IR 05000354/1999002
| ML20206U859 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/14/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206U850 | List: |
| References | |
| 50-354-99-02, 50-354-99-2, NUDOCS 9905260041 | |
| Download: ML20206U859 (20) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-354 License Nos:
NPF-57 Report No.
50-354/99-02 l
Licensee:
Public Service Electric and Gas Company Facility:
Hope Creek Nuclear Generating Station Location:
P.O. Box 236 Hancocks Bridge, New Jersey 08038 Dates:
March 8,1999 - April 18,1999 Inspectors:
M. G. Evans, Senior Resident inspector
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S. M. Pindale, Senior Resident inspector J. D. Orr, Resident inspector
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J. G. Schoppy, Jr., Senior Resident inspector j
i T. H. Fish, Operations Engineer Approved by:
Gienn W. Meyer, Chief, Projects Branch 3 Division of Reactor Projects
9905260041 990514 f
PDR ADOCK 05000354 G
PDR q
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EXECUTIVE SUMMARY Hope Creek Generating Station NRC Inspection Report 50-354/99-02 This integrated inspection included aspects of operations, engineering, maintenance, and plant support. The report covered a six-week period of resident inspection; in addition, it included inspections by regional inspectors regarding refueling activities and operator training.
Ooerations Operators performed well during numerous refuf outage activities and returning the plant to 100% power. An operating crew responded appropriately and conservatively to a service water system strainer malfunction and heavy grass accumulation on the service water system traveling screens.
Maintenance and system engineerin0 provided prompt support to repair the C service water pump strainer. Extensive corrective actions were completed such that the service water system
was not degraded from the marsh grass event. The system engineers initiated corrective actions to make the service water strainers more reliable. (Section 01.1)
PSE&G performed refueling activities safely and conservatively overall. Nonetheless, an incorrect fuel bundle was installed into the core due to procedure non-compliance, personnel error, and degraded fuel pool coordinate markings. PSE&G responded quickly and appropriately to address this problem. (Section 01.2)
PSE&G corrected NRC-identified problems regarding operator attendance during licensed operator requalification training segments and documentation prior to the completion of the program. (Section 08.1)
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Maintenance Due to mislocated thermocouples PSE&G contractors, in part through inattention to detail, did
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not correctly monitor reactor vessel flange temperature prior to and during head stud tensioning for about forty minutes. The problem was self-identified by a maintenance technician. Later in the evolution, refuel floor technicians did not promptly report another temperature monitoring problem to reactor operators in the control room. PSE&G initiated prompt corrective actions to ensure that flange temperature. (Section M1.1)
Overall, maintenance performance during refuel outage eight was good. PSE&G identified some human performance problems including instances of inattention to detail. The inspectors concluded that PSE&G remained focused on improving human performance. (Section M1.2)
Operators and maintenance technicians performed a well controlled emergency diesel generator (EDG) loss of power and loss of coolant accident surveillance. Operations management led a multi-disciplined root cause assessment team in the resolution of a degraded il
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condition affecting the EDG output breaker. Troubleshooting efforts demonstrated safety focus, excellent system and design basis knowledge, and thorough root cause assessment.
(Section M2.1)
Enaineerina Thorough in-vessel visual inspections identified circumferential cracking in non-pressure retaining portions of the source range monitor (SRM) and intermediate range monitor (IRM) dry tubes. PSE&G developed appropriate precautions to ensure that the dry tubes would not become dislodged prior to core reload and would remain operabie during the next operating cycle. PSE&G intended to reevaluate the condition of the SRM/lRM dry tubes in the subsequent refuel outage. (Section E1.1)
PSE&G staff corrected IST violations associated with a modification to the control area chilied water system chillers. (Section E8.1)
PSE&G staff corrected a violation on the documentation for the control area chilled water system chillers licensing basis. (Section E8.2)
Engineering continued to develop actions in response to an industry-generic issue related to setpoint drift of main steam safety relief valves. PSE&G previously implemented actions to i
prevent further setpoint problems, however, similar results occurred with the modified valves.
PSE&G intended to closely monitor valve performance and associated SRV maintenance practices. (Section E8.3)
Plant Suooort The inspectors observed several activities that were supported by radiation protection. In general, the support was in the spirit of ALARA. However, the inspectors identified three examples of operators not effectively communicating with radiation protection.
The inspectors determined that these activities had pot 3ntial radiological implications. The operations manager agreed with this assessment and determined that the operators should be reminded to notify radiation protection prior to causing any change in radiological conditions.
(Section R1.1)
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TABLE OF CONTENTS EXEC UTIVE SU M MARY..................................................... li TABLE OF CONTENTS...................................................... iv l. Operations................................................................ 1
Conduct of Operations............................................ 1 01.1 General Observations..................................... 1 01.2 Fuel Handlina Ooerations................................... 2 O2-Operational Status of Facilities and Equipment........................ 4 O2.1 H ousekeepina........................................... 4
Miscellaneous Operations issue..................................... 5 08.1 (Closed) Unresolved item 50-354/98-302-02.................... 5 l l. M ainten ance............................................................ 6 M1' ' Conduct of Maintenance......................................... 6 M1.1 Reactor Vessel Head Tensionina............................. 6 M1.2 Refuel Outaae 8 Maintenance General Observations.............
i M2 Maintenance and Material Condition of Facilities and Equipment...........
M2.1 Emeraency Diesel Generator Testina.......................... 8
111. E ngineering............................................................ 9
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E1 Conduct of Engineering........................................ 9 '
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E1.1. SRM/lRM Dry Tube Crackina............................. 9
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E8-Miscellaneous Engineering issues..................,............. 11
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E8.1 (Closed) Violation 50-354/98-05-01
..........................11 E8.2 (Closed) Violation 50-354/98-05-03.......................... 11
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- E8.3 (Ooen/ Closed) Licensee Event Report 50-354/99-003.............
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IV. Pla nt S uppo rt.......................................................... 13 R1 Radiological Protection and Chemistry (RP&C) Controls................
R1.1 General Observations.................................... 13
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V. M a nagement Meetings.................................................... 14 X1.
Exit Meeting S ummary........................................... 14 INSPECTION PROCEDURES USED........................................... 15 ITEMS OPENED AND CLOSED............................................... 15
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- LIST OF ACRONYMS USED................................................. 16 h
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Reoort Details
Summary of Plant Status i
At the beginning of this inspection, Hope Creek was progressing with refuel outage 8 activities and the reactor was defueled. The outage was completed and the unit was placed on line on March 31,1999. At the end of the inspection period, the unit was operating at full power.
1. Operations
Conduct of Operations 01.1 General Observations a.
Inspection Scope (71707. 61726)
The inspectors observed numerous control room and field activities associated with the refuel outage, plant startup and power operations. Significant evolutions observed included:
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LOP /LOCA integrated testing ;
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shutdown margin demonstration;
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reactor vessel cavity letdown;
heat up for reactor pressure vessel in service leak test;
power ascension activities (e.g. control rod testing and withdrawal);
reactor core isolation cooling system testing;
main turbine roll; a
follow up after marsh grass fouling at the service water intake structure;
b.
Observations and Findinas
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In general, operators performed well in coordinating with maintenance technicians and system engineers. A consistent approach to safe operations was observed in each evolution. Operators conducted thorough pre-job briefs and utilized effective three way communications. The inspectors observed the control room operators make excellent use of industry operating experiences during the cavity letdown and shutdown margin demonstration pre-job briefs. The inspectors concluded that operator performance during the shutdown margin demonstration had greatly improved over that during refuel outage 7.
On April 1,1999, the main control room received high differential pressure (d/p) alarms for the service water pump discharge strainers. The C and D service water pumps were operating. Operators responded promptly to the service water intake structure and observed heavy marsh grass accumulation on the service water pump intake traveling screens. About 3 minutes after the high d/p alarms, the C service water strainer motor tripped. The equipment operator was unable to reset the motor. Operators entered the abnormal procedure for service water system malfunction. The A service water pump
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2 was started, and the C service water pump was secured. The inspectors reviewed control room logs and interviewed operators. The inspectors judged that the operators responded appropriately and in accordance with the abnormal operating procedure.
Follow up actions by the operating crews included additional monitoring of traveling screen conditions and grass accumulation.
Maintenance and system engineering provided prompt support to repair on the C service water pump strainer. One of the two strainer screen elements was discovered " blown out" and damaged from the excessive grass accumulation and d/p that developed across the screen. PSE&G analyzed the plant indications and determined that the excessive grassing had contributed to the strainer motor failure, which then caused additional grass accumulation on the screen elements and led to their failure. The strainer motor and screen elements were replaced. The system engineer initiated a corrective action evaluation to consider routine replacement of service water strainer screen elements and the strainer motor. System engineers also performed inspections of the safety-related heat exchangers and verified that the marsh grass had not caused a degraded condition on the heat exchanger tube sheets.
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Conclusions Operators performed well during numerous refuel outage activities and returning the plant to 100% power. An operating crew responded appropriately and conservatively to a service water system strainer malfunction and heavy grass accumulation on the service water system traveling screens.
Maintenance and system engineering provided prompt support to repair the C service water pump strainer. Extensive corrective actions were completed such that the service water system was not degraded from the marsh grass event. The system engineers initiated corrective actions to make the service water strainers more reliable.
l 01.2 Fuel Handlina Ooerations
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Inspection Scope (60710)
The inspectors observed fuel handling operations including core offload and reload from the control room and from the refueling bridge. The inspectors also reviewed the details related to two PSE&G identified problems that had occurred during fuel movement activities.
b.
Observations and Findinos The inspectors observed several fuel moves from the reactor vessel to the spent fuel pool and from the spent fuel pool back into the reactor. The following general observations were noted throughout the reviewed activities.
Control room operators remained aware of and satisfactorily completed technical
specification requirements associated with refueling operations;
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Radiation protection conducted thorough radiological controls briefings and
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ensured exposure was maintained ALARA (As Low As is Reasonably Achievable);
The refueling senior reactor operator (SRO) demonstrated excellent control, 3-
point communications, and independent verification of core alterations; and The refueling SRO, bridge operator, and refuel bridge spotter effectively used
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self-checking, independent verification, end 3-point communications to control core alterations. Example: the bridge opers+or demonstrated self-checking to identify and take appropriate corrective actions for a double blade guide that remained grappled following an attempt to disengage the grapple.
Notwithstanding these positive observations, there were two performance problems that PSE&G-identified during the conduct of fuel handling operations. The first problem was a mis-oriented fuel bundle in the spent fuel pool. On March 15,1999, PSE&G identified that during core offload the fuel bundle in location AC-65 had been installed in a northeast orientation vice the northwest orientation specified by the approved core offload sheet (rotated 90'). In response, PSE&G initiated an Action Request 990315147 to complete an evaluation. PSE&G also briefed this mis-orientation with all fuel handlers, discussing the potential for improperly mis-orienting a fuel bundle in the reactor core.
A second more significant error was identified on March 16,1999. During the core reload PSE&G identified a fuel bundle installed in the reactor core incorrectly.
Specifically, on March 15, a fuel bundle was removed from the location adjacent to the intended location in the spent fuel pool and installed in the reactor core (wrong bundle, correct location). Upon discovery, PSE&G implemented several actions including 1)
immediately suspending core alterations pending further evaluation, 2) removing the involved individuals from further fuel handling operations pending assessment and correction of their performance weaknesses,3) cleaning fuel pool coordinate markers,4)
providing lessons leamed, and 5) performing a random sample inspection of various fuel pool and reactor cavity bundle locations. After the error was identified, reactor engineering also performed a shutdown margin analysis and determined that adequate shutdown margin existed in the off-normal core configuration. Reactor engineering later developed a move sheet to restore reactor and spent fuel pool configuration to normal.
The inspectors independently reviewed the details associated with this error, and reviewed PSE&G's evaluation and found it to be appropriate. PSE&G documented this adverse condition in Action Request 990316250 for detailed evaluation and correction.
There were three individuals involved with the fuel handling operations; the SRO observed and directly supervised the core alterations; a refuel bridge operator manipulated the bridge, and spotter validated reactor core and spent fuel pool locations during core alterations. The individuals (except for the SRO) were contract personnel from another nuclear facility. In their evaluation, PSE&G determined that the root causes for the improper fuel move was procedure non-compliance, personnel error (inattention i
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4 to detail), and less than ideal fuel pool coordinate markings. The inspectors agreed with this determination.
i Procedure HC.OP-SO.KE-0001(O), Refueling Platform and Fuel Grapple Operation, specifies that bridge personnel ensure the fuel grapple is at the correct core or fuel pool coordinates after the grapple has rested on the fuel bundle. However, PSE&G determined that the spotter identified the target location (fuel pool coordinates) while the grapple was three to four feet above the fuel bundle. Because of the parallax effect associated with vision through over twenty feet of water, the final position could not have
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been determined accurately then and an error resulted. Other performance weaknesses were apparent, including the failure to correct degraded fuel pool coordinate labels (some missing or difficult to read) along with the less than optimal lighting in that area of the fuel pool. Because of these deficiencies, refuel bridge personnel had been counting from known wall locations to identify target locations, a technique that did not ensure a positive location identification.
The inspectors determined that PSE&G completed timely and effective corrective actions for this fuel handling problem. The error had no adverse impact on shutdown margin in i
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the reactor core. PSE&G's random sample of fuel pool and reactor core bundle locations did not identify additional errors, nor did the final core configuration verification.
Refuel bridge personnel resumed fuel handling operations on March 17. These two errors constitute a non-repetitive, PSE&G identified and corrected violation of procedure
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HC.OP-SO.KE-0001(Q), which is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-354/99-02-01)
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Conclusions PSE&G performed refueling activities safely and conservatively overall. Nonetheless, an incorrect fuel bundle was installed into the core due to procedure non-compliance, personnel error, and degraded fuel pool coordinate markings. PSE&G responded quickly and appropriately to address this problem.
Operational Status of Facilities and Equipment O2.1 Housekeepina (71707)
The inspectors walked down several safety-related pipe chases, steam sensitive areas, the reactor building and the turbine building. The inspectors also observed PSE&G perform a closecut inspection of the suppression chamber. Each area had been appropriately cleaned up after refuel outage activities. The inspectors found that all temporary equipment was properly secured. During the suppression chamber closeout inspection, the inspectors found that system engineers, senior reactor operators, maintenance supervisors, and plant management performed a very detailed inspection
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of the area. PSE&G had completed a very good cleanup in the suppression chamber following extensive maintenance and testing activities. Overall, housekeeping was observed to be very good.
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Miscellaneous Operations issue
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08.1 (Closed) Unresolved item 50-354/98-302-02: Five Licensed Operators in Jeopardy of Not Completina the Reaualification Proaram a.
Inspection Scope (92901)
Inspectors reviewed PSE&G's response to NRC-identified problems in the Hope. Creek licensed operator requalification training (LORT) program.
b.
. Observations and Findinos During an NRC inspection of the LORT program in November and December 1998, the NRC identified that as of December 4,1998, five licensed operators had not completed all required segments of their requalification program. The 24 month cycle associated with the LORT program was to conclude on December 31,1998, and those five
' operators were in jeopardy of not successfully completing the program within the defined 24 month period required by 10 CFR55.59(a)(1).
PSE&G corrected the problems prior to the program's completion. PSE&G attributed the reason for this problem to be less than adequate training management oversight of records program implementation and resource control. PSE&G also determined that there was inadequate implementation of the new computerized system (SAP) database for training records. In response, PSE&G training staff initiated action for the operators and others not identified by the inspectors to make up all missed training classes before December 31,1998. Training management later reported that this initiative was completed before the requalification period ended. PSE&G also provided training to operations training instructors and training management personnel to review regulations goveming training records, intemal requirements and guidance for training records, and training record processing related to the new SAP computer system. During the week of February 8,1999, the inspector reviewed training records for the five operators identified during the initial inspection, and sampled records of an additional twenty five operators.
The review indicated that the operators had made up missed training prior to December 31,1998. This item is closed.
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Conclusions PSE&G corrected NRC-identified problems regarding operator attendance during
licensed operator requalification training segments and documentation prior to the completion of the program.
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11. Maintenance M1 Conduct of Maintenance M1.1 Reactor Vessel Head Tensionina a.
Insoection Scope (61726. 62707)
The inspectors reviewed the circumstances related to contractor maintenance personnel inadequately monitoring the reactor vessel flange temperature during head stud tensioning.
b.
Observations and Findinas On March 24,1999, PSE&G began reactor vessel head tensioning after completing core reload and cavity drain down evolutions. Contractors were performing the work.
Temporary thermocouples were installed by magnets on various locations including the head flange and some studs. Additional thermocouples were intended to be installed on the vessel flange. Technical specification 4.4.6.1.4.b required that the reactor vessel flange and head flange temperatures be verified to be at least 70*F within 30 minutes prior to and every 30 minutes thereafter during reactor vessel head tensioning.
Contractors began tensioning the head at 7:50 a.m. on March 24,1999. At 8:30 a.m.,
technicians not initially involved with the thermocouple installation verified the proper installation of the thermocouples. These technicians discovered that the thermocouples intended to monitor the vessel flange had been incorrectly applied to the head flange.
The thermocouples were promptly and properly relocated to the vessel flange in accordance with the original work order.
Later on the same day, a control room operator noticed that the temperature for the vessel flange had drifted about 9*F in one 30 minute interval. The control room operators contacted the refuel floor technicians. The technicians reported that the thermocouple had already been discovered dislodged and had been properly reapplied.
The control room operators stopped the reactor vessel head tensioning evolution to discuss the importance of temperature monitoring and good communication between the control room and refuel floor. Head tensioning was resumed after PSE&G was convinced that refuel floor technicians would properly communicate the status of flange temperature monitoring problems.
PSE&G initiated an action request, AR 990324150, to develop long term corrective actions for the problems on March 24,1999. PSE&G determined that the initial problem with mislocating the vessel flange thermocouples involved some inattention to detail on the part of the technician and mis-applying past experience at another boiling water reactor that did not have similar flange sizes. The head flange at Hope Creek was much thicker and the technician did not visually verify that the thermocouples were actually on the vessel flange. The technician's view was partially obstructed by reactor head tensioning equipmen.
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The inspectors agreed with PSE&G's conclusion that vessel flange temperature was above 70*F during the times temperature was not monitored on the vessel flange.
i PSE&G considered the vessel temperatures, head flange temperatures, and ambient temperature to draw the conclusion. The inspectors determined that PSE&G fortuitously satisfied the LCO action statement while surveillance requirements were not met. This l
non-repetitive,' PSE&G identified and corrected violation of technical specifications is being treated as a Non-Cited Violation, consistent with Section Vil.B.1 of the NRC Enforcement Policy. (NCV 50-354/99-02-02)
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Conclusions
Due to mislocated thermocouples PSE&G contractors, in p:-t through inattention to j
detail, did not correctly monitor reactor vessel flange temperaiJre prior to and during head stud tensioning for about forty minutes. The problem was self-identified by a maintenance technician. Later in the evolution, refuel floor technicians did not promptly report another temperature monitoring problem to reactor operators in the control room.
PSE&G initiated prompt corrective actions to ensure that flange temperature measurements would be accurate and any further problems would be promptly reported
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to the control room.
M1.2 Refuel Outaae 8 Maintenance General Observations
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Inspection Scope (62707)
The inspectors observed several safety-related maintenance activities performed during refuel outage 8. The inspectors also reviewed the circumstances surrounding some
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b.
Observations and Findinas The inspectors observed several maintenance activities including: LOP /LOCA integrated testing, A emergency diesel generator (EDG) electronic govemor troubleshooting, A EDG speed pickup sensor troubleshooting, reactor core isolation cooling (RCIC) system testing, reactor vessel moisture separator latching, and core reload. The activities involved participation by maintenance technicians, operators, and system engineers. in general, the inspectors observed good maintenance practices by all individuals.
Technicians and operators were knowledgeable of the job or evolution at hand, and system engineers were involved ensuring that activities were properly managed and problems resolved. Details of LOP /LOCA testing are discussed in section M2.1 of this report. Details of refuel floor activities are discussed in sections 01.2 and M1.1 of this report.
PSE&G identified some maintenance errors during the refuel outage, including performing inspection on the wrong offsite power source bus duct to the safety-related 4kV busses, wiring a local power range monitor (LPRM) string backwards, improperly applying a test current to a safety-related molded circuit breaker, a fuel assembly misorienting in the spent fuel pool, loading an incorrect fuel assembly in the reactor
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8 vessel, and improperly applying temporary thermocouples on the reactor vessel flange during head tensioning. Each example involved some element of inattention to detail.
PSE&G entered each of the problems into the corrective action program. The problems occurring on the refuel flocr are d;scussert in detailin sections 01.2 and M1.1 of this report. The other problems are violatiore af minor significance and are not subject to formal enforcement action.
The problem regarding the bus duct inspection also involved a potential for personnel injury. Operators deenergized and tagged out the one independent offsite power source bus duct to the safety-related 4kV busses. The tagout was applied correctly, but maintenance technicians visually inspected the other, still energized, offsite power cables. The power cables were insulated, providing protection, but the maintenance technicians were misled believing the cables were deenergized. PSE&G appropriately responded once the problem and potential personnel safety issues were identified. The problem was identified by PSE&G technicians when confusion arose prior to performing the bus duct inspection on the other power source at a later date. PSE&G initiated an extensive root cause investigation and developed immediate corrective actions to prevent reoccurrence. The inspectors determined that the corrective actions were appropriate.
In accordance with their corrective action program, PSE&G was addressing each problem individually and intended to determine if a collective evaluation would be needed for the human performance problems.
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Conclusions Overall, maintenance performance during refuel outage 8 was good. PSE&G identified some human performance problems including instances of inattention to detail. The inspectors concluded that PSE&G remained focused on improving human performance.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Emeroency Diesel Gjtngrator Testino a.
Insoection Scope (61726)
The inspectors observed portions of the C emergency diesel generator (EDG) 18-month
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Loss of Power / Loss of Coolant Accident (LOP /LOCA) test from the control room and the C core spray pump room, in addition, the inspectors evaluated PSE&G's corrective actions in response to a failure of the EDG output breaker to close during one test segment and observed the associated retest.
b.
Observations and Findinos '
On March 15,1999, operators conducted HC.OP-ST.KJ-0007, Integrated Emergency Diesel Generator 1CG400 TEST-18 Months. This surveillance verified the EDG
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automatic start feature, emergency bus load shedding, and emergency bus load sequencing during simulated loss of offsite power (LOP) and loss of coolant accident (LOCA) conditions. Operators conducted the test within the scheduled outage window and effectively used the surveillance procedure to control this complex and infrequently performed evolution. During Section 5.5 (simulated LOP to the alternate infeed power supply), the EDG outpu' Sreaker failed to close as designed. Operators documented the failure in action request 990315238, completed the surveillance test, and declared the C EDG inoperable.
Operations management established and led a multi-disciplined root cause assessment team to evaluate the condition. The team syst3maticahy and methodically considered a broad spectrum of possible failure mechanisms involving previous breaker testing, old work orders, procedure adequacy, human performance, recent design change packages, testing restoration, and component failures. The operations department received high quality support from the maintenance and engineering departments in this effort. Based on available indications and detailed circuitry analysis, the team eliminated implausible causes and developed a well-focused troubleshooting plan. Maintenance technicians and engineers implemented the plan, and based on "as-found" conditions and logic card bench testing, determined that a failed input on the " program select"
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portion of the output breaker control logic card caused the failure. The failed logic card
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was replaced and operators successfully completed Section 5.5 of the 18-month i
surve:llance on March 20.
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Conclusions Operators and maintenance technicianc performed a well controlled emergency diesel generator (EDG) loss of power and loss of coolant accident surveillance. Operations management led a multi-disciplined root cause assessment team in the resolution of a degraded condition affecting the EDG output breaker. Troubleshooting efforts j
demonstrated safety focus, excellent system and design basis knowledge, and thorough root cause assessment.
Ill. Enaineerina E1 Conduct of Engineering E1.1 SRM/lRM Dry Tube Crackina
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a.
Inspection Scope (37551)
The inspectors reviewed PSE&G's interim resolution to indications of cracking on a non-pressure retaining portion of the source range monitor (SRM)/ intermediate range monitor (IRM) dry tubes identified during the refuel outage.
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Observations and Findinas PSE&G performed in-vessel visual inspections (IWI) on top portions of the SRM/lRM dry tube assemblies. The inspections are performed every other refuel outage and were incorporated into the IWI program in response to an industry wide problem with dry tube cracking identified in 1984. The cracking has been observed in the non-pressure retaining portions of the SRM/lRM dry tubes. The problem was transmitted to the industry in General Electric Service Information Letter (SIL) No. 409. On March 4,1999, PSE&G identified cracking indications on all 12 of the SRM/lRM dry tubes in the location that was similarly described in GE SIL No. 409.
PSE&G initiated an action request, AR 990304178, to track resolution of this issue.
PSE&G also developed precautions during the full core reload to ensure continued operation of the SRM/lRM dry tubes. ' Specifically, after three of four fuel assemblies surrounding the dry tubes were set in place, PSE&G performed an additional visual examination to verify that no dislodging occurred during fuelloading. PSE&G also determined that dislodging while loading the fourth fuel assembly was unlikely, because
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the three installed fuel assemblies would support the dry tube laterally if it did become bumped. A similarjustification was made for continued power operations. The cracks were under a compressive force and lateral movement was not possible with the surrounding four fuel assemblies preventing movement.
A 10CFR50.59 safety evaluation was developed and presented to the station operations review committee (SORC) for approval prior to restart. The inspectors attended the SORC meeting and determined the SORC members questions to be thorough and probing. Experts were present and answered the questions with sufficient detail. The safety evaluation applied to only one operating cycle. PSE&G intended to inspect and reevaluate the condition of all SRM/lRM dry tubes during the next refuel outage scheduled in April 2000, c.
Conclusions Thorough in-vessel visual inspections identified circumferential cracking in non-pressure retaining portions of the source range monitor (SRM) and intermediate range monitor (IRM) dry tubes. PSE&G developed appropriate precautions to ensure that the dry tubes would not become dislodged prior to core reloed and would remain operable during the
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next operating cycle. PSE&G intended to reevaluate the condition of the SRM/lRM dry tubes in the subsequent refuel outage.
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E8-Miscellaneous Engineering issues E8.1 (Closed) Violation 50-354/98-05-01: Failure to Establish IST Reauirements: and
[ Closed) Violation 50-354/98-05-02: Failure to Maintain Safety Related Nitroaen Bottle Pressure Reaulators at the Proper Setooint a.
Insoection Scope (92903)
The inspectors reviewed PSE&G's actions in response to violations of in service testing (IST) requirements associated with a design change to the Hope Creek safety-related control area chilled wat6r rystem chillers.
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Observations and Findinas in the first violation, Hope Creek personnel failed to establish IST requirements and periodically perform ISTs for two check valves. PSE&G attributed this failure to -
personnel ctror. In response, station personnel satisfactorily tested the check valves, developed applicable IST procedures for the valves, revised the design change procedure to require a separate sign-off for the IST review, and reviewed this event with design engineering personnel. The inspector determined these actions were r">propriate
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and confirmed that PSE&G personnel had completed them.
in the second violation, NRC inspectors found that the backup pneumatic pressure regulators for the chiller condenser cooling water pressure control valves were set below I
minimum design requirements. Also, operational tests had not been performed to i
ensure that the regulators would remain properly set in accordance with design requirements. PSE&G attributed the cause of the violation to personnel error. In response, station personnel restored the pressure regulator settings to the required values, gave interim guidance to the operators on the operation of the pneumatic supply system, developed IST procedures to include periodic verification of the pressure regulator settings, and reviewed this event with design engineering personnel. The
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inspector determined the response was appropriate and verified that PSE&G personnel
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c.
. Conclusions PSE&G staff corrected IST violations associated with a modification to the control area chilled water system chillers.
E8.2 LQlesed) Violation 50-354/98-05-03: Ineffective Corrective Action for a Chanae of Desian
, Basis for the Safety Related Ch;11er Ooeration.
a.
Insoection Scope (92903)
The inspectors reviewed PSE&G's actions in response to a violation for failing to promptly correct a condition adverse to quality that affected the safety-related control
area chilled water system (CACWS) chillers.
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b.
Observations and Findinas On December 10,1997 PSE&G engineers determined that the minimum cooling water inlet temperature for the CACWS chillers should be changed to a more limiting value of 70 degrees F, from 55 degrees F. However, station personnel did not incorporate this change into plant documents until May 7,1998, when guidance was provided to operators specifying the new 70 degrees F minimum cooling temperature.
PSE&G attributed tne cause of th's violation to personnel error. In response, station personnel completed an evaluation to determine the correct minimum temperature for chiller operation and revised the applicable system operating procedure accordingly.
Also, lessons learned from this violation were communicated to the engineering staff.
The inspector confirmed station personnel properly implemented these actions.
c.
Conclusions PSE&G staff corrected a violation on the documentation for the control area chilled water system chillers licensing basis.
E8.3 (Open/ Closed) Licensee Event Report 50-354/99-003: As-Found Values for Safety Relief Valve Lift Setooints Exceeded Technical Specification Allowable Limits a.
Inspection Scope (37551. 90712)
The inspectors performed an onsite inspection and verified PSE&G's corrective actions described in a licensee event report (LER) for several safety relief valves (SRV) whose setpoints were found to have drifted outside allowable values.
b.
Observations and Findinas On February 28,1999, PSE&G received the results of SRV setpoint testing for the 14 SRVs following the completion of operating cycle 8. The as-found setpoints for six of the 14 SRVs were out of specification. The technical specification allowable limit was +/-
3%. The six as-found setpoints ranged from 3.2% to 5.6% beyond the valves' nominal lift setpoint. All six SRV setpoints had drifted high.
The SRVs used at Hope Creek are a two-stage Target Rock design.' Five of the 14 SRVs also function as s'itomatic depressurization system valves. Setpoint drift for SRVs has been an ongoing industry-wide issue and several industry-initiated corrective actions have been taken. The most recent of these corrective actions implemented at Hope Creek involved modifying the pilot discs for seven of the 14 SRVs w'A platinum ion implantation. This effort was an attempt to resolve SRV setpoint dre. due to corrosion bonding of the pilot disc to the pilot seat. Operating cycle 8 was the first cycle during which the implanted pilot discs were in service. Three of the six out of tolerance SRVs contained the implanted rr: terial. Further testing and visualinspection of these three SRVs indicated the absence of pilot disc adherence due to corrosion bondin.
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PSE&G conducted further evaluation of test and inspection data, and concluded that the drift for the three SRVs with platinum ion implanted discs appeared to have been caused by friction on the sliding surfaces resulting from less than adequate maintenance
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performed by the SRV vendor prior to installation for operating cycle 8.
i PSE&G has actively participated in the Boiling Water Reactor Owners Group (BWROG)
committee associated with SRV setpoint problems. PSE&G engineers informed the inspectors that PSE&G plans to continue supporting the BWROG in working toward a resolution of the setpoint drift issue. Platinum ion implantation was implemented on an additional six SRV pilot discs during the recently completed refueling outage. Currently, 13 of the 14 SRVs now have this modification. PSE&G stated that the results of the i
platinum ion implantation will be further evaluated after the current operating cycle to evaluate its effectiveness at reducing setpoint drift.
During the recently completed refueling outage 8, all SRVs that were installed were inspected, refurbished and satisfactorily retested at a test facility. PSE&G informed the SRV vendor that performed tne maintenance and testing on the SRVs of the concerns regarding prior maintenance activities. PSE&G plans to propose a technical audit of the vendor's maintenance practices for BWROG consideration.
PSE&G initiated an action request, AR 990303177, to document the out of specification SRV setpoints. The inspector reviewed the AR and found it to be comprehensive. The inspectors found that PSE&G was providing sufficient attention to this condition, including the near term and planned corrective actions.
,
c.
Conclusions Engineering continued to develop actions in response to an industry-generic issue related to setpoint drift of main steam safety relief valves. PSE&G previously implemented actions to prevent further setpoint problems, however, similar results occurred with the modified valves. PSE&G intended to closely monitor valve performance and associated SRV maintename practices.
IV. Piant Suonort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Observations (71750)
- The inspectors observed several activities that were supported by radiation protection.
In general, the support was in the spirit of ALARA. However, the inspectors identified three examples of operators not effectively communicating with radiation protection as
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follows:
Reactor water cleanup letdown was initiated to the main condenser for the first
time during the refuel outage, and radiation protection was not notifie.
The pre-job brief for the initial main turbine roll did not include any radiation
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protection participation.
I An equipment operator walked into potentially contaminated streams of water
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'
while responding to overflowing reactor feed pump equipment drains without first contacting radiation protection support or providing protection for his shoes.
I The inspectors determined that these activities had potential radiological implications.
The operations manager agreed with this assessment and determined that the operators should be reminded to notify radiation protection prior to causing any change in radiological conditions.
V. Manaaement Meetinas
X1 Exit Meeting Summary The inspectors presented the preliminary inspection results to plant management led by Mark Bezilla at the conclusion of the inspection on April 22,1999. PSE&G acknowledged the findings presented.
The inspectors asked PSE&G whether any materials examined during the inspection should be
considered proprietary. No proprietary informatbn was identifie I l
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l INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 60710:
Refueling Activities IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
{
IP 92901:
Followup - Plant Operations IP 92903:
Followup - Engineering IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors j
i ITEMS OPENED AND CLOSED Opened / Closed 50-354/99-02-01 NCV Fuel handling operations. (Section O1.2)
j 50-354/99-02-02 NCV Missed technical specification requirement during reactor i
vessel head tensioning. (Section M1.1)
50-354/99-003 LER As-found values for safety relief valve lift setpoints exceeded technical specification allowable limits. (Section
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E8.3)
Closed 50-354/98-05-01 VIO Failure to establish in service testing requirements.
(Section E8.1)
50-354/98-05-02 VIO Failure to maintain safety related nitrogen bottle pressure regulators at the proper setpoint. (Section E8.1)
50-354/98-05-03 VIO Ineffective corrective action for a change of design basis for the safety related chiller operation. (Section E8.2)
50-354/98-302-02 URI Five licensed operators in jeopardy of not completing the requalification program. (08.1)
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LIST OF ACRONYMS USED
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d/p Differential Pressure EDG Emergency Diesel Generator IRM Intermediate Range Monitor j
IST in Service Testing IWI in-vessel Visual Inspection LOCA Loss of Coolant Accident LOP Loss of Offsite Power LORT Licensed Operator Requalification Training LPRM Local Power Range Monitor l
NRC Nuclear Regulatory Commission PDR Public Document Room PSE&O Public Service Electric and Gas RCIC Reactor Core Isolation Cooling RP&C Radiological Protection and Chemistry
)
ALARA As Low As is Reasonably Achievable CACWS Control Area Chilled Water System l
LER Licensee Event Report l
SRV Cafety Relief Valves J
BWROG Boiling Water Reactor Owners Group f
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AR Action Report SIL Service Information Letter SORC Station Operations Review Committee SRM Source Range Monitor
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SRO Senior Reactor Operation
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j