddenly increased speed to the high speed stops, causing reactor water level to increase rapidly to the high level alarm setpoint ( +40"). At the time, the "A" and "C" RFPs were in automatic control and the "B" RFP was tagged out for maintenance. The nuclear controls operator and the nuclear shift supervisor promptly took manual control of "A" and "~" feed pumps, terminating the level increase at +44". Reactor water level was returned to and maintained at its normal level ( +35") by manual control of feed pump speed. A failed dynamic compensator card was found in the "C" RFP control logic. The licensee replaced the card with one from the "B" RFP circuitry, tested operability, and returned feed pump control to automatic within four hours of the transien.
The inspector reviewed this event in detail (including chart recorder traces and annunciator logs) with operations personnel. The inspector concluded that operators had acted promptly and effectively in terminating the transient and manually controlling reactor water level until
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feed pump control could be returned to autc>matic. Good support was also noted from Instrument and Control (I&C) technicians and the technical staff system engineer. The
'inspector noted, that these individuals were only recently qualified in their position.
RADIOWGICAL CONTROIS Inspection Activities PSE&G' s conformance with the radiological protection program was verified on a periooic basis. These inspection activities were conducted in accordance with NRC inspection
procedures 71707 and 9370 * Inspection Findings 3~ Salem Post Accident Sampling System (PASS)
The inspector reviewed the Salem PASS, including the administrative controls for system operability. A Technical Specification (TS) interpretation (TSI number ADM-6.8.4.E), datf(
May 8, 1990, requires the PASS to be operable in Modes 1, 2 and 3. A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> TS action*
statement (TSAS) applies if PASS becomes inoperable. This also requires the licensee to inform the NRC and initiate action to restore the syste The inspector confirmed the licensee's use. of the TSI. The inspector reviewed work order (WO) 910617069 on a leaking PASS valve which required removing the system from servic for less than one day. From discussions with system engineers, operators, chemistry personnel, and* from reviewing other WOs, the inspector further confirmed that the licensee considers correcting PASS deficiencies a high priorit On August 20, 1991, the inspector* performed a walkdown of the Salem PASS in the auxiliary building. This walkdown* was accompanied by training department personnel. inspector noted that the PASS material condition was good, and that training personnel we competent and knowledgeable of PASS operatio Chemistry Observations
. During surveillance test observations (Section 4.3.1.A), the inspector observed two in-pla chemistry technicians perform reactor coolant sampling and analysis. ProCedures SC.CH CA.ZZ-0325(Q), "Boron By Titration", and SC.CH-SA.ZZ~0222(Q), "Sampling Reactor Coolant System and Residual Heat Removal Outlet", were observed. The inspector concluded.that chemistry technicians were proficient in their duties, and that the procedu were correctly implemente *
During this review, the inspector noted three material deficiencies associated with the Unit 2 primary sampling cabinet: a handle on a sample valve was broken (missing), a toggle switch was rt1alfunctioning, and a valve position light was not working. None of these deficiencies prevented sampling. However, there were no equipment deficiency tags identifying these problems. The inspector questioned chemistry management personnel regarding these item The licensee* reviewed' records and determined that these deficiencies were previously identified and were scheduled for work. The licensee stated that while the tags were not
_required by procedure, the tags did provide information about system status and would be poste. Hope Creek Post Accident Sampling System (PASS)
During the week of July 29, 1991, the inspector noted that the liquid sampling portion of the Hope Creek PASS operation was out of service due to flow blockage in the water return line to the torus (through solenoid valves SV643A and B). Consequently, the ability to take
residual heat removal and reactor coolant post accident samples was prevented. The inspector reviewed a work order that was initiated to effect repair (WO 910712101) and noted that this operability problem was identified to the Chemistry Department by the Training Department on July 12, 1.991, via a written feedback for On July 19, 1991, the licensee initiated an incident report (Report No.91-111) which identified that pipe sealant material had apparently been introduCed into the PASS. return line while performing containment local leak rate testing during the last refueling outage (January
- February 1991). The report noted the post-outage testing on PASS in early March 1991 indicated that the PASS was functioning properly at that time. Subsequently, the licensee completed repairs in accordance with WO 910712101; and the PASS was successfully tested and returned to service on August 5, 199 From review of related documents and interviews with chemistry, training, and operations personnel, the inspector learned the following 'relative to previous PASS operability problems. Attachment 4 provides a summary of the sequence ofevent During an emergency drill exercise on March 15, 1991, the drill observer and two chemistry technicians operating the PASS identified that there was insufficient flow to collect a representative sample. Emergency Preparedness personnel documented these findings but did not report the deficiency to station management and chemistry supervision for resolution until May 17, 1991. Upon receipt of the notification, the Chemistry Department personnel reviewed the reported deficiency and tested the system. At that time the Chemistry Department noted that the PASS appeared to be functioning properly and took no further action.
Concurrently, from the period between M~y 16 and July 9, 1991, the Training Department conducted trailling of chemistry technicians on the PASS. Several times during this period, some ~ning instructors and chemistry technicians identified intermittent flow problems, such that representative samples could not be reliably obtained from use of the PAS Reportedly, these problems were identified to the Chemistry Department several times but never documented until July 12, 199 Technical Specification (TS) 6.8.1 requires that proCedures be established, implemented and maintained covering the applicable procedures recommended 1n Appendix A of Regulatory Guide 1.33, and the procedures required to implement the requirements of NUREG-073 Further, TS 6.8.4.C requires the establishment, implementation, and maintenance of a program to include procedures for post-aecident sampling and analysis, including provision for the maintenance of sampling and analysis equipmen Accordingly, the licensee established Procedure HC.CH-EO.SH-OOOl(Q), "Post Accident Sample Panel Operation.". Section 2.9 of that procedure requires the PASS sample team to immediately inform chemistry supervision when any problems encountered during sampling, in order to effect resolution. The inspector noted that the sample* team's failure to inform chemistry supervision, until May 17, 1991, of the inability to obtain a representative sample due to flow problems on March 15, 1.991, constituted an example of a violation of TS 6. Procedure NC.NA-AP.ZZ-0009(Q), "Work Control Process,... Revision 2, was also established in accordance with TS 6.8.1. Sections 3.1 and 5.2, requires personnel to initiate work requests and recommend the hanging of Equipment Malfunction Information System (EMIS) tags for malfunctioning components or systems. The inspector noted that failure of licensee personnel to document and initiate work requests for the frequent and intermittent PASS flow problems that prevented representative sample acquisition, and to recommend the posting EMIS tags on the equipment, for the period between May 16 and July~' 1991, constituted a second example of violation of TS 6. 8.1. (50-354/91-16-01)
The inspector noted that the licensee's regard for the importance of maintaining the PASS operable was inconsistent relative to the attention afforded the Salem PASS (See Section 3.2.1.A). For example, the licensee had not established a Technical Specification Interpretation for TS 6.8.4. relative to the expected operability requirements for the Hope Creek PASS. As a result, a lower consideration has been applied to the maintenance and operability of the Hope Creek PASS. Consequently, even after WO 910712101 was initiated on July 12, the system remained out of service until August 5, 1991, since repair wa considered as a low priorit As a result of an independent assessment of this matter by the plant's Quality Assurance Department, the licensee has initiated action to direct more management attention oversight to the operability and maintenance of the Hope Creek PASS, including the identification of root causes and more immediate corrective actions for identified deficiencie *
- MAINTENANCE/SURVEILLANCE TESTING Maintenance Inspection Activity The inspectors observed selected maintenance activities on safety-related equipment to ascertain that these activities were conducted in accordance with approved procedures, Technical Specifications, and appropriate industrial codes and standards. These inspections were eonducted in accordance with NRC inspection procedure 6270 Portions of the following activities were observed by the inspector:
Work Request (WR)/Order Unit (WO) or Procedure Description Salem 1 Troubleshooting Plan lA Safeguards Equipment Cabinet (SEC)
Salem 1 Troubleshooting Plan Loop.13 cold leg temperature instrument Salem 1 W0910617069 Post Accident Sampling System (PASS)
Salem 1 W0910905104 Reactor Trip Breaker "A" Replacement Hope.Creek W0910712101 PASS Hope Creek W0910819083 PASS Hope Creek Various
"B" Reactor Feed Pump shaft seizure investigation and repair Hope Creek Various Rosemount transmitter replacements The maintenance activities inspected were effective with respect to meeting the safety objectives of the maintenance progra.2 Surveillance Testing Inspection Activity The inspectors performed detailed technical procedure reviews, witnessed in-progress surveillance testing, and reviewed completed surveillance packages. The inspectors verified that the surveillance tests were performed in accordance with Technical Specifications, approved procedures, and NRC regulations. These inspection activities were conducted in accordance with NRC inspection procedure 61726.
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The following surveillance tests were reviewed, with portions witnessed by the inspector:
Salem Salem 1 Salem 2 Hope Creek Procedure N S 1.MD-FT.SEC-0003(Q)
UC 18.1.013 Reactor Engineering Manual - Part 9 OP-ST.GK-001 lC SEC Reactor Trip Breaker Operability Moderator Temperature Coefficient
"B" Control Room Emergency Filter Monthly Surveillance The surveillance testing activities inspected were effective with respect to meeting the safety objectives of the surveillance testing progra * Inspection Findings 4. Salem Steam Generator Pressure Protection Channel Inoperable Due to Personnel Error On August 7, 1991, a technician identified that the two lead-lag controller switches were in the test position for the Unit 1 No. 13 steam generator (SG) pressure channel III, rendering that channel inoperable. This condition was identified during the performance of the SG pressure channel III surveillance test. The licensee determined that the two test switches were inadvertently left in that position during the previous channel functional test on July 12, 1991. Upon discovery, the lead-lag controller switches were returned to normal position, thereby restoring the channel to an operable status. Since the Technical Specification (TS)
operability requirements were not satisfied (inoperable channel to be placed in tripped condition) from July 12 - August 7, 1991, the licensee reported this event to the NRC in accordance with 10CFR50. 73 reporting requirements (30-day licensee event report - LER No. 91-26).
The licensee determined that the root cause of this event was personnel error due to inadequate self-verification of the technician performing the surveillance test on July 12; 1991. The event will be reviewed with applicable personnel. the licensee stated that in addition, the functional surveillance test procedure will be revised to require an independent verification of the lead-fag controller test switch position restoration. The inspector noted that an independent verification should have already been part of the test procedure, as the test is performed on a safety related system. The licensee stated that independent verification is currently required for such procedure steps, however, the test procedure for this event was
- developed prior to the current requirement and had not yet been revised. The licensee
initiated action to review similar procedures to assure that independent verification checks are accomplishe.
There is orie SG main steamline pressure monitor for each of the four SGs, which provide input to several safety-related circuits, including safety injection (SI). The four main.
. steamline pressure signals are divided into two protection sets; Protection Channel ill (No and 13 pressure channels), and Protection Channel IV (Nos. 11 and 14 pressure channels). The affected SI signal is high steam flow coincident with either low-low average reactor coolant temperature or low steamline pressure. Any two of the four low steamline pressure. signals will satisfy the low steamline pressure trip coincidence. The purpose of the lead-lag controller is to amplify the incoming steamline pressure signal such that the SI is initiated before the actual steamline pressure reaches the trip setpoint value. Accordingly, the lead~lag controller is credited in the accident analysi The licensee's analysis of the above condition identified that a potentially unanalyzed condition existed. A failure of Protection Channel IV (single failure) would result in the delay of pressure channel Nos. 11 and 14 to provide the safeguard actuation signals
- necessary if called upon during a small steamline break. Under this condition, coincident with the inoperable pressure channel No. 13, the required SI would be delayed due to the mispositioned lead-lag controller switches. The licensee stated that only the small steamline break accident was of concern, since the lead-lag function is not pertinent for larger
- steamline breaks involving large and immediate steamline pressure drops. This concern was conservatively reported to the NRC upon discovery on September 6, 1991 in accordance with 10CFR50. 72 reporting requirements (unanalyzed condition). The licensee is currently reviewing this issue to determine whether the condition is bounded by existing accident analyse The inspector reviewed LER No. 91-26 and found it to be acceptable. However, the report references a continuing review of this matter to determine its safety significance, but does n(
indicate that a supplemental LER will be provided.* The inspector discussed this concern with the licensee, who stated that a supplemental LER will be submitted upon completion of their review. The inspector had no further questions at this tim *
. Unit 2 Moderator Temperature Coefficient (MTC) Measurement On August 21, 1991, the inspector observed implementation of a surveillance test "MTC Measurement" on Unit 2. The test was required per Technical Specification (TS) 4.1.1. to ensure the value of MTC meets TS requirements when 300 ppm critical Boron concentration is achieve *
The tesi *involved maintenance, operations, reactor engineering and chemistry personnel, ar was performed in accordance \\\\'.ith the procedure, Reactor Engineering Manual (REM)-Part 9. The inspector observed test activities from the control room and in the chemistry lab (s
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section 3.2.1.B). The inspector concluded that the test was weli planned and conducte Personnel performance was commendabl During test procedure review, the.inspector noted that procedure REM-Part 9 does not follow
- the surveillance procedure format as required by Administrative Procedures NC.NA-AP.ZZ-0032 and AP-12. In particular, acceptance criteria were not included in the procedure step However, it was included in an attachment to the procedure. The inspector discussed this with licensee personnel. Their response was that the REM procedures were being currently revised by both the reactor e!lgineering section and by the Procedure Upgrade Project to *
meet administrative procedure format requirements. The inspector reviewed the licen*see's schedule to upgrade REM procedures and concluded it to be acceptabl * Unit 2 Engineered Safeguard Feature (ESF) Actuation During Surveillance Testing On August 26, 1991, an ESF actuation occurred while operating at 100% power when the 2A safeguards equipment cabinet (SEC) was inadvertently actuated during surveillance testing. While performing test procedure No. S2.MD-FT.4KV-000l(Q), "ESFAS Instrumentation Monthly Functional Test - 2A 4kV Vital B11s Under Voltage" a technician applied an electrical jumper across contacts in the wrong relay. This action actuated the 2A SEC, which automatically completed an electrical load shed on the 2A 4kV vital bus, started the No. 2A emergency diesel generator (EDG), and sequentially started associated safety rel.ated components. All systems functioned as designed. Control room operators entered procedure No. AOP-ELEC-4kV-A and verified the automatic actions. The 2A vital bus was subsequently restored to a normal lineup. The 2A EDG was subsequently secured and returned to a standby status. Operation of the unit was unaffec_ted by this even The inspector reviewed this event and determined that the cause was similar to. a previous ESF actuation that occurred on June 6, 1991, discussed in NRC Inspection Reports 272 &
311/91-15 and 272 & 311/91-19. The cause of the June 6, 1991 event was determined to be personnel error caused by human engineering deficiencies. Specifically, technicians are required to install an electrical jumper across contacts on the underside of relays, which are located on the inside of 4kV vital bus cubicle doors, an.ct are positioned approximately nine inches from the floor. Adjacent relays are located about 1/2 inches apart. On June 6, 1991, the technician accidentally touched an adjacent relay while approaching the relay to be jumpered. On August 26, 1991, the technician properly located and identified the proper relay while standing up (the label is above the relay). However, after he positioned himself on the floor to install the jumper, the technician inadvertently connected the jumper to the adjacent rela,
As a result of this latest event, Plant Operations requested that further undervoltage relay
- testing be suspended until an appropriate hardware change is implemented to prevent further occurrences. There are three 4kV vital buses per unit, and due to equipment concerns (NRC Unresolved Item 311/91-05-01), the undervoltage testing is being conducted on a weekly
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frequency. The subsequent licensee actions included (1) installing color coded test jacks to the jumper connection points so that the jumpers can be readily installed and removed, (2)
providing additional relay labelling to the underside of the relays, and (3) changing the associated test procedure to reflect the above test jack implementation and use. The inspector verified the implementation of the above changes and did not identify any deficiencie *
- . The "inspector noted that the licensee's previous corrective actions were not timely when considering the existing high testing frequency. However, the actions implemented.following this event appear to be effective in preventing future similar occurrences. Longer term corrective actions, as stated in previous licensee event reports and NRC inspection reports are also planned by the licensee. The inspector had no further question *
4. Hope Creek High Pressure Coolant Injection (HPCI) Initiation During Surveillance Testing On August 15, 1991, technicians were performing a drywell pressure (B21-N694A) channel calibration. The procedure required the installation and use of a test device which simulated*
the operation of a number of relay contacts in the logic circuitry. After receiving the fourth drywell high pressure alarm (as expected), the nuclear controls operator (RO licensed)
noticed the HPCI steam admission valve HV-FOOl stroking open. After verifying that other plant parameters were normal, the operator judged the initiation signal to be spurious and*
tripped the HPCI turbine before ir:ijection to ~e reactor vessel occurred. The HPCI system -
was then returned to its standby configuratio The licensee's immediate investigation did not reveal the cause of the spurious initiation signal. The test was rerun using a different test device with satisfactory results and no unexpected actuations. The licensee also determined that personnel error had not been involve The inspector verified that the test device is included in the licensee's measurement and test equipment (M&TE) program, although no periodic ciilibration of the device is required..
Extensive bench testing of the test device did not reveal any malfunction~.
As of the close of this reporting period, the licensee's investigation was ongoing. The inspector noted that the licensee's actions to date were both appropriate and extensive. The inspector had no further questions at this tim "D" Emergency Diesel Generator (EDG) Start Failure Followup As discussed in NRC Inspection Reports 354/91.-12 (Section 4.3.3.B) and 354/91-14 (Section 4.3.2), the licensee had been pursuing the cause of the May 22, 1991 failure of the "D" EDG to start as required during a surveillance test. In Special Reports 91-03-00 and 91-03-
1'1<
)
01 (Supplement), the licensee described the investigation which determined that probable cause of the failure was a lack of fuel boost when the start signal was received. Such a
- condition could have been caused by either mechanical failure. or a mispositioning of the minimum/maximum fuel position switch. The licensee found no mechanical failures and personnel who performed the surveillance tests stated that no repositioning of the fuel boost position switch occurred between test runs. The licensee no~ that the EOG had been successfully started eleven successive times since the failure and that the conditions leading to the failure could not be reproduced. Consequently, the licensee concluded that the root cause of the failure was indeterminate. Corrective actions* consisted of enhancing the applicable surveillance procedures by including a verification of proper fuel boost position switch position prior to the initiation signa *
- The inspector followed closely the licensee's efforts to identify the root cause of the start failure. In addition to activities noted in NRC Inspection Report 354/91-14, the inspector
_ reviewed both special reports and interviewed the EDG system engineer on a frequent basi The inspector concluded that; although the licensee was unable to definitely identify a root cause, his investigation had been rigorous and expansive, and included the participation of the diesel vendor (Colt Pielstich) and training personnel. The inspector noted that while the investigation did not rule out the possibility of a personnel error, evidence of such was* not found. As a preventive measure, the EDG system engineer intends to provide specific training to the equipment operators during their initial or requalification training, as
- appropriate, to enhance their knowledge of diesel generator performance and control.
EMERGENCY PREPAREDNESS Inspection Activity The inspector reviewed PSE&G's conformance with 10CFR50.47 regarding implementation of the emergency plan and procedures. In addition, licensee event notifications and reporting requirements per 10CFR50. 72 and 73 were reviewe.2 Inspection Findings Hurricane Bob Preparations A hurricane warning was issued for the nearby coastal areas during the period August 18-19, 1991. The site was forecast for high winds and tidal conditions. The licensee made preparations at both Salem and Hope Creek stations for this forecasted condition including:
Tracking the hurricane's progress, Reviewing Emergency Classification Guides (ECGs),
Implementing abnormal operating procedures, Monitoring meteorological instrumentation,
_
Inspecting the site and all outside areas for non-secure item **
Closing water tight doors, Briefing affected personnel on required actions, Verifying operability of offsite and emergency power sources, and Ensuring availability of diesel fuel oi The inspectors contacted each control room and discussed preparations_ with the on-shift senior nuclear shift supervisor. The inspectors also reviewed the associated procedures and ECGs, verified licensee actions, and provided site coverage. The inspectors concluded that the licensee was proactive in their approach to hurricane preparation Full Participation Onsite Accountability Drill
- On Friday, August 23, 1991, the licensee conducted a training drill at the Hope Creek station which included a full scale onsite assembly-and accountability scenario, which included all personnel within the protected area at both Hope Creek and Salem. The inspector participated in the drill and observed licensee performance in the Hope Creek Technical *
Support Center (TSC) from initial staffing to the drill critique. The inspector observed that the drill objectives (demonstrating a coordinated emergency response to simulated plant *
events and *a timely and accurate personnel accountability) were generally met with appropriate coaching and interruptions by the drill controllers. Although some of the emergency response team members were_ new, team performance appeared goo Personnel demonstrated proactive interest in the drill scenario. For example, although fuel pool cooling pump and heat exchanger: status was not reflected on any of the TSC status boards, engineering personn~l identified that fuel pool cooling would be lost and provided the emergency duty officer (EDO) with a conservative time estimate for the spent fuel pool temperature to reach the boiling point. A drill critique conducted at the end of the ex~rcise.
with the command team provided good feedback on the team's performanc.
SECURITY Inspection Activity PSE&G's conformance with the security program was verified on a periodic basis, including the adequacy of staffing, entry control, alarm stations, and physical boundaries. These inspection activities were conducted in accordance with NRC inspection procedure 7170.2 Inspection Findings No noteworthy findings were identified.
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- ENGINEERING/TECHNICAL SUPPORT Salem Failure and Replacement of Unit 1 J.{.eactor Coolant System Temperature Instrument On August 11, 1991, the Salem Unit 1 Reactor Coolant System (RCS) 13 loop oold leg narrow range resistance temperature detector (RTD) exhibited signs of intermittent failure; and the channel was declared inoperable. The following day, Salem Instrumentation and Control inspected and tested the RTD and its associated circuitry. The RTD performed satisfactorily and was declared operable. On August 13, 1991, however, the channel again behaved.erratically and was again declared inoperable. Subsequently; the circuit was again tested, and the RTD was found to have out-of-specification resistance readings. After the failed RTD was removed, a new spare RTD was resistance checked, documen_ted,* and installed while the unit was maintained at powe A newly installed RTD requires post-installation verification of RTD accuracy to within +l-0.5 degrees F. At Salem, RTD accuracy is normally verified during plant start-up, at normal operating temperature and zero power, with as near to isothermal equilibrium RCS conditions as possible established prior to the testing. In this case, with the unit operating at power, plant conditions prevented the gathering of the necessary data to utilize the normal methods of RCS RTD accuracy verification. RCS RTD accuracy is significant because the RTD provides an input to two reactor protection system setpoints, over-pressure differential temperature and over-temperature differential temperature.. The Westinghouse methodology
_for protection system setpoints (WCAP 12103) assumes an accuracy of +I- 0.5 degrees F for the temperature input *
To confirm the required RTD accuracy after installation, Salem system engineers and PSE&G corporate engineers, with consultation from Westinghouse, developed an alternative analysis to be performed in lieu of the normally performed RTD cross calibration. The method of confirmation consisted of the evaluation and analysis of vendor calibration reports and letters, reactor engineering state point data gathered with the plant at power, data from Westinghouse WCAP 12103, and the pre-installation bench check data of the replacement RTD. The licensee also performed an engineering review and safety evaluation of this method of RTD accuracy verificatio Data was collected at steady state conditions with the new RTD installed and compared to the expected T-cold reading, which was an average T-cold calculated from data from the four previous Unit 1 refueling cycles. The new RTD reading was found to be 0.308 degrees F above the calculated value, which was within one staridard deviation of the calculated average and the 0.5 degrees F accuracy requirement. Each of the other three loops also read slightly (from 0.125 to 0.458 degrees F) higher than their calculated value, indicating that some portion of the higher than expected reading of the new RTD was due to a slightly elevated
a slightly elevated RCS temperature. The licen8ee concluded that the process used provided adequate assurartce that the new RTD accuracy is. within the +/- 0.5 degrees F limit. The 13 loop cold leg narrow range temperature channel was declared operable on August 17,
' 19,9 '
The resident inspector followed the licensee resolution of this _matter from the time the original RTD was determined to be failed until the analysis of the replacement RTD verified it met the required accuracy standards. Through discussions with the involved PSE&G
engineers, attendance at several station management meetings at which the resolution of the problem was planned, and a review of the state point data results and engineering evaluation, the inspector determined that the licensee had used prudent engineering practices and a conservative safety approach in the restoration of the 13 loop T-cold channel. No inadequacies were *noted in the licensee's actions or oonclusions in this matte Unit 1 Shutdown Required By Technical Specifications Due to Equ_ipment Failure On September 5, 1991, an "auto-test fault" alarm was received at 5:09 p.m. from the lA safeguards equipment cab~net (SEC) at Unit 1. Control room operators attempted to reset the SEC per Operating Procedures; however, the SEC would not reset, rendering it inoperable. *
Technical Specification (TS) 3.3.2.1 was entere,d and a unit shutdown from 100% power was commenced per the TS Action requirements. The licensee replaced the installed lA SEC chassis with a spare chassis and satisfactorily completed a SEC functional test. The unit shutdown was terminated at 50% and the lA SEC was declared operable at 9:00 p.m. Unit operation was. unaffected by this event and the plant was subsequently returned to full powe The spare chassis was* the one which was previously removed from the lA SEC on August 15, 1991 (See Section 2.2.1.A of this report). The two failed circuit board cards had been replaced and the chassis was functionally tested satisfactoril The fospector reviewed the initial conditions prior to the event and. the licensee's event response, including immediate actions and conformance with TS requirements. *Prior to the lA SEC failure, multiple safety-related components were made inoperable at 12:42 a.m. on September 5, 1991, due to the tag-out of the No. 12 nuclear service water (SW) header for valve maintenance. Specifically, one of the two intermediate-head safety injection (SI)
pumps and one of the two high-head SI pumps were rendered inoperable due to the loss of SW supply flow for the pumps' lubricating oil coolers. The pumps were properly tagged out *
of service. Additionally, other safety pumps were rendered inoperable (but remained available) due to the loss of SW supply flow to the associated room coolers. The operability of one residual heat removal and two compo'nent cooling water system pumps were technically.affected by the room coolers being out of servic The inspector also reviewed impact of the loss of the lA SEC as related to equipment already made inoperable due to the SW header outage and the TS applicability required actions. The inspector concluded that'the appropriate TS Action requirements were properly
entered and implemented. The inspector noted that the loss of the lA SEC rendered the automatic starting and load sequencing for the lA emergency power source inoperabl However, station abnormal and emergency* opeIC!-ting procedures have provisions which direct operators to manually start and load the diesel generators if required under accident conditions. The inspeetor concluded that redundant system components remained available throughout this event, procedures adequately addressed postulated design basis conditions, and plant safety was not compromise.2 Hope Creek Filtration, Recirculation and Ventilation System (FRVS) Heater Failure Update During this reporting period the licensee continued the implementation of corrective actions to resolve the issues surrounding the May and July, 1991, FRVS heater failures, as discussed in NRC Irispection Report 354/91-14, Section 7.2.A. The licensee's investigation following the July 6, 1991, fuse failures determined that a build-up of heat inside the panels during the*.
ten hour surveillance run caused a degradation in the current carrying capability of the fuses to a level below the fuse rating. As an interim fix, the heater doors were removed from the recirculation and vent fan panels. Additionally, non-essential heat producing components (e.g., disconnect switch and indicating relays) identified by thermography were de".'energized or remove On July 30, 1991, the panel vendor, Nutherm International, informed the NRC of a potential deviation in the design safety function of the FRVS panels. Following their evaluation, CFR21 notification was made on August 9, 1991. This notification, however, applied only to the "A" vent and "D" recirculation panels (1AC045 and 1DC043 respectively) and stated that with the doors removed the two panels were operable for both normal and accident condition The inspector noted that test data had been developed by the licensee using two panels as representative of the eight affected panels (1AC043:-1FC043, 1AC045 and 1BC045);
therefore, the Part 21 determination should have included the other six panels. The licensee agreed and stated that their concerns had been communicated to the vendor by letter dated August 19, 1991. The inspector reviewed licensee arid vendor material documenting testing, environmental qualification (EQ), and reportability issues, including Licensee Event Reports (LER)_ 354/90-07-01 and 91-07-02, and concluded that the licensee's root cause investigation, and corrective actions were thorough and appropriately addressed the outstanding issues
concern!ng 10CFR21 reportability and-degraded EQ of ce~n components. LER 354/91-07-02 appeared to adequat~ly document the resolution of the FRVS heater design versus actual required capacities and the effects of degraded EQ components.. While the licensee's initial root cause analysis for the May 1991 fuse failure was inadequate, actions undertaken as a result of the July 1991 failures were extensive, thorough and appropriately managed. The licensee continues pursuing resolution of the Part 21 issue with Nuther **
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18 Reactor Building Ventilation Backdraft Isolation Damper Investigation In September 1990, the licensee completed an investigation into. the spare parts inventory for reactor building* supply and exhaust ventilation systems backdraft steam isolation damper These are designed to prevent steam from a postulated pipe break from entering non-affected areas of the building through the ventilation ductwork. The results of the licensee's investigation were documented in a letter (SCI-90-0371) dated September 21, 1990, detailing a number of apparent discrepancies relating to environmental qualification (EQ), document inaccuracies, and available spare part *
An earlier evaluation of $e ductwork in February-April, 1988, had determined that none of the 26 pairs of backdraft dampers were included in the EQ program and that substantial further evaluation and documentation would be required to assess the impact on the affected *
systems. The licensee initiated the appropriate engineering efforts to resolve these issues, efforts which were in progress at the time of the 1990 investigation. The licensee
determined that only three of the 26 pairs of backdraft isolation dampers should be (and consequently were entered) in the EQ program. All three were in the Filtration, Recirculation and Ventilation System (FRVS). The equipment qualification maintenance and surveillance (EQMS) information sheet for these dampers, M717-DMPR-004, lacked all the *
appropriate EQ data. For example, revision 0 of this data sheet specified a ten-year replacement interval, but listed the gasket material as unknown.
. The inspector reviewed revision 1 (approved on May 1, 1991) to M717-DMPR-004 and verified that the gasket material was specified and that other discrepancies noted by the 1990 investigation appeared adequately addressed. Justification for the use of four caulk type compounds and a ten-year life time was documented in an April 25, 1991 memorandum to the M717 EQ file. The licensee is currently compiling and will procure the appropriate spare parts for all. the backdraft dampers; Completion of procedure enhancements and documenf updates is tentatively scheduled for the end of 1991. The inspector considered this time frame appropriate considering' the minor safety significance of the remaining issue.
SAFETY ASSESSMENT/QUALITY VERIFICATION Salem Significant Event Response Team (SERT) Report Review During this report period, the resident inspector staff reviewed two SERT reports that had been prepared for the General Manager-Salem Operations. SERT Report No. SSR 91-03 reported the investigation of the Salem.Unit 1 lB Vital Bus Undervoltage (UV) Relay events of June 6 and June 13,1991,(see NRC Inspection Report 272/91-15, Section 4.3.2.C) and SERT Report No. SSR 91-04 assessed the Salem Unit 1 Reactor Trip and Lightning Strike of June 16, 1991 (see NRC Inspection Report 272 & 311/91-19, Section 2.2.3).
The reports documented the findings, condusions and recommendations of the two SERTs that had been formed in accordance with PSE&G Nuclear Administrative Procedure NC.NA-AP.ZZ-0061 (Q), "Significant Event Response Team Management." This procedure states that the purpose of a SERT is to provide for "independent assessment of selected events, trends or certain repetitive situations" and to "en~ure that all relevant aspects of an event or
situation have been considered and appropriate corrective actions identified to prevent recurrence. "* The inspectors' review of the two reports revealed that both SERTs had accomplished these functions and adequately reviewed each event; a good questioning attitude and a proper safety perspective was note *
The inspectors identified one weakness in SERT Report No. SSR 91-03. The SERT did not identify the fact that a certain human engineering deficiency (the location of.the relays and the lack of test jacks for surveillance testing of the UV relays), had a direct impact on the June 6, 1991 event, and was previously identified by Salem technicians. Th_e correction of this deficiency might have prevented these events. When informed by the inspector of this finding, the licensee ackn*owledged that the information should have *been considered in the report and that it would be considered in the resolution of the UV relay testing concer Other than the one identified weakness, the inspector concluded that the SERT process had*
been effectively utilized in the assessment of the two events.
8.2. *Hope Creek Hope Creek Comprehensive Scram Review Following the May 7, 1991 unplanned scram at Hope Creek (NRC Inspection 354/90-12), a team from Onsite Safety Review, Off site Safety Review, Station Quality Assurance and Human Performance Enhancement Syste_m reviewed the 12 scram events since August 26, 1988. The licensee's team used the Management Oversight and Risk Tree (MORT) proces This team assessed event-specific and management:..related factors that contributed to or allowed the scrams to happe The team's conclusions and recommendations focused on the following:
Establishing scram reduction responsibility, Communicating scram-specific quality expectations, Establishing employee involvement and feedback processes, Re-emphasizing the Scram and Power Reduction Elimination Committee (SPRE),
Ensuring t_hat training and job content are in a climate of procedure adherence and routinized work,
,.
Enhancing balance-of-plant maintenance, and effecting latent error reduction in all maintenance,
, Implementing Significant Event Review Team (SERT) recommendations, Balancing plant versus people-oriented *corrective actions, and Reviewing the value of the action tracking *syste The inspector reviewed the final report dated July 29, 1991. The inspector concluded that the licensee's assessment, conclusions and recommendations appeared appropriate. Overall, this effort was thorough, well managed, and effectively conducted. - LICENSEE EVENT REPORTS (LER), PERIODIC AND SPECIAL REPORTS FOLLOWUP PSE&G submitted the following licensee event reports, and special and periodic reports, _
which were reviewed for accuracy and evaluation adequac Special and Periodic Reports Semi-Annual Fitness For Duty Performance Data dated August 19, 199 Salem and Hope Creek Monthly Operating Reports for July 199 Salem Unit 1 Inservice Inspection Activities for the Ninth Refueling Outage (NLR-N91128).
Salem and Hope Creek Semi-Annual Effluent Release Reports for period January 1 to June 30, 199 Salem Unit 1 Special Report 91-2 addressed the "A" Reactor Trip Breaker failure of July 25, 1991 (See NRC Inspection Report 272/91-19, Section 2.2.1.B).
Hope Creek Special Report 91-03-01 (Supplement) See Section 4.3. No unacceptable conditions were note Salem LERs Unit 1 LER 91-07, Revision 1 updated an event regarding automatic starting of the motor driven auxiliary feedwater pumps during an outage. The event was reviewed in NRC Inspection 272/91-05. No inadequacies were noted relative to this LE,,;
LER 91-08, Revision 1 updated a reactor protection system actuation while shut down in Mode 5. The event was reviewed in NRC Inspection 272/91-05. No inadequacies were noted relative to this LE * LER 91-21, Supplement 1, addressed an additional containment penetration overcurrent protection device Technical Specification 3/4.8.3.1 noncompliance discovered after the submission of the original-LER (See NRC Inspection Report 272/91-19, Section 4. 3.1. C).
No inadequacies were noted relative to this supplemen LER 91-24 addressed the Unit 1 reactor trip due to a lightning strike on June 16, 1991, which was discusSed in NRC Inspection Report 272/91-19, Section 2.2.3. No inadequacies were noted relative to this LE *
LER 91-25 concerned a radiation monitor spike (1R45C) and containment ventilation isolation on July 27, 1991, due to equipment failure.* No inadequacies were noted relative to this LER:
LER 91-26 addressed the No. 13 steam generator pressure protection channel inoperability resulting from the lead-lag switch incorrect setting, which is discussed in Section 4.3.1.A of*
this repor *
LER 91-27 (See Section 2.2. l.A)
Unit 2 LER 91-06, Revision 1 updated an event regarding failure of a control roorri radiation monitor due to equipment failure. The event was reviewed in NRC Inspections 311/91-05, 09. No inadequacies were noted relative to this LE. LER 91-08 addressed the vital 4KV bus undervoltage relays that were *found to have setpoints below the Technical Specification minimum allowed. value (see NRC Inspection Report. 311/91-15, Section 4.3.2.E). No inadequacies were noted relative to the LE LER 91-09 addressed the spurious start of the No. 21 motor driven auxiliary feedwater pump of June 30, 1991, which was reviewed in NRC Inspection Report 311/91-19, Section 2.2.1.A. No inadequacies were noted relative to the LE LER 91-11 addressed the July 30, 1991, discovery that the non-radioactive liquid waste discharge radiation moni~or system (RMS) channel 2R37 setpoint was not in compliance with Technical Specification 3.3.3.8. Higher capacity pumps had been installed in the system in June 1985, and the corresponding RMS setpoint changes were not implemented until _the July 30, 1991 discovery. The licensee attributed the root cause of the event to inadequate design
. *
review and subsequently completed a design change to modify the 2R37 setpoint. The
- inspector reviewed the licensee actions, determined them to be adequate, and noted no inadequacies relative to the LE LER 91-10 concerned a radiation monitor spike (2R45C) and containment ventilation
. isolation on July 23, 1991, due to equipment failure. No inadequacies were noted relative to this LE Hope Creek LERs LER 91-07-01 (See Section 7.2.A).
LER 91-07-02 (See Section 7.2.A).
LER 91-16 described an isolation of the Reactor Core Isolation Cooling (RCIC) System due to a spurious signal from the steam leak detection system. A NUMAC temperature sensor card associated with the inboard RCIC steam isolation valve failed high; the resulting signal caused the valve to fully close. After determining that the signal was spurious, repairs were initiated and RCIC returned to operable status. The safety significance of this first-time event was minimal. No *significant discrepancies were noted in this LE *
1 EXIT INTERVIEWS/MEETINGS 1 Resident Exit Meeting The inspectors met with Mr. C. Vondra and Mr. R. Hovey and other PSE&G personne periodically and at the end of the inspection report period to summarize the scope and findings of their inspection activitie Based on Region I review and discussions with PSE&G, it was determined that this report does not contain information subject to 10 CPR 2 restriction.2 Specialist Entrance and Exit Meetings Inspection Reporting Date(s)
Subject Report N Inspector 8/12-20/91 Emergency 272&311/91-24;
- Preparedness 354/91-17 Amato 8/19-23/91 Radiological 354/91-15 Mann Control Environmental 272&311/91-22; Monitoring 354/91-18 Peluso
8/26-30/91
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1 Hope Creek Meeting Motor Operated Valve (MOV) Program Enforcement Conference An enforcement conference was held on September 9, 1991, in the NRC Region I King of Prussia office to discuss a number of issues arising from NRC Inspection 354/91-80, conducted July 15-19, 1991, at Hope Creek, pertaining to the operator, testing, and safety evaluation of Motor Operated Valves as described in NRC Generic Letter 89-10. Attachment 1 is a list of attendees, and Attachments 2 and 3 describe the licensee's presentation. The NRC's conclusions from this conference will be provided to the licensee in a separate correspondence.
ATTACHMENT 1 ENFORCEMENT CONFERENCE LIST OF ATTENDEES
- September 9, 1991 NUCLEAR REGULATORY COMMISSION J. Wiggins, Deputy Director, Division of Reactor Projects (DRP)
W. Lanning, Deputy Director, Division of Reactor Safety (DRS)
R. Blough, Chief, Reactor Projects Branch 2, DRP J. White, Chief, Reactor Projects Section No. 2A, DRP P. Eapen, Chief, Systems Section, DRS J. Durr, Chief, Engineering Branch, DRS K. Lathrop, Resident Inspector R. Matakas, Investigator, Office of Investigations D. Holody, Enforcement Officer J. Yerokun, Project Engineer B. Westreich, Reactor Engineer
, K. Smith, Regional Counsel W. Butler, Project Director, NRR S. Dembek, Project Manager, NRR J. Colaccino, Mechanical Engineer, NRR E. Sullivan, Section Chief, NRR T. Scarbrough, Senior Mechanical Engineer, NRR PUBLIC SERVICE ELECTRIC AND GAS COMPANY S. Miltenberger, Vice President & Chief Nuclear Officer T. Crimmins, Jr., Vice President - Nuclear Engineering J. Hagan, General Manager - Hope Creek L. Reiter, General Manager - QA & Nuclear Safety Review D. Jagt, Manager - Nuclear Engineering Design G. Englert, Jr., Nuclear Engineering Standards Manager J. Ranalli, Mechanical Engineering Manager R. Brown, Principal Engineer - Licensing F. Thomson, Manager - Nuclear Licensing & Regulation K. Suomi, Senior Nuclear Maintenance Supervisor R. Binz, Principal Engineer OTIIER M. Sesok, Atlantic Electric Site Representative C. Dell, Nuclear Engineer, State of New Jersey
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- ps~G Public service
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- Electric and Gas
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Company
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NRC ENFORCEMENT CONFERENCE ATTACHMENT 2 MDV PROGRAM SEPTEMBER 9, 1991 HOPE CREEK 6BERATIN8 STATION
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AGENDA MDV PROGRAM INTRODUCTION/OVERVIEW SAFETY ASSESSMENT RESULTS PROGRAM DESCRIPTION PROGRAM DESCRIPTION DEVIATION SUPPLEMENT 3 RESPONSE TORQUE SWITCH SETTING VENDOR INFORMATION CONTROL PSE&G ASSESSMENT OF POTENTIAL VIOLATIONS SUMMARY T. M. CRIMMINS J.. A. RANALLI J. A. RANALLI J. A. RANALLI J. A. RANALLI J. J. HAGAN G. E. ENGLERT F. X. THOMSON
- T. M. CRIMMINS
- MOV PROGRAM INTRODUCTION/OVERVIEW OVERVIEW OF INSPECTION FINDINGS
- POTENTIAL DEVIATION-THE LICENSEE HAD NOT ESTABLISHED A DETAILED GL 89-10 PROGRAM DESCRIPTION BY JANUARY 1, 1991 AS COMMITTED
- POTENTIAL VIOLATIONS
-PROVIDING INACCURATE AND INCOMPLETE INFORMATION TO THE NRC IN LICENSEE LETTER DATED MARCH 8, 1991 CONTRARY TO 10CFR5 MODIFICATION OF MOV TORQUE SWITCH SETTINGS WITHOUT AN ENGINEERING OR SAFETY EVALUATION-FAILURE TO REVIEW. EVALUATE AND INCORPORATE
- cERTAIN VENDOR INFORMATION WHICH PROVIDED INFORMATION FOR THE MAINTENANCE OF SAFETY-RELATED MOVs
MOV PROGRAM OVERALL SAFETY. ASSESSMENT
- DETAILED REVIEW OF SUPPLEMENT 3 ANALYSIS VERIFIES VALVES WILL CLOSE UNDER*DBA CONDITIONS
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- SUBSTANTIAL MARGINS EXIST DUE TO TORGUE SWITCH BYPASS CIRCUITRY-ENHANCES VALVE THRUST CAPABILITY TO ACHIEVE DISK-TO-SEAT OVERLAP-ALLOWS ACCOMODATION OF HIGHER VALVE FACTORS*
-REMOVES RATE OF LOADING AS A CONCERN FOR THE DURATION OF BYPASS-VALVE ASSEMBLY HAS ADEQUATE STRUCTURAL CAPABILITY TO WITHSTAND STRESSES
- SUBSEQUENT EVALUATIONS CONFIRM OPERABILITY CONCLUSIONS 91EC1-1B
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MDV PROGRAM-SCHEDULE 1989 l
1990 I 1991 1992 I
1993 1994 I
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OFiIGINAL I f HAsE I II 1 j PHASE III I
I I
9 12
i PROCW.M DE8CllPTION.
1989 I 1990 I, 1991 I 1992 I 1993 1994 PRESENT
PHASE II II I I PHA~E II 12.
10
i PROaRAM DE8CRtPnoN
":111: [, 1 -.~ 1
MDV PROGRAM PROGRAM DESCRIPTION PHASE I ACTIVITIES
- DEVELOP ENGINEERING REQUIREMENTS AND IMPLEMENTATION PLANS
- IDENTIFY 15 MOVs FOR DETAILED ANALYSIS
- 9 SALEM VALVES
- 6 HOPE CREEK VALVES
- IDENTIFY SCOPE ITEMS FOR INCLUSION IN PHASE II PROGRAM 91EC1-19A
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MOV PROGRAM PROGRAM DESCRIPTION LESSONS LEARNED - PHASE I
- PRINCIPAL ENHANCEMENTS IDENTIFIED FOR PHASE I EC1-20 INCLUDE:
-ENHANCEMENT OF CURRENT NUCLEAR DEPARTMENT POLICY GOVERNING OVERALL MAINTENANCE PROGRAM-DEVELOPMENT OF PROGRAMMATIC STANDARDS FOR CONTROL OF FUTURE MOV ACTIVITIES
- -CONSOLIDATION AND RECONCILIATION OF EXISTING DATA SOURCES
- DEVELOPMENT OF IMPROVED METHODS AND PROCEDURES FOR MOV TESTING AND MAINTENANCE
MDV PROGRAM PROGRAM DESCRIPTION PHASE II ACTIVITIES
- IMPLEMENT POLICY AND FORMAL PROGRAM REGUIREMENTS TO MEET GL 89-10 RECOMMENDATIONS -
-DEVELOP PROGRAM IMPLEMENTING PROCEDURES A MANAGEMENT POLICY
- DATA COLLECTION/RECONCILIATION
- FUNCTIONAL EVALUATIONS
- CORRECTIVE ACTION
- TRENDING-DEVELOP IMPROVED MAINTENANCE METHODS
- TESTING AND MAINTENANCE
- ESTABLISH DETAILED SCHEDULES
- COMPLETION BY JUNE 1994
91EC1-~98
MDV PROGRAM PROGRAM DESCRIPTION MOV PROGRAM PROJECT PLAN
- PROJECT CONTROLS NI> FUNCTIONAL RESPONSIBILITIES
- ORGANIZATIONAL MA TRIX MOV PROGRAMMATIC STANDARD ~C.DE-PS.ZZ-0033(G)
- APPENDICES OF PROGRAMMATIC STANDARD 91EC1-34
- APPENDIX Ai
.VALVE POPULATION
- APPENDIX A2 VALVE PRIORITIZATION.
- APPBl>IX A!
WALKDOWN DATA COLLECTION-APPOOIX A~
OPERATING CONJITION EVALUATION
- APPENDIX A!5 ELECTRICAL CAPABILITY REVIEW
- APPENDIX AS MECHANICAL CAPABILITY REVIEW
- APPENDIX A7 DIAGNOSTIC TEST DATA REVIEW
- APPENDIX AS CAPABILITY ASSESSMENT
"'" APPENDIX A9 DATABASE SYSTEM
- APPENDIX A10 MOVATS DIAGNOSTIC DATA CONSOLIDATION
- APPOOIX A11 SOFTWARE REQUIREMENTS SPECIFICATIONS
- APPOOIX A12 _ DATA COLLECTION SPECIFICATION
- APPROVAL OF THESE DOCUMENTS IN TOTAL WILL BE ACCOMPLISHED BY 10/31/91
- ONGOING ACTIVITIES:
- DATA GA THEA ING-MOV EVALUATIONS TO SUPPORT MAINTENANCE ACTIVITIES
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MDV PROGRAM
_ PROGRAM DESCRIPTION DEVIATION NRC FINDING
- PSE&G FAILED TO MEET ITS COMMITMENT TO ESTABLISH AN APPROVED PROGRAM DESCRIPTION BY JANUARY 1. 1991 PSE&G RESPONSE
- WE AGREE WITH-THE DEVIATION AS STATED ROOT CAUSE OF DEVIATION
- WE FAILED TO RECOGNIZE AND ADDRESS THE OVERCOMMITMENTS ON OUR VALVE ENGINEERING RESOURCES IN A TIMELY MANNER
- TURNOVER OF KEY PERSONNEL-DIVERSION OF RESOURCES TO OTHER SIGNIFICANT ISSUES
- WE FAILED TO UTILIZE OUR MONITORING AND CONTROL PROCESSES FOR TRACKING OUR GL 89-10 PROGRAM DESCRIPTION COMMITMENT 10 MONTH SLIP IN SCHEDULE RESULTED 91EC1-15
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MOV PROGRAM PROGRAM DESCRIPTION DEVIATION CORRECTIVE ACTIONS
- VICE PRESIDENT - NUCLEAR ENGINEERING (VPNE) HAS COUNSELED ALL MANAGEMENT PERSONNEL DIRECTLY INVOLVED
- VPNE LETTER ISSUED TO ALL NUCLEAR ENGINEERING PERSONNEL REITERATING EXPECTATIONS
- CONTROL OF COMMITMENTS
- SCHEDULE ADHERENCE-TIMELY RECONCILIATION OF RESOURCE ISSUES-REQUESTED MANAGERS TO REVIEW OTHER REGULATORY PROGRAMS TO ENSURE SIMILAR PROBLEMS DO NOT EXIST I
- AN INDEPENDENT REVIEW HAS BEEN INITIATED TO ASSESS THE EFFECTIVENESS OF OUR REGULATORY COMMITMENT TRACKING PROCESS (12/31/91)
- PROJECT PLAN WAS RE-ESTABLISHED IN OCTOBER 1990
- PHASE I ACTIVITIES COMPLETE
- PHASE I I. ACTIVITIES UNDERWAY-FINAL COMPLETION IN JUNE 1994
- MDV PROGRAM SUPPLEMENT 3 RESPONSE NRC FINDINGS PSE&G STATED THAT RESPONSE TO SUPPLEMENT 3 WAS BASED ON COMPREHENSIVE EVALUATION OF THRUST REQUIREMENT THE FOLLOWING DISCREPANCIES AND INACCURACIES WERE NOTED BY THE NRC DURING THE INSPECTION:
- THRUST VALUES REPORTED WERE NOT DERIVED FROM DETAILED EVALUATIONS BUT WERE OBTAINED FROM VENDOR DATA SHEETS
- * PSE&G HAD 3 SETS OF EVALUATIONS INDICATING THAT
-*
REPORTED THRUST VALUES MIGHT BE
. NON-CONSERVATIVE
- LACK OF TECHNICAL JUSTIFICATION FOR USE OF NON-CONSERVATIVE DISK AND STEM FRICTION FACTORS eREPORTED THRUST MARGINS DID NOT INCLUDE CONSIDERATION OF INSTRUMENT UNCERTAINTY AND RATE OF LOADING.
- REPORTED THRUST VALUES WERE NOT BASED ON DESIGN BASIS DP TESTIN REPORTED THRUST VALUES WERE BASED ON TORQUE SWITCH TRIP, WHEREAS TORQUE SWITCH IS BYPASSED DURING ACCIDENT CONDITIONS 91EC1-14
. ~
'*
MOV PROGRAM SUPPLEMENT 3 RESPONSE.
PSE&G RESPONSE
- RESPONSE WAS 1 COMPLETE AND ACCURATE 1 BASED UPON OUR ENGINEERING JUDGEMENT AND EVALUATION
- DIFFERING PROFESSIONAL OPINIONS EXIST WITHIN THE INDUSTRY ON CERTAIN TECHNICAL ISSUES
- WE FAILED TO ADEGUATELY COMMUNICATE THE BASIS FOR OUR CONCLUSION OF OPERABILITY-SUBSEGUENT RE-EVALUATION OF DATA CONFIRMED OPERABILITY CONCLUSION STATED IN OUR
. RESPONSE
(
MOV PROGRAM SUPPLEMENT 3 RESPONSE NRG _FINDINGS
. *THRUST VALUES REPORTED WERE NOT DERIVED FROM DETAILED EVALUATIONS BUT WERE OBTAINED FROM VENDOR DATA SHEETS
- PSE&G HAD 3 SETS OF DATA INDICATING*THAT.
91EC1-23 REPORTED THRUST VALUES MIGHT BE NON-CONSERVATIVE
- MOV PROGRAM SUPPLEMENT 3 RESPONSE PSE~G.RESPONSE REPORTED THRUST VALUES WERE TAKEN FROM VENDOR VALVE DATA SHEETS
- REPRESENTED 1 BEST AVAILABLE 1 DESIGN INFORMATION SELECTION OF VENDOR THRUST VALUES FOR USE IN SUPPLEMENT 3 WAS BASED ON ENGINEERING REVIEW OF AVAILABLE INFORMATION
- VENDOR DATA -
- *
- PHASE I RESULTS
- SAFETY ASSESSMENT ENGINEERING REVIEW OF AVAILABLE INFORMATION IDENTIFIED THE FOLLOWING:
- VENDOR THRUST VALUES WERE CONSERVATIVELY DERIVED
- DIFFERENTIAL PRESSURE
- UNDERVOL T AGE
- PHASE I RESULTS INCLUDED CONSERVATIVE ASSUMPTIONS
- WELD END PREP vs SEAT DIAMETER-BYPASS CIRCUITRY NOT INCLUDED
- SAFETY ASSESSMENT CREDITED BYPASS CIRCUITRY FOR PROVIDING SIGNIFICANT THRUST CAPABILITY BEYOND MINIMUM REQUIRED THRUST 91EC1-24A
- MOV PROGRAM SUPPLEMENT 3 RESPONSE PSE&G RESPONSE (CONT)
DRAFT ENGINEERING RESPONSE TO SUPPLEMENT 3 ESTABLISHED THE BASIS FOR OUR OPERABILITY CONCLUSION
- RECONCILED DIFFERENCES IN MINIMUM THRUST VALUES ON BASIS OF CONSERVATIVE ANALYSIS ASSUMPTIONS
- CONCLUDED DESIGN BASIS VALUES (i.e. VENDOR DATA) REPRESENTED "BEST AVAILABLE* INFORMATION
- IDENTIFIED TORQUE SWITCH BYPASS CIRCUITRY AS
PROVIDING ADDITIONAL ASSURANCE OF VALVE CLOSURE CAPABILITY UNDER OBA CONDITIONS
-ALLOWS THRUST UP TO MOTOR RATED TORQUE-GREATLY EXCEEDS TORQUE SWITCH TRIP THRUST WE* FAILED TO ADEQUATELY COMMUNICATE THIS BASIS IN OUR RESPONSE TO SUPPLEMENT 3 BASED ON RE-EVALUATION OF DATA AVAILABLE PRIOR TO 3/8/91 AND THE OPERABILITY BASIS DESCRIBED IN OUR DRAFT RESPONSE, WE CONCLUDE THAT OUR OPERABILITY DETERMINATION PROVIDED IN RESPONSE TO SUPPLEMENT 3 REMAINS VALID
- RECENT EVALUATIONS CONTINUE TO CONFIRM THE OPERABILITY OF THESE VALVES 91EC1-24B
MOV PROGRAM SUPPLEMENT 3 RESPONSE NRC FINDING LACK OF TECHNICAL JUSTIFICATION FOR USE OF NON-CONSERVATIVE DISK AND STEM FRICTION FACTORS PSE&G RESPONSE
- CURRENT DISK AND STEM FRICTION FACTORS ARE THE RESULT OF BEST AVAILABLE VENDOR RECOMMENDATIONS *
- OUR PHASE II PROGRAM WILL RE-EVALUATE THE VALVE FACTORS TO BE USED
- DIFFERING PROFESSIONAL OPINIONS EXIST WITHIN THE INDUSTRY ON THE EXACT VALUES TO BE USED
- 151 + 51 MARGIN IS ASSIGNED TO ACCOUNT FOR ENGINEERING UNCERTAINTIES AND EQUIPMENT INACCURACIES
- BYPASS OF TORGUE SWITCH ENHANCES VALVE THRUST*
CAPABILITY, THEREBY ALLOWING HIGHER VALVE FACTORS TO BE ACCOMODATED 91EC1-25
- MOY PROGRAM SUPPLEMENT 3 RESPONSE NRC FINDING REPORTED THRUST MARGINS DID NOT INCLUDE CONSIDERATION OF INSTRUMENT UNCERTAINTY AND RATE OF LOADING PSE&G RESPONSE
- REPORTED THRUST MARGINS WERE NOT ADJUSTED FOR INSTRUMENT INACCURACIES NOR DID THEY ADDRESS RATE OF LOADING EFFECTS
- 151 + 51 MARGIN IS ASSIGNED TO ACCOUNT FOR
INSTRUMENT INACCURACIES AND ENGINEERING UNCERTAINTIES
- ACTUAL MARGINS ARE GREATER CONSIDERING TORQUE SWITCH BYPASS CIRCUITRY
- -ACTUAL DEVELOPED THRUST COULD BE EQUIVALENT TO MOTOR RATED TORQUE-DEFEAT OF BYPASS OCCURS AFTER DISK-TO-SEAT OVERLAP IS ACCOMPLISHED
- RATE OF LOADING IS NOT A CONCERN WHEN THE TORQUE SWITCH IS BYPASSED.
- FINAL DISPOSITION OF THESE ISSUES IS REQUIRED UNDER PSE&G PHASE II PROGRAM 91EC1-26
--
MOV PROGRAM SUPPLEMENT 3 RESPONSE NRC FINDING REPORTED THRUST VALUES WERE NOT BASED ON DESIGN BASIS DP TESTIN REPORTED THRUST VALUES WERE.
BASED ON TORQUE SWITCH TRIP, WHEREAS TORQUE SWITCH IS BYPASSED DURING ACCIDENT CONDITION.
'
.
PSE&G RESPONSE
- REPORTED THRUST VALUES REFLECTED MOST RECENT DIAGNOSTIC RESULTS OBTAINED UNDER STATIC CONDITIONS.
-REPORTED THRUST VALUES WERE BASED ON TORQUE SWITCH TRIP.
- IN-SITU DESIGN BASIS DP TESTING IS NOT PRACTICABLE FOR THESE VALVES
- BYPASS CIRCUITRY PROVIDES ADDITIONAL CAPABILITY -
TO CLOSE VALVES 91EC1-27-ALLOWS THRUST UP TO MOTOR RATED TORQUE CAPABILITY-MOTOR RATED TORQUE IS MUCH GREATER THAN TORQUE SWITCH TRIP THRUST-DIAGNOSTIC TRACES SHOW SUBSTANTIAL DISK-TO-SEAT OVERLAP WHEN BYPASS IS DEFEATED
- MOV PROGRAM
SUPPLEMENT 3 RESPONSE ACTIONS TAKEN TO DATE
- RE-EVALUATED EXISTING DATA TO CONFIRM NO OPERABILITY*
ISSUE
- AN INDEPENDENT ASSESSMENT HAS BEEN CONDUCTED BY OUR NUCLEAR SAFETY REVIEW GROUP TO EVALUATE THE CIRCUMSTANCES LEADING TO OUR SUPPLEMENT 3 RESPONSE
- VPNE HAS DISCUSSED THE ISSUE WITH DIRECT REPOR-TS AND REITERATED HIS EXPECTATIONS RELATIVE TO COMMUNICATIONS WITH THE NRC ACTIONS TO BE COMPLETED
- GL 89~10 SUPPLEMENT 3 RESPONSE WILL BE RESUBMITTED BY 9/30/91 TO CLARIFY THE BASIS FOR OUR OPERABILITY DETERMINATION
- LETTER TO BE ISSUED TO ALL NUCLEAR DEPARTMENT PERSONNEL-TO EMPHASIZE OBLIGATION TO PROVIDE CLEAR INFORMATION IN COMMUNICATIONS WITH NRC
- TO EMPHASIZE THAT TECHNICAL ISSUES MUST BE ADEQUATELY EXPLAINED
- LETTER WILL ALSO BE DISCUSSED WITH PERSONNEL THROUGHOUT THE NUCLEAR DEPARTMENT VIA *ROLLDOWN* BY MANAGEMENT/SUPERVISION
- TRAINING TO BE HELD WITH ALL LICENSING DEPARTMENT PERSONNEL TO ENSURE EXPECTATIONS/RESPONSIBILITIES ARE UNDERSTOOD RELATIVE TO COMMUNICATIONS WITH THE NRC 91EC!-37
.
-~
MOV PROGRAM SUPPLEMENT 3 RESPONSE SUMMARY
- OUR RESPONSE TO GENERlC LETTER 89-10, SUPPLEMENT 3 WAS COMPLETE ANJ ACCURATE BASED ON ENGINEERING JUDGEMENT ANO EVALUATION AND THEREFORE NOT IN VIOLATION OF 10CFR5 *WE FAILED TO ADEQUATELY COMMUNICATE THE BASIS FOR OUR OPERABILITY CONCLUSION
- NO INTENT TO MISLEAD THE NRC*
- SUBSTANTIAL MARGINS EXIST TO ASSURE PROPER VALVE FUNCTION UNDER DESIGN BASIS ACCIDENT CONDITIONS
- TECHNICAL CONCERNS RAISED IN GENERIC LETTER 89-10 WILL BE ADDRESSED UNDER OUR PHASE II PROGRAM-DIAGNOSTIC EQUIPMENT INACCURACIES
- APPROPRIATE VALVE FACTORS
- ACTIVELY INVOLVED IN INDUSTRY GROUPS FOLLOWING MOV ISSUES
- WE WILL RESUBMIT OUR SUPPLEMENT 3 RESPONSE TO CLARIFY THE BASIS FOR OUR OPERABILITY DETERMINATION BY 9/30/91 91EC1-28
MOV PROGRAM TORQUE SWITCH SETTING NRC FINDING
- MODIFICATION OF MOV TORGUE SWITCH SETTINGS WITHOUT A DOCUMENTED ENGINEERING OR SAFETY EVALUATION DESCRIPTION OF DEFICIENCY
- LIMITER PLATES REMOVED FROM 2 RCIC VALVE MOTOR OPERATORS
- TORQUE SWITCH SETTINGS INCREASED TO OBTAIN i
REGUIRED-THRUST
.
- NO DOCUMENTED ENGINEERING EVALUATION PERFORMED PSE&G RESPONSE
- MOV P~OGRAM*
TORQUE SWITCH SETTING EVALUATION OF DEFICIENCY
- SWITCH SETTINGS UNDER ADMINISTRATIVE CONTROL OF MAINTENANCE. PROCEDURE-SYSTEM ENGINEER APPROVAL WITHIN PROCEDURE-LIMITORQUE AND VALVE MANUFACTURER CONCURRENCE OBTAINED
- SWITCH SETTING INTENDED TO PROVIDE OESISH BASIS SEATING THRUST-LIMITER PLATE SETTING BASED ON VENDOR CALCULATION
&LESS ACCURATE THAN DIRECT MEASUREMENT
- NO DOCUMENTATION OF EVALUATIONS PERFORMED 91EC1-29
- MOV PROGRAM
TORQUE SWI-TCH SETTING ROOT CAUSE
- PAST PROCEDURAL CONTROLS FAILED TO ADDRESS DOCUMENTATION REQUIREMENTS CORRECTIVE ACTIONS 91EC1-30
- MAINTENANCE PROCEDURES REVISED TO REQUIRE DEFICIENCY REPORT (DR) INITIATION TO RE-EMPHASIZE DOCUMENTATION REQUIREMENT-DR REQUIRES 50.59 PROCESS
- COMPLETED REVIEW OF POST-STARTUP RECORDS FOR SIMILAR OCCURRENCES
- 13 AFFECTED VALVES IDENTIFIED-ENGINEERING DISCREPANCY EVALUATION PROCESS USED TO EVALUATE
- INITIAL OPERABILITY SCREENING
- PRIORITIZED BASED ON SAFETY SIGNIFICANCE (PRA)
- EN6INEERIN6 EVALUATION BASED ON LIMITORQUE MAINTENANCE UPDATE 89-01-INITIAL SCREENING AND EVALUATION INDICATES NO SAFETY SIGNIFICANCE FOR LIMITER PLATE REMOVAL.
- COMPLETION OF PRE-STARTUP DOCUMENTATION REVIEWS BY NOVEMBER 15, 1991
MDV PROGRAM TORQUE SWITCH SETTING PRESENT PROGRAM FOR CONTROL OF TORQUE SWITCH SETTINGS
- DEFICIENCY REPORT REQUIRED IN ORDER TO EXCEED SPECIFIED SETTINGS-REQUIRES DOCUMENTED ENGINEERING EVALUATION-ENSURES 50.59 IS ADDRESSED 91EC1-31
MOV PROGRAM TORQUE SWITCH SETTING SUMMARY
- OUR PAST EVALUATIONS WERE NOT FORMALLY DOCUMENTED
- NO SAFETY SIGNIFICANCE
- RECORDS ARE BEING REVIEWED FOR SIMI~AR OCCURRENCES AT BOTH STATIONS *
-ENGINEER DISCREPANCY EVALUATION PROCESS WILL BE INITIATED FOR SIMILAR OCCURRENCES IDENTIFIED
- REVISED PROCEDURES REQUIRE DOCUMENTATION OF FUTURE ADJUSTMENTS 91EC1-32
9iEC1.-4 MDV PROGRAM VENDOR INFORMATION_CONTROL
- NRC FINDING
- FAILURE TO REVIEW, EVALUATE, INCORPORATE AND MAINTAIN CERTAIN VENDOR TECHNICAL INFORMATION
- DESCRIPTION OF DEFICIENCY
..
-
'
.
-DURING INSPECTION PSE&G WAS UNABLE TO RETRIEVE THREE REQUESTED DOCUMENTS:
.ALIMITORGUE MAINTENANCE UPDATE 89-1
- LIMITORGUE MAINTENANCE UPDATE 90~1
.AMOVATS ENGINEERING REPORT * PSE&G RESPONSE-LIMITORGUE MAINTENANCE UPDATES SHOULD HAVE BEEN RETRIEVABLE-LACK OF RETRIEVABILITY OF MOVATS ENGINEERING REPORT DOES NOT SUPPORT FINDING
~
PSE&G VENDOR INFORMATION CONTROL PROGRAM VENDORS,,
,,
,,
ENGINEERING NUCLEAR PQA VENDOR R&A
~
DEPARTMENT PERSONNEL CONTACTS PURCHASING
+
a LOG RE GUEST EVALUATE MISSING INCORPORATE
, DOCUMENTS
+
..
t
,,
FILE DOCUMENT
-
COMPARE
--
MAINTENANCE
-
-
. LIST 91EC1-36
_\\;
MOV PROGRA~
VENDOR INFORMATION_CONTROL EVALUATION OF DEFICIENCY - MOVATS DOCUMENTS 91EC1-5
- MOVATS ENGINEERING REPORT MOVATS ISSUED REV. 0 REPORT INTERNALLY ON 1/3/91-PSE&G BECAME AWARE DURING TEAM INSPECTION ON 7/16/91
- REVIEW OF MOVATS PROCESS A ENGINEERING REPORTS NOT ISSUED DIRECTLY TO INDUSTRY
- VENDOR INFORMATION ISSUED UNDER TECHNICAL_
NOTICES-MOVATS ENGINEERING REPORT 5.0 WAS NOT INTENDED FOR INDUSTRY USE
- ABSENCE OF ENGINEERING REPORT NOT INDICATIVE OF PROGRAM DEFICIENCY
- MOVATS TECHNICAL NOTICES-MOVATS ISSUED 10 TECHNICAL NOTICES BETWEEN
- 1988 AND 1990 NOT RETRIEVABLE BY PSE&G ARECEIVED BY INDIVIDUALS AT PSE&G AFAILED TO FOLLOW PROCEDURE TO INPROCESS INTO PSE&G SYSTEM
MOV PROGRAM VENDOR INFORMATION CONTROL EVALUATION OF DEFICIENCY - LIMITORQUE DOCUMENTS 91EC1-6
- LIMITORQUE MAINTENANCE UPDATES 89-1, 90-1-LIMITORQUE MAINTENANCE UPDATE 89-1 ISSUED 12/89
.A. RECEIVED BY INDIVIDUAL AT PSE&G (1/90)
& FAILED TO FOLLOW PROCEDURE TO INPROCESS INTO PSE&G SYSTEM-LIMITORQUE MAINTENANCE UPDATE 90-1 ISSUED 5/90
& LIMITORQUE UNABLE TO PRODUCE MAILING. LIST
.A.NO PSE&G PERSONNEL ACKNOWLEDGED RECEIPT-MUG MEETING PARTICIPATION 7/90
&PSE&G PERSONNEL REQUESTED AND RECEIVED ALL LIMITORQUE MAINTENANCE UPDATES
&FAILED TO FOLLOW PROCEDURE TO INPROCESS INTO PSE&G SYSTEM
- LIMITORQUE MAINTENANCE UPDATE 88-1
- ISSUED BY LIMITORQUE 8/88 BUT NOT RETRIEVABLE BY PSE&G
&TRANSMITTED TO INDIVIDUALS AT PSE&G
&NO PSE&G PERSONNEL ACKNOWLEDGED RECEIPT
MOV PROGRAM VENDOR INFORMATION _CONTROL SAFETY SIGNIFICANCE
- MOVATS TNs 88-01 THROUGH 04, 89-01 THROUGH 04 AND 90-0i (TEN TOTAL INCLUDING SUPPLEMENTS)
-ENGINEERING REVIEW INDICATES NO SAFETY SIGNIFICANCE ISSUES-APPLICABLE CONTENT TO BE INCORPORATED INTO THE MOV PROGRAM
- LIMITORGUE MUs 88-1. 90-1-ENGINEERING REVIEW INDICATES NO SAFETY SIGNIFICANCE ISSUES-APPLICABLE CONTENT TO BE INCORPORATED INTO THE MDV PROGRAM
- LIMITORGUE MU 89-1 91EC1-35-EVALUATION OF 13 HO~E CREEK VALVES FOR WHICH LIMITER PLATES WERE REMOVED INDICATED NO SAFETY SIGNIFICANCE
J
MOV PROGRAM VENDOR INFORMATION _CONTROL ROOT CAUSE ANALYSIS
. *LACK OF UNDERSTANDING BY LIMITORQUE AND MOVATS OF PSE&G POINTS OF CONTACT FOR CORRESPONDENCE
- INADEQUATE TRAINING-LESS* THAN ADEQUATE UNDERSTANDING BY SOME-PSE&G PERSONNEL OF ELEMENTS OF VF.NOOR DOCUMENT CONTROL PROGRAM 91EC1-7
.)
-
-
MDV PROGRAM VENDOR INFORMATION_ CONTROL RECENT PROCESS IMPROVEMENTS
- ISSUED NUCLEAR DEPARTMENT ADMINISTRATIVE PROCEDURE TO CLARIFY THE OPERATING-EXPERIENCE FEEDBACK PROGRAM (3/90)
- PUBLISHED POLICY TO CLARIFY REQUIREMENTS FOR INPROCESSING OF VENDOR TECHNICAL DOCUMENTS RECEIVED BY PSE&G INDIVIDUALS (5/90)
. *ESTABLISHED PROGRAMMATIC STANDARD TO FORMALIZE VENDOR CONTACT PROGRAM (12/90)
91EC~-B
- UPDATED IMPLEMENTING PROCEDURE TO ALIGN TRANSMITTAL ANO INPROCESSING OF VENDOR TECHNICAL DOCUMENTS WITH PSE&G POLICY (7/91)
- ISSUED LETTERS TO VENDORS (3/91 - 7 /91) TO:
-REITERATE PSE&G INPROCESSING REQUIREMENTS-ESTABLISH AND MAINTAIN DOCUMENT TRANSMITTAL LISTINGS
)'
j
MOV PROGRAM VENDOR INFORMATION CONTROL
, CORRECTIVE ACTIONS
- INPROCESSED DOCUMENTS IDENTIFIED AS MISSING INTO VENDOR TECHNICAL DOCUMENT PROGRAM (B/16/91)
- HAD LIMITORGUE AND MOVATS MODIFY THEIR MAILING
LISTS TO CONFORM TO VENDOR CONTACT PROGRAM
91EC1-9 (8/23/91)
- COMPLETED SAMPLING OF OTHER VENDORS TO ASSURE RECEIPT OF ISSUED DOCUMENTS (9/5/91)
- ISSUED LETTER UNDER VICE PRESIDENT SIGNATURE TO ALL PSE&G PERSONNEL REITERATING IMPORTANCE OF AND THEIR RESPONSIBILITY TO VENDOR DOCUMENT CONTROL POLICY, PROGRAM AND PROCEDURES (9/6/91)
MDV PROGRAM VENDOR INFORMATION CONTROL.
CORRECTIVE ACTIONS (cont)
- WILL REVIEW VENDOR DOCUMENT CONTROL PROCESS TO DETERMINE POSSIBLE ADDITIONAL IMPROVEMENTS (11/29/91)
- WILL FORMALIZE PROCESS FOR VALIDATION AND RECONCILIATION OF LISTS FROM VENDOR CONTACT PROGRAM (11/29/91)
- PERFORMING COMPARISON OF PSE&G PROGRAM TO THOSE OF UTILITIES RECOGNIZED BY INPO FOR THEIR VENDOR INFORMATION PROGRAMS (11/29/91)
- EVALUATE ADDITIONAL TRAINING NEEDS AND IMPLEMENT ADDITIONAL TRAINING AS REQUIRED
- WILL PROVIDE OVERSIGHT OF THE VENDOR CONTACT PROGRAM THROUGH USE OF GA VENDOR AUDITS/SURVEILLANCES
- ON PERIODIC BASIS WILL PERFORM AN EFFECTIVENESS REVIEW OF VENDOR TECHNICAL DOCUMENT PROGRAM IN ADDITION TO NORMAL GA AUDITS OF THE PROGRAM
.
~.
MOV PROGRAM VENDOR INFORMATION CONTROL SUMMAR *PSE&G UNABLE TO RETRIEVE DOCUMENTS THAT SHOULD HAVE BEEN RETRIEVABLE
- PERFORMED ASSESSMENT OF THIS SITUATION
- MADE RECENT PROCESS IMPROVEMENTSi TOOK IMMEDIATE CORRECTIVE ACTION At<<> Catl4ITTED TO LONG-TERM CORRECTIVE ACTIONS
- VENDOR INFORMATION CONTROL COtif>l.EX ISSUE
- RECOGNIZED NEED FOR CONTINUAL IMPROVEMENT
MDV PROGRAM PSE&G ASSESSMENT OF POTENTIAL VIOLATIONS GL 89-10, SUPPLEMENT 3 RESPONSE e PSE&G BELIEVES THAT A * 10CFR50. 9 VIOLATION IS *NOT APPLICABLE FOR THE FOLLOWING REASONS:
-A PSE&G REVIEW.OF SEVERAL EVALUATIONS WAS PERFORMED THAT PROVIDED A SOUND TECHNICAL BASIS FOR DATA IN SUPP 3-INSPECTION REPORT ISSUES IDENTIFIED WERE ADDRESSED BASED ON BEST TECHNICAL INFORMATION AVAILABLE
- DIFFERING PROFESSIONAL OPINIONS EXIST. ON SOME TECHNICAL ISSUES-FAILED TO ADEQUATELY COMMUNICATE THE LOGIC FOR OUR CONCLUSION OF OPERABILITY BOTH IN OUR SUPPLEMENT 3 RESPONSE AND DURING THE INSPECTION
- INFORMATION PROVIDED WAS NOT INACCURATE OR INCOMPLETE
- NO SAFETY SIGNIFICANCE -RELATED TO SUPP 3 ISSUE-ACTIONS UNDERWAY TO IMPROVE EFFECTIVENESS OF COMMUNICATIONS WITH THE NRC-PAST PERFORMANCE RELATIVE TO WRITTEN COMMUNICATION/RESPONSE TO NRC ISSUES HAS BEEN GOOD
- MOV PROGRAM
PSE&G ASSESSMENT OF POTENTIAL VIOLATION TORQUE SWITCH SETTING AND VENDOR DOCUMENT CONTROL *
ISSUES
- PSE&G DOES NOT DISPUTE THE VIOLATIONS
- SEVERAL MITIGATING FACTORS APPL~
-COMPREHENSIVE CORRECTIVE ACTIONS TAKEN/UNDERWAY-DEMONSTRATED NO SAFETY SIGNIFICANCE-CONTINUING FOCUS ON IMPROVING SUBJECT PROGRAMMATIC AREAS-PAST PERFORMANCE ON IDENTIFICATION AND CORRECTIVE ACTIONS OF DEFICIENCIES HAS BEEN GOOD
- PSE&G BELIEVES THAT ESCALATED ENFORCEMENT SHOULD NOT BE APPLIED TO THESE ISSUES 91EC1-13
Ge
~*
oe **
@. **
0*
~-(
- 0e MOV PROGRAM SU.MMARY
- NO SAFETY ISSUE IDENTIFIED WITH OPERABILITY OF THE MOVs
- PSE&G RECOGNIZES THAT MORE MANAGEMENT ATTENTION/OVERSIGHT IS NEEDED FOR MOV PROGRAM
- FULLY PLANNED PROJECT IN PLACE-PROGRAM DESCRIPTION BY 10/31/91-PROGRAM COMPLETION BY 6/94 (ORIGINAL.
COMMITMENT)
- FAILED TO ADEQUATELY COMMUNICATE THE LOGIC FOR OUR CONCLUSION OF OPERABILITY IN OUR SUPPLEMEN-3 RESPONSE
- DATA PROVIDED WAS SUPPORTED BY REVIEW OF SEVERAL EVALUATIONS/BEST TECHNICAL DATA AVAILABLE
- ROOT CAUSES DETERMINED FOR TORQUE SWITCH/VENDO DOCUMENT ISSUES
- COMPREHENSIVE CORRECTIVE ACTION TAKEN/CONTINUING
- ESCALATED ENFORCEMENT NOT WARRANTED 91EC1-33
'-
Ps~G Public Service
~
Electric and Gas
.
Campany ATTACHMENT 3 NRC
- ENFORCEMENT
'
'
'
- CONFERENCE
MDV PROGRAM SEPTEMBER 9, 1991 HOPE CREEK 8BERATIN8 STATION
- I llRC BlfPORCBllDr.l' COllPBRBllCB SBP.tBMBR'R 9, 1991 PRESBllTATIOll ROTES Introduction. * * * * * * * * * * *. * * * * * * * * * * * * * * * * * * * * * * * * * *
i overall Safety A&sesa.ent **************************
Pr'ograJI l>escription. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
Proqra. Description Deviation **** e*****************
Suppla.e.n.t 3 Respons.e ***************** _ ****.******** ~
Torque SVitch Setti.IlC). * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
Vendor Information control ************************* 11 III. SUPPLBllEHTARY DATA -
IHSPBCTIOll REPORT 354/91-80
"
ti_
- . ;i.i~*-*..... 9711.h
'ftlia-rDllaving provid-a written diacuaaion of topica diacumed at the*sept-ber 9, 1991 Bnforcwnt Conferenc In addition, iaauea covered in the inspection report, but not addressed at the enrorccment conference are discussed *
(
OVJWALJ* SAlETX A8SISSKENT A detailed review of our Supplement 3 analysi* ha* verified that the six identified valves will close under daaiqn baaia accident condition Thi* concluaion i* baaed on the fact that substantial margin*
exist.in all ca*** due to torque *witch bypaa* circuitr The bypass circuitry aaaur** that adequate thrust i* available to overcome potential valve factor concern* and to achieve disk-to-seat overla The atructural capability of the valve asaamblies are adequate to witbatand th* atreaaea associated with torque switch bypaa* circuitr Subaaquent evaluations, as well as a review of data available prio~ to our Suppl.. ent 3 response, have confirmed our operability conclusion&.
MOY PROGRAM SCHEDULE Phase I of our Generic Letter 89-10 proqram was originally scheduled for implementation between December 1989 and October 1990 while Phase II was scheduled to begin with the completion of Phase I in October 1990 and to continue through June 199 Due to unplanned burdens on our valve engineering resources, which were not adequately monitored and managed, Phase I was not completed until June 199 Althouqh the initiation of our Phase II was delayed until June 199i, PSE'G management is committed to take whatever actions are necessary to complete Phase II in accordance with the original schedule (June 1994) *
The schedule for completion of our proqram description has slipped ten months from January 1, 1991 to October 31, 199 PRQGRAM DESCRIPTION Phase I Activities Phase I of our Generic Letter 89-10 proqram included four major activitie The first activity involved development of engineering require.ants and proqraa implementation plan The second activity consisted of initiating a pilot proqram by performing a detailed analysis for 15 selected valves (9 Salem valves and 6 Hope Creek valves).
As a result of the Phase I delay, we were aware of concerns with the six valves which were addressed by Generic Letter 89-10, Supplement 3 prior t performing th* detailed analysi The six Supplement 3 valves were therefore the six Hope Creek valves evaluated during Phase The third activity involved assessing our existing MOV proqram against the Generic Letter 89-10 recommendation This activity was intended to identify any necessary improvements to our existing proqra The final major activity consisted of identifying scope items for inclusion in Phase II of the program.
Le*aons Learned - Phase I A9 a result of Lessons Learned durinq Phase I, nuaerous *
illprov.-enta were identified for incluaion in Pbaae II. The principal 1-proveaenta included enhancement of the current nuclear department policy governinq the overall HOV program, develop*ent of programmatic.standards for control of future HOV activities, consolidation and reconciliation of existing data
- sources, and development of improved methods and procedures for HOV testing and maintenanc Phase II Activities Phase II includes two major activitie The firat activity consists of implementing policy and forJaal proqraa requirements to meet Generic Letter 89-10 objectivea. Thi* activity includes developinq prograJ1 implementing procedures which cover aanaqement policy, data collection/reconciliation, functional evaluations, corrective action, and trendin Development of improved maintenance methods for maintenance and testing are also contained within this activit The second activity involves establishing detailed schedules for proqram development and implementatio our programmatic standard alonq with its twelve appendices constitute our MOV ProqraJ1 description and will be completed and approved by October 31, 199 A project plan, which outlines organizational and functional responsibilities relative to our MOV program, ha* also been prepare In addition to the development and processing of the programmatic standard, its appendices, and our p*roject plan, current onqoinq activities, include data gathering and MOV evaluations to support.aintenance activitie Phase II is scheduled to be completed by June 1994 *
...
..
PROGRAM DESCBIP'l'ION Dl\\1llTIOB ISSUE
"In a sublli~tal dated August 31, 1990, to ~e NRC, the licensee stated that a detailed GL 89-10 proqra. de*cription will be available onsit* on January 1, 199 However, at the ti.. of the inspection, the licensee had not established an approved proqram description."
PSE&G BESPONSE We agree with the deviation aa state ROOT CAUSE One root cause of the deviation a was failure to recoqnize and address the over-commitments on our valve engineering resources in a timely manne The over-commitments resulted in a ten month slip in schedule primarily due to turnover of key personnel and diversion of resources to other significant issue*.
Other
.
significant issues included ECCS and MSIV concerns at Salea which resulted in self-imposed shutdowns during the Spring, SWIJller, and Fall of 199 A second root cause was failure to utilize our monitoring and control process for tracking Generic Letter 89-10 program description commitment Had this commitment been properly tracked, an alternative means of alerting management to the approaching commitment would have existe COBREC'fIVE AC'l'IONS The Vice President - Nuclear Engineering has counseled all management personnel directly involve A letter has been sent from the Vice.President - Nuclear Engineering to all Nuclear Engineering Department personnel communlcatin9 his expectations relative to control of commitments, schedule adherence, and timely reconciliation of resource iaaues and requesting managers to review all regulatory programs to ensure similar problems do not exist in other area An independent review will be conducted to asaesa the effectiveness of our regulatory commitment tracking proces This review will be completed by December 31, 199 The project plan*was re-established in october 199 Phase I activities are now complete, Phase II activities are underway, and final completion is expected by June 1994 *
suPPT.EMJ!NT 3 RE5PQHSE PSEiG S"'f!Q*X BESPQNSI TO lOCFB.50.9 CONCER Although the technical basi* for our conclusiona was not adequately communicated, our response to suppleJ1ent 3 was co-.pleta and accurate based on engineering judg... nt and evaluatio Although thrust value* ware taken fro* valve data sheets, these values were considered to be the be*t available deaign informatio our operability conclu*ion waa based on margins available considering torque *witch bypaa* circuitr All available data and evaluation* were reconciled prior to our supplement 3 reapon***
Although the baai* for our operability conclusion was not clearly co.. unicated, an adequate and documented basis existed prior to our Suppleaent 3 respons We do not believe that our Supplement 3 responae waa either inaccurate or incomplete, and therefore, a lOCFRS0.9 violation is not justifie *
Each of the individual findings relative to the potential 10CFR50.9 violation is discussed belo The firat two findings have significance relative to lOCFRS0.9, however, although we consider the other three inspection findings to be important issues, they do not relate to any issue specifically addressed.in supplement Although they are technical issue* targeted for resolution under GL 89-10~ the*e issues were not required to be resolved prior to our Supplement 3 respons At the ti*e of our Supplement 3 response, we were not in a position to accelerate our proqram to address the issues covered under the final three finding Since acceleration of our program for the subject valves was a suggestion and not a requirement, we chose to make use of the best information available at the tim The issues described in the final three findings will be addressed in.Phase II of our progra SPECIFIC FINDINGS FINDING
"Thrust valu.. provided were not derived from detailed evaluatio.., but were obtained from valve data sheet Some of this data vam provided by the valve manufacturer (Anchor Darling)
as part of the.original plant design in the later 1970's."
"The licensee had three sets of evaluations (dated December.,
1990; February, 1991; and March, 1991;) performed by two licensee contractors indicating that the thrust requirements provided in the response might be nonconservative.* However, *the existence of this data was not mentioned in the licensee's March 8, 1991 letter to the NRC."
SUPfT,OONT 3 ftESPONSI PSE&G PQSl'.TXOH We agree that thrust values provided in our Suppleaent 3 response ware obtained froa valve data sheet OUr operability conclusion wa* based on our evaluation and reconciliation of all available information and our determination that po*itive aargin existed for all valves when -credit waa taken for torque switch bypass circuitr The valve data *heat value* repreaented the bast available desiqn information and were, therefore, the thrust values provided in our Supplement 3 respons Prior to submittal of our Supplement 3 response, the following reviews were performed by PSE, The valve data *beet thrust values were reviewed and determined to be conservatively derive The Phase I results which consisted of vendor evaluations were reviewed and found to be conservativ All available data-and evaluations, including the three sets of evaluation* noted in the NRC finding, were reconciled, and the values fro* the valve data sheets were determined to be the best available desiqn informatio The reconciliation and review of available information was documented in the draft engineering response to Supplement The draft engineering response demonstrated that, with credit taken for torque switch bypass circuitry, substantial margins exist for all of the subject valves irrespective of which of the various required thrust values were use The torque switch bypass circuitry allows thrust up to motor rated torque which greatly exceeds torque switch trip thrus The draft engineering response was overlooked and not provided to NRC inspectors during the inspectio Based on the above, we conclude that, although the basis for our operability conclusion was not clearly communicated, an adequate and documented basis existed prior to our Supplement 3 response to conclude that the valves were capable of performing their intended safety function under design basis accident condition Information we considered material to drawing this conclusion was contained in our Supplement 3 response *
-
suPPIPDT 3 usPQBsz FINDING
"The licensee did not have any technical justification for their use of a non-conservative 0.3 di*k factor and 0.15 ste11 friction factor in datenaining their required and available thrust capabilities."
- PSE&G PQSITION cµrrent disk and stea friction factor* are the result of best available vendor reco11J1endation Differing professional opinions exist within the induatry relative to apecific values to be used for disk factor and *t-friction facto OUr Phase II program will re-evaluate valve factors based on industry consensu We have historically included a aarqin of 15.t +/- 5% in our MOV maintenance procedures to address engineering
uncertainties and instrument inaccuracie Torque switch bypass allows application of full rate motor torque which allows accommodation of higher valve factor FINDING
"The thrust margins shown in the response did not include the effects of diagnoatic instrument inaccuracies or the rate of loading effects."
PSE&G POSITION Reported margins were not adjusted for instrument inaccuracie A margin of 15% +/- st has been historically included in our MOV maintenance procedures to address engineering uncertainties and instrument inaccuracie The margins reported in our response are in fact the allowance for engineering uncertaintie The actual margins would be greater based on the higher thrust which could be developed when the torque switcn is bypasse Additionally, rate of loading is considered to be a concern only when there is an interface between the torque switch and spring pac As auch, rate of loading is not a concern over the majority of the valve stroke due to the bypass circuit. Final disposition of th*** issues will be accomplished during Phase II of our MOV proqrma *
SUPPTPJiHT 3 *usPQNSE FINDING
"Th* liceJW** obtained the thruat valuea fro* teat who**
condition* were siqnificantly different fro* tho** durinq desiqn baaia accident conditions. Specifically, th* licensee provided th* thrust values when the torque switches tripped durinq the tea However, during an accident, the torque *witches will be bypassed and the available thrust will be dependent entirely on motor thrust capability."
PSE&G PQSITION The reported thrust values reflected the most recent diaqnostic results obtained under static condition* and were based on torque switch tri The bypass circuitry allows thrust up to actor rated capabilit Motor rated torque exceeds torque switch trip thrust and therefore torque switch trip thrust values are conservative. It is noted that defeat of the bypass circuitry oecurs after disk-to-seat overlap is accomplishe Differential testing will be performed under Phase II of our proqram where practica *
ISSUE
"However, the licensee still has to perform detailed evaluations to determine the margin of safety available after considering such factors as instrument accuracies and rate of loading.*
PSE&G PQSITION Instrument accuracies and rate of loading were previously addresse Based on the information provided, reasonable assurance of operability existed due to bypass circuitry even when instrument uncertainty is considere Final disposition of these issues will be accomplished during Phase II *
SVPPI:fi'EHT 3 RISPQNSE
ACTIONS TAD! TO DATI AND ACTIONS TO BE COMPLETED
We are *ensitive to the significance of these issues and the implicatiOJW of our less than clear communicatio As such, we have taken actiona to confira our previous* conclusions and investiqated the circumatancea leading to the issuance of our respons Additionally, we have identified follow-up activities to avoid sblilar occurrence* in th* futur Identified actions are as follows:
Actions Taken To Date We have re-evaluated existinq data to confirm our oriqinal operability conclusion In addition an independent
.
assessment has been conducted by our Nuclear Safety Review Grou The Vice President Nuclear Operations has discussed the issue with E&PB management and reiterated his expectations relative to communications with the NR Actions To Be completed A letter will be issued to all Nuclear Department Personnel to emphasize our obligation relative to clear and complete*
information in our communications with the NR Traininq*
will be conducted within the Licensinq Department to ensure that expectations and responsibilities are understood relative to communication with the NR our Supplement 3 response will be revised by September JO, 199 SQMMARY our response to Generic Letter 89-10, supplement 3 was accurate and complete based on engineering judgement and evaluation and therefore, we are not in violation of 10CFR5 Technical concerns.raised in Generic Letter 89-10 will be addressed under Phase II of our proqram. * Technical concerns *to be addressed include diagnostic equipment inaccuracies, rate of loading, and conservative valve factor Substantial J1&rqina exist to assure proper valve function under design ba*i* accident condition PSE&G is actively involved in industry qroups following MOV issues includinq MUG, BWROG, and EPR *
Although we failed to adequately communicate the technical basis for our conclusions, there was no intention to mislead the NR Our supplement 3 response will be resubmitted to clarify the logic for our operability determination *
TORQUB SWITCH SETTING Tb* findiDIJ, aa deacribed in th* inapection report, state* -
modification Qf MOV torque switch aettinga without a dOCU11ented engineering or safety evaluation. Specifically, the iillitar plates w*re r8JIOved fro* 2 RCIC valve aotor operator*.
Th***
plates were removed *o that the awitch aettinga could be increased to obtain th* thruat require No engineering evaluation had been documente PSE'G agrees with this findin our understanding of the iaaue, back during startup of Hope Creek, included the following elements:
The switch setting was intended to provide the design basis seating thrust for the valve; The placement of the liaiter plate was based on the vendor's calculation: this methodoloc;y (calculation and subsequent placement) was less accurate than direct measurement would yiel The switch settings were administratively controlled by the maintenance procedure: namely, the approval of the systell engineer was required, and verbal occurrence from Limitorque and the valve manufacturer was obtained prior to the removal of a liaiter plate *
However, we did not document the evaluation process or the telephoned concurrence We conclude that the root cause was that the past procedural controls failed to adequately address documentation requirement We have revised our maintenance procedures to require that a Deficiency Report (DR) be initiated for any future removal of limiter plate ORa require the 50.59 process, thereby ensuring the proper documentation of such evaluation We have done a review of post-startup maintenance records to identify similar occurrence To date, we have identified 13 valves with limiter plates remove These have been written up on DEFs (Diacrepancy Evaluation Forms) and are being evaluated under the Snqineering Discrepancy Evaluation proces This process *tart* with an initial operability screening, next, the DEF is prioritized based on its safety significance (PRA), and then evaluated in light of Limitorque Maintenance Update 89-our screeninq and evaluation shows*there is no safety significance for the removal of limiter plates on the 13 Hope Creek valves *
We will coaplete our review of all pre-startup documentation by Novmaber 15, 199 Again, our 1resent program for the control of torque switch settings ~ires the initiation of a deficiency report in order to exceed specified setting The deficiency report will document the engineering evaluation and ensure that 50.59 is addresse In summary:
our past evaluation were not formally documented; our review has shown no significant safety concerns:
The records at the Salea station are alao being reviewed for similar occurrences:
Any such discrepancies noted will also be evaluated using th~ DEF process: and We have revised our procedures to ensure that any future adjustments will be documented *
- VllfDOR INFORKATIOH CQNTBOL ISSQI HBC Finding Failure to review, evaluate, incorporate and maintain vendor technical information is a potential violation of lOCFRSO, Appendix B, Criterion Qescription of Qeficiency During Inspect~on No. 50-354/91-80, PSE'G waa unable to retrieve three (3) documents requested. by the inspection team:
Limitorque Maintenance Update (MU) 89-1,
"Maximwa Torque Switch Setting* and Other Issues" (Dated 10/89)
Limitorq\\le Maintenance Update (MU) 90-1,
"Hydraulic Lock and Torque Spring Assembly Relaxation" (Dated 5/90)
KOVATS Engineering Report (ER) 5.0,
"Equipment Accuracy Summary" (Dated 1/3/91)
PSE&G Position Limitorque MUs 89-1 and 90-1 should have been retrievable through our Vendor Document Control Programs at the time of the inspectio Lack of retrievability of KOVATS ER 5.0 does not support the NRC finding and thus is not a basis for a violatio PSE&G will elaborate on this position later in this document under
"Evaluation of Deficiency".
Vendor Information Control Program PSE&G is extr.. ely sensitive to the issues involved with Vendor Information Control. It is a complex issue requiring diligence on our part a* we coordinate with many (approximately 300)
vendors and process a multitude (approximately 7500) documents per yea The processes we have are in accordance with INPO Good Practice and have been embellished over the years to ensure the integrity of our programs *
The PSE'G Vendor Infonaation Control Proqr.. i* coapoaecl of four (4) pri.Jiary el... nt* each of which have aaaociated proc***
procedur.. addr***inq th* logginq, evaluation, incorporation and 11aintenance of certain Vendor Inforaatio These eleJMmt* were eatabliahed alonq the organizational lines of responsibility of our Enginaarinq, Procurement Quality Assurance, Reliability and Assessment; and Purchaaing Organization The four (4) primary element* are auppl-ented by two secondary elements in order to maximize the integrity of our progrm These secondary elements are our Vendor Contact Progrma; and the requirement to and provision for individual Nuclear Department personnel receiving Vendor Information to in-proca** same into our Vendor Information Control Progrm There has been continuing improvement/modifications made to these programs over the last few year Additional improvements that we will be makinq aa a result of the inspection findings are discussed in the following section Evaluation of Pef iciency MOVATS Engineering Report *
MOVATS issued ER 5.0, Revision o internally on 1/3/91 *
PSE&G first became aware of ER 5.0's existence during the inspection on 7/16/91 when requested to retrieve it by the inspection teal *
A subsequent review of the MOVATS process indicated that:
-
ERs such as ER 5.0 are internal MOVATS documents and, as such, are not directly_ transmitted to the industry (including PSE&G) *.
- Relevant information resulting from ERs would have been iaaued to the industry by MOVATS in the form of Technical Notice *
KOVATS ha* not yet incorporated the results of ER into a Technical Notice (TN):
- The report was issued internally for the purpose of consolidating data and to be used as a training ai PSE&G is aware that ER 5.0 involves instrument accuracies associated with MOVATS equipment which is currently an unresolved industry issu PSE&G personnel actively participate in t~e Motor
- '
Operated Valve Uaar'* Group (MUG) and are cloaely follovinq thi* iaau* to reaolutio PSBiG i* in receipt of, and in the process of raviewinq, a letter fro* Weatinghouae ITI KOVATS dated August 16, 1991 pertaininq to this iaau *
Based on the preceding discussion, tha absence of ER 5.0 from our Vendor Document Control System is not indicative of a deficiency in our progra.
MOVATS Technical Notice* (TN*)
Further evaluation of documents issued by MOVATS indicated that tha following ten (10) TH* had been distributed but were not retrievable through PSE'G's Document control Systeil:
TN 88-01 "Differential Pressure Thrust Calculations" TN 88-02 "Spring_ Pack Response to Stea Loads" TN 88-03 "Use of the AC Motor Load Unit" TN 88-04 "2151 Mainframe" TN 89-01 "Locked Rotor Condition Due to Grease Relief Kit" TN 89-02 "Spring Pack Response Under Differential Pressure" TN 89-03 "Use of AC Motor Load Unit" TN 89-03 Supplement 1 "Use of AC Motor Load Unit."
TN 89-04 "Output Imbalance in HBC Gearboxes and Use of the.BART System for Measuring Gearbox output Torque" TN 90-01
"Rate of Loading"
- The.. TN* were transaitted directly ta individual PSB'G personnel who had participated in the KOVATS trainingpraqraa a* evidenced by the KOVATS
.. !ling lis Hane at the acknowledged PSB'G-recipients of the subject TNs.had atteapted ta in~praceas the.a into our Document Control* syatea inaatar as they were perceived to be peraonal copie.
Limitorque Maintenance Update& 89-1 and* 90-1
On 12/22/89, Limitorque issued MU 89-MU 89-1 was directly transmitted ta individual PSE'G personnel that participate in the MUG aa evidenced by the Limitorque Mailing Lis *
In January 1990, MU 89-1 waa received by a Maintenance Engineer at Hope Creek for information under a Limitorque cover letter dated 12/22/8 *
- No further action was.taken insofar aa it was perceived as a personal copy and assumed to be in our syste In May 1990, Limitorque apparently transmitted MU 90-1
- Limitorque can not provide evidence (i.e.,
Mailing List) that it was transmitted directly to PSE, No PSE'G personnel known to be on Limitorque*s recent mailing list acknowledge receipt of MU 90-.
In July 1990, the MUG held their summer naeeting
- Salem System Engineer obtained copies of all MUs issued to date directly from Limitorque after they were identified at the MUG meetin He wa* not aware of the procedural guidance with regard to the in-processing of vendor documents obtained in this manne.
Limitorque Maintenance Update 88-1 Further evaluation of documents issued by Limitorque indicated that one (1) additional Maintenance Update,,
MU 88-1, had been distributed but was not retrievable through PSE&G's Document Control Program *
JIU 88-1 had been tranmaitted directly to two (2) PSE&G perac>Jmel who had been involved in.valve maintenance
~iviti** at th* tiae of i*auance a* evidenced by Liaitorqu*'* :aailinq li* One PSB'G recipient left PSB'G'a employment shortly after MU 88-1 wa* i**ue The other individual, who i* no lonqer involved with valve maintenance, could not acknowledge receipt of MU 88-1.*
Safety Significance MOVATS Technical Notices
TN 88-01 through 88-04, 89-01 throuqh 89-04, and 90-01
- All ten (10) MOVATS Technical Notice* are undergoing the required Bn9ineerinq review in accordance with PSE'G'* proces A preliminary review of the content indicates that there are no safety siqnificant issues involve All applicable information that i* included in these Technical Notices will be factored into PSE&G's MOV progra Limitorque Maintenance Updates
MU 88-1 and MU 90-1
- Both Limitorque Maintenance Updates are undergoing the required Enqineerinq review in accordance with PSE'G'a proces A preliminary review of ~e content indicates that there are no safety siqniticant issues involve All applicable inf oraation that is included in these Maintenance Updates will be factored into PSEiG'* MOV proqral *
JIU 89-1
- Evaluation of 13 Hope Creek valves for which Limiter Platea were removed indicated no safety significance.
Suryey of Qther Vandgrs PSB*G con~ed a *ample (lOt) of our vendors and requested them to identity updated document* that they have provided to PSE*G in the last tvo (2) year*.
The vendors identified a total of 57 document* that had been transmitted during the 2 year perio PSE&G was able to retrieve lOOt of the document* through our Vendor Docwlent Control Syat...
PSE*G therefore concludes that: 1) this issue i* limited to MOVATS and Limitorque, and 2) reaultad from th* direct interface between the two vendors and the individual PSE*G personnel involved with MOV maintenance iaaue Root Cause Analysis
Lack of clear understanding by Limitorque and MOVATS of the PSE*G Vendor Document Control Proqra. and their responsibilities regarding points-of-contact for correspondenc *
Inadequate training was a major contributing factor as evidenced by the untcuailiarity on the part of some PSE&G personnel with "elements" of the Vendor Document Control proqra Specifically, personnel not normally in the "mainstream" of the Vendor Document Control Programs did not understand their responsibility regarding the in-processing of documents received directly fro*, vendor Recent Process Improyements PSE&G has, and will continue to recoqnize.that vendor information control is a very complex and extremely important issue which must be thoroughly and continuoualy addressed in order to operate our nuclear units safely and reliabl To thia end, PSE&G is committed to work with our vendor* and the industry to demonstrate continued illproveaent in this area to our customers, regulators, and ourselve *
On 3/7/90, PSE&G issued Procedure NC.NA-AP.ZZ-0054(Q)
Revision o, "Operating Experience Feedback (OEF)
Program", which clarified the Operating Experience Feedback Program.
-On 5/31/90, PSE'G iaaued Procedure VPH-EDP-01 Reviaion 1, *vendor OOCU.ent Control*, which provide* the policy for our Vendor Technical OOCU.ent (VTD) Proqra This raviaion included the addition of Hope creek Station applicability throughout the text, and clarified the requirement* for receiving, reviewing and di*tributing V'1'D on 12/31/90, Programmatic standard DE-PS.ZZ-003l(Q)
Revision o, *vendor Contact Prograa*, was iaaued which formalizacl our vendor contact prograa as required by GL 90-0 On 3/8/91 and 6/12/91, PSE*G issued-letters to vendors for the purpose of clarifying docu.ent transmittal requirements and PSB'G contact* (Engineering, Procurement Quality Assurance, Reliability and Assessment; and purchasing); and establishing a document tracking systa. with each vendo On 7/29/91, Procedure NC.DE-AP.ZZ-0006(Q) Revision 3,
- vendor Document Control Program*, was issue In part, this revision aliqned the responsibilities of Nuclear Department organizations with those outlined in procedure VPN-EDP-0 On 7/30/91, PSE&G issued follow-up letters requesting a response by 8/22/91 to all Suppliers With Approved Quality Systems (SWAQS) lists and GL 90-03 *category b" vendors who had not responded to our previous letters (this included Limitorque and KOVATS).
Corrective Actions
The following corrective actions have been completed:
- All document_s requested by tbe NRC during the the inspection which were not retrievable through our system have since been obtained, in-proceased to our Vendor Document Control System and are currently undergoing revie Limitorque and MOVATS, as directed by PSE&G, has aoditied their mailing list to conform with our Vendor Contact Proqra A sampling of other vendors to assure receipt of all documents issued to PSE&G has been complete *
- A letter ha* been iaaued, under Vice Pr**idential aignature, to all PSB'G Nuclear Departlaent Personnel reiterating the importance ot and their r.. ponaibility to the Vendor Dc>cwlent Control Policy, Proqru and Procedure *
Tbe following corrective actions will be completed by November 29, 1991:
- All Vendor Doculllent Control Proqrcua Procedures will be reviewed to ensure clarity and determine additional area* tor potential iJaprov-en As part of this review, PSB'G will addreas procedural consiatency and idsntif ication of proper responsibilitiea, authorities and interface The process for validating and reconciling vendor-supplied document distribution lists will be formalized; this will provide confirmation upon receipt of docwaents transmitted by vendor An evaluation will be conducted to compare PSE'G's programs to those of other utilities that have been recoqnized by INPO as having excellent Vendor Information Control Proqrams *
The following on-going corrective actions will be implemented after completion of the programmatic changes described above:
- oversight will be provided to the Vendor contact Program via periodic QA Audits and Surveillances of the vendor This will ensure that: 1) vendors participating in our program are maintaining mailing and document distribution lists as requested, and 2) vendors not participating in our progru are spot-checked for docwnents which have been transmitte Periodic effectiveness reviews of our Vendor Document Control Programs will be conducte * - Training needs will be evaluated and additional training programs implemented as required *
!\\
Snparv*
PSB*G va* \\lllabl* to acce** th* thr** (3) Lillitorqua Maintenance Update* a.Riii ten (10) llOVATS Technical Notice* which *hould bave been ratri.vabl* through our Vendor ~nt Control Sy*te Upon asaa*a*ant of thia *ituation, root, cau*** have b9en identif iad and inmediate corrective actions take To prevent recurrence and further strengthen our vendor information control process, PSB*G ha* co11Jaitted to the actions identified abov PSE'G reeoqnizes that vendor information control is a very complex and extraJaely important i*aue which auat be thoroughly and continuously addressed in order to operate our nuclear units safely and reliabl To this and, PSEl&G is coJlllllitted to work with our vendor* and the industry to demonstrate continued improvement in this area to our customers, requlators, and ourselves *
i *
. I
-*
SUPPIDQPl'AJlY I1'PORMATIOll - IHSPECTIOH REPORT 354/91-80 nt: (S.ction 2.0, Para. 2, p. 4)
- Tb* scope of the proqrua i* li.Jlited-*to only active function valv***
GL 89-10 and Suppl-ent 1 reccmaended that all MOVs in safety-related ayst-. be included in the MOV program scop Supplementary Data:
Th* scope of the program ia not lillited to only active function valves and encoapasaec a large population of valve* baaed on the-criteria outlined in Appendix Al of Programtic Standard HC.DE-PS.ZZ-0033 (Q).
The baaic criteria for including valves in the progrua is &Ullllarized as followa:
a)
MOV is required to perform an active aafety function; b)
Operability of the MOV is required by Technical Specifications; or, c)
MOV operation is required in the course of performinq Operating, Abnormal and Emergency Operating Procedure*.
We believe these criteria meet the intent of the Generic Letter recommendation. NRC ColllJllent: (Section 2.0, Para. 2, p. 4)
- The licensee intends to use the BWR owners Group recommendations for desiqn baais review However, the licensee has not performed a detailed review to determine the applicability of the owner's group recommendations to Hope Cree Supplementary Data:
The use of the applicable owners group methodologies, with regard to design ~i* flow and differential pressure calculations, is an inherea!ly conservative approach which we believe does not require a detailed revie The prograa will consider any of the unique da*iqn attributes of the systea and valve functions as part of the design basis determinations and will adjust the methodologies as appropriat * NRC Comment: (Section 2. o, * Para. 2, p. 4)
- Inaccuracies of the diaqnostic equipment have not been adequately addressed..
(
-Suppleaentary Data:
At the tU. of the inspection, a draft Programaatic Standard wa*
provided vbich ducribed th* MOV Proqr The illpleaentinq Appendic** vill *uppl.. ent thi* *tandard and will def in* total progrma illpleaantation. It va* always our intention to acldr***
th* inaccuracie* of diagno*tic equipaent through total prograa developmen PSEfrG has taken initial po*itive actions in.this.
are A conference was bald at Artificial I*land on Auguat 30, 1991 to begin the process of validating th* accuracy of th*
diagnostic equip.. nt in use at Artificial I*land. Attendees included representative* fro* th* equipment aanufacturer and also from two other regional utiliti** which utilize *imilar equipmen The purpose was to pool effort* and r**ourcea with the regional utility users and the manufacturer to provide an efficient and cost effective accuracy deter11ination progra In addition to this regional effort, PSE'G is alao participating in the diaqnostic equipment manufacturer'* conference on this topic scheduled for the 9th, 10th and 11th of SaptemberG Thi* topic is being actively addressed by PSEfrG due to the senaitivity of th*
isuue. NRC Comment: (Section 2.0, Para. 2, p. 4)
- Testing where "practicable" has not been clearly defined to preclude deviations tro* the intent of the generic lette Supplementary Data:
The MOV Proqrammatic Standard uses the term practical instead of practicabl The evolution of the MOV proqrma will more clearly/
define the use of the term "practical" and its impact on the
total population that can be safely tested at pressures at or near the maximum capabilit. NRC CoJIDlent: (Section 2.0, Para. 2, p. 4)
- Periodic Verification has not been adequately addresse supplementary Data:
Periodic re-verification of switch setting adequacy will be performed on a frequency not to exceed 5 years or 3 refueling outages as outlined in Section 4.4.4 of the HOV Programmatic standar This frequency will be evaluated and may be increase or decreased based upon MOV specific evaluations as data is compile The requirements for periodic setting verification following maintenance activities will also be clearly delineate
/\\
(Section 2.0, Para. 2, p. 4)
- Th* prograa doe* not add.re** aection h. of the generic lett:er relating to *failure**, *corrective actiona* and
- trm111inq*.
Suppl.. entary Data:
Section 4.6.1 of the MOV Programaatic Standard addressea Section h. of th* Generic Lette. NRC co... nt: (Section 2.0, Para. 2, p. 4)
- The proqraa does not contain sufficient details to demonstrate how th* reco.. endad achadul** of the generic letter will be impleaente supplementary Data:
A achedule will be developed once the MOV population has been selected and prioritized as outlined in Appendices Al and A2 of the MOV Proqrammatic Standar. NRC Comment (Para. 3.1, p. 4)
There waa a lack of co11JDunication between the Maintenance and Engineering departments. * For example, th* maintenance department adjusted the minimwa required MOV thrust values provided by engineering by +lot to +2o Neither maintenance nor engineering knew the source or exact reason for this adjustmen Also, neither organization could explain how diagnostic inaccuracies are addressed in the licensee'* MOV progra Supplementary Data:
An explanation of 10-20t target thrust window over the design minimum was provided to the inspectors during the inspectio This explanation waa provided by the Engineering Departmen The Maintenance Supervisor had properly deferred the question to the Engineering Depart.ent as being outside his area of expertis We take exception to the conclusion and its basi.
eooFurtheraore,_ th* licensee had not evaluated or incorporated MOVATS Engineering Report 5.0 "Equipment accuracy SWDlllary* which provides recommendations on how to account for *rate of loading" effect supplementary Data:
According to MOVATS personnel, Engineering Report 5.0 had never been disseminated to utilities.
. 22
The inapector vaa uncertain bow thi* docu.ant bad co-into the Region'* pm**-ioa *
~-::: -:_
- R.Bta of 1-..lincr' will be addr-*ed in our proqr-a* indicated in suppl*ara* 1, i.e., a* a consideration in th* u.e of teat r-ulta tor *iailar valve *
PSB'G ia aware of and i* actively involved with the induatry i the reaolution of instruaent inaccuracy iaauea.~
10. NRC Co.. ent: (Section 3.5, Para. 2, p. 6)
Procedure NC.NA-AP.ZZ-0050 (Q), *station Teating Proqr-* does not clearly define the required po*t.aintenanco taating (PMT)
- following maintenance activiti*** Al*o, PMT for actuator replaceaent and valve packinq adjustllanta did not require.
differential presaure or diaqnoatic t-tin Suppl..-ntary Data:
Attachment 4 (Page 4 of 6) of th* subject procedure indicates the appropriate Post Maintenance Teatinq and Operability reteat requireaent* for actuator replace11ent and valve packing adjustment Possibly the inspectors did not have all pages or*
the procedure in questio *
However, it is important to note that NC.NA-AP.ZZ-0050(Q) is an acblinistrative pr04:edure that i* used in the Planninq anc:l Implementation process for guidelines. It i* not a stand-alone documen In conjunction with th* IST proqr-and the Plant Technical Specifications, the Syatma Engineer, Maintenance supervisor, Planner, and Licensed Operator, diacusa and establish the retest requir8ll8nts on a case by case basi currently, based on our co11J1itted response to NRC Bulletin 85-03, and the interpretation of ASMB Section XI requirements for Pre and Post Maintenance LLRT's, we do not necessarily perfora diagnostic testing after a valve packing adjuat.en This statement considers the requirement to maintain packing gland torque within the quidelinea of our Chesterton Repack Proqra Based on tM anticipated result* of the proqram associated with qeneric l.tt.r 89-10, we do foresee a chanqe to the requirements for differential preaaure or diaqnostic testing of motor operated valve. NRC co.. ent: (Section 3.5, Para. 3, p.3) reqardinq the schedule for MOV overhau Supplementary Data:
In conjunction with the existing trending proqram to assess historical motor operated valve failures, reco11J1endation h. of Generic Letter 89-10 requires that an adequate trendinq proqram
b9 **tablimhed to analyze HOV failur.. and diagno*tic t.. t r.. ul ta to verity th* adequacy of our Maintenance Progru.. we ccmaittect tw tbe inspector. that reco-m.tion h. of GL 89-10 would b9 ***flll&t*ly addres*ed in our final approved veraion of th* progr-.tic atandard for 11Gtor operated valve progr In the dran projeCtplan for th* GL 89*10 JIOV Progru, th*
re*ponsibilitiea tor each Jlellber of th* project t... are delineate One of th* reaponaibilitiea li*ted for the Reliability and A8****aent Departaent i* to "*valuate th*
lonq-tera llilintenanc* raquir.aenta (rebuild) baaed on reliability centered maintenance (RCM) approach."
- It is anticipated that the reliability aMa..maent perforaad for HOV* will include an evaluation of vendor aaintenance requirement*, EQ proqru requireaent*, available condition monitoring and predictive aaint~ce technique*, and an interpretation of current induatry practice Thi* approach is expectecl to satisfy th* inspector*' concern of an adequate overhaul progr&Jl/*checlule. *
11. NRC Co111J1ent: (Section 3.5, Para.4, p. 6)
The inspectors noted a concern with th* general valve maintenance procedure HC.MD-GP.ZZ-002(Q) general in*pection criteria (paragraph 5.2).
They felt that this paragraph and accoapanying check list/data sheet did not provided adequate guidance to
~rfora :meaningful inspectio Supplementary Data:
PSE&G's position regarding this procedure as discussed with the inspectors was: This is a "general" valve :maintenance procedur.
The intent of thi* paragraph is to dOCW1ent a "general" overall visual inspection of a valve and operator for obvioua siC)n8 of degradation, et Exampl.. of degradation that would be typically looked for are~loo.e-bolting, external leakage, et Thi* paragraph.relies on training that a typical boiler repair :mechanic would have to allow him to be qualif iecl to repair valve The use of the words "general inspection criteria" in paragraph 5.2 was not meant to imply inspection to rigoroua and absolute standards, rather the sense of a "general inspection check". We intend to add detailed inspection criteria consistent with the 89-10 program development: *
12. llJIC Cc:nnent: (Section 3.5, Para. 4, p.6)
'!'be i.118pr I.**r.lKed a concern with th* preventive aaintenance procedure ~1'1.ZS-004) requir-nt to aaintain the lubricant in tba liai%, tliU.f:cb gear box at 90 - lOOt ful _suppl~Data: -----
Thi* 90t ainillua requir... nt had been put into th* procedure ba*ecl on a letter fro* Liaitorque Corporation to Bechtel, Inc. on April 11, 1984 in r**ponse to th* specific conc.rn about qreaae leve.
,
When the inspector* provided U8 with a copy of an LER fro* a PWR utility, reqardinq a failure of an HOV due to exc.a*ive lubricant, we took 1-ediate actio We contacted th* Maintenance Departaent of th* affected utility, and other utilitiea to di*cu** how to iapl... nt th* practice of not fillinq loo we have already initiated a procedure revi*ion to incorporate the lesaons learne. NRC Comment: (Section 5.0, Para. 2, p.9)
Th* inspector* noted that th* vendor dravinq for RRR valve*
HV-F015A and B.indicated a direction of flow which appeared backward* for a qlobe valve *
Supplementary Data:
A walkdown of valve lBCHV-FOlSA perfonaed on 7/20/91 verified that the installed poaition is such that flow is under the sea Anchor Darlinq waa contacted on 7/22/91 and thi* information (i.e. installed po*ition) was discuased and confirme A review of P'ID'* Fabrication Iaoaetrics and Inatallation Specif !cation* have identified that these ~alvea were installed-with flow under the *eat and th* Anchor Darlinq valve drawing was not revised to reflect this installed conditio Based on the Bechtel in8tallation dOCUJDent *Y*t.., thia wa* allowed since the hiqhest tt..-clocullent, in this caae in*tallation specification 10855-P20..-*-.pecified how these valves were to be installe We have i~**** a- *docuaent only" d-iqn change, deaiqn change No. 4HC-Oa:M._$'__,_ to reviae the valve drawing for 1BCHV-F015A'B to identify tb&t flow ia under the seat. *
We have also reviewed all "safety related" qlobe valves used for thrott*linq at Hope Creek for any *i*ilar drawinq discrepancies (based on a computer run of all *afety related qlobe valves from MMIS and review8d-P,ID's, Fabrication Isometrics, Valve Manufacturer drawings, etc.) *
- ~
~*r Valv* Kanufactur*r drawing* do not contain a
. tba-. explicitly ic:l9ntifi** if flow i* over or under the *-t. *
bave v*rifi-4 tbat th-* valv** ar* install41d properly allilf va plan an r*vi*ing th*** Valv* Kanufactur*r
- drawing* as put of duign cbang* *ao *. 4HC-0234. HRC comient: (S*ction 6.0, Para. 4, p. 9) *
- 'l'h* licensee acknowl9d9ed that th-* iaau** will be proi)9rly addressed in their r*vised r**pons* to Suppleaent 3 of th*
9*neric latter a* w*ll aa in th* d*v*lopaent and impl.. antation Of th* GL 89-10 MOV Prograa.*
Suppl.. entary Data:
our revised respon** to Suppl..ant 3 will clarify our ba*i* for concluding that th* valves will function und*r d**ign ba*i*
conditions and will address th* que*tion* contain9d in th* June 25, 1991 Request for Additional Infonaation *
~.*: '
I.
ATI'ACHMENT 4 HOPE CREEK POST ACCIDENT SAMPLING SYSTEM (PASS) SEQUENCE OF EVENTS
.
Date(s) - 1991 January-February Early March March 15 May 16 May 17 May 16-July 9 July 12 July 19 July 12-August 5 August 5 Third Refueling Outag Post outage testiri.g - PASS operabl Emergency drill - PASS had no detectable sample flow noted during system operatio Training personnel noted low sample flow during PASS operation Chemistry received written emergency drill observation that.
PASS had no detectable sample flow on March 15. Chemistr supervision checked PASS operation and concluded that it was operable.
Training continued on PASS - low sample flow was still observed. Training informally communicated these findings to chemistry supervisio Chemistry received training feedback form documenting PASS flow problems. Work Order (W0910712101) writte Incident Report (91-111) initiate PASS inoperable during troubleshooting on system. Licensee determined that pipe sealant material caused the blockag PASS repaired and declared operable.