IR 05000272/1999004
| ML18107A427 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/06/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18107A426 | List: |
| References | |
| 50-272-99-04, 50-272-99-4, 50-311-99-04, 50-311-99-4, NUDOCS 9907120315 | |
| Download: ML18107A427 (22) | |
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Docket Nos:
License Nos:
Report N Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
U.S. NUCLEAR REGULATORY COMMISSION 50-272, 50-311 DPR-70, DPR-75
REGION I
50-272/99-04, 50-311/99-04 Public Service Electric and Gas Company Salem Nuclear Generating Station, Units 1 & 2 P.O. Box236 Hancocks Bridge, New Jersey 08038 April 19, 1999 - May 29, 1999 S. A. Morris, Senior Resident Inspector F. J. Laughlin, Resident Inspector H. K. Nieh, Resident Inspector T. F. Burns, Reactor Engineer G. Smith, Senior Physical Security Specialist P. R. Frechette, Jr., Physical Security Specialist J. T. Furia, Senior Radiation Specialist L. A. Peluso, Radiation Physicist L. Harrison, Reactor Engineer Glenn W. Meyer, Chief, Projects Branch 3 Division of Reactor Projects 9907120315 990706 PDR ADOCK 05000272 G
EXECUTIVE SUMMARY Salem Nuclear Generating Station NRC Inspection Report 50-272/99-04, 50-311/99-04 This integrated inspection included aspects of operations, engineering, maintenance, and plant support. The report covers a six-week period of resident inspection and includes the results of announced inspections by regional inspectors who reviewed of physical security measures, inservice inspection and occupational exposure controls, and a review of the year 2000 progra Operations The conduct of operations was professional and safety-conscious. Operators responded to an unanticipated transient during low power in a timely, appropriate manner. (Section 01.1)
Operators responded appropriately following a Unit 1 reactor trip caused by a grounded control rod circuit. Technicians promptly identified and corrected the cause of the groun (Section 01.2)
Operators safely and effectively coordinated Unit 2 refueling outage activities. (Section 01.3)
Maintenance Human performance errors contributed to inadequate containment fan cooler maintenance and an unplanned technical specification 3.0.3 entry. Continued management attention to human performance was warranted. (Section M 1.1)
PSE&G applied considerable operations, maintenance, and engineering efforts before determining and correcting all contributing causes of the pressure induced gasket leaks on a containment fan cooler unit (CFCU). Although these efforts resulted in increased CFCU unavailability and a one time extension of the technical specification allowed outage time, PSE&G made reasonable attempts to correct the conditions. PSE&G's initial assessment of the leaking CFCU's contribution to a grounded rod control circuit, which resulted in an automatic reactor trip, had been incorrect. (Section M 1.2)
PSE&G's corrective actions for a high head injection check valve failure were reasonable. The 10 CFR 50.72 four-hour report was made two days after the event, which was not timel PSE&G subsequently determined a sufficient basis to retract this report and adequately addressed untimely reporting to the NRC through the corrective action system. (Section M2.1)
For the selected areas observed, PSE&G performed acceptable inservice inspections (ISi)
which included adequate ASME Code program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and contractor oversight. The inspections performed were thorough and of sufficient extent to determine the ii
integrity of the components inspected. Nonconforming conditions were properly documented and resolved in accordance with established requirements. (Section M2.2)
PSE&G's replacement of two ASME Code class 1 valves was well-planned, controlled and coordinated, and included appropriate QA involvement. (Section M2.3)
Engineering Inattention to detail resulted in no update to some Unit 2 reactor engineering procedures and no re-scaling of a pressurizer pressure instrument, which were identified by operators prior to the Unit 2 startup. PSE&G's corrective actions for these issues were reasonable and timel (Section E2.1)
PSE&G took appropriate actions in response to industry operational experience concerning potential stress corrosion cracking of fuel assembly components. A Westinghouse safety assessment provided reasonable assurance for safe plant operation. (Section E2.2)
Plant Support The hot particle exposure event of April 27, 1999, did not result in a skin overexposure to a worker. PSE&G's response to the event was appropriate and timely, and the event was appropriately entered.into the corrective action program. (Section R 1. 1)
PSE&G implemented an effective radiation protection program for the Unit 2 refueling outag Control of work in high radiation areas, especially work inside the bioshield involving steam generator activities, was appropriately planned and implemented, which minimized occupational exposures. Some minor radworker practice issues resulted in a number of clothing contaminations. Prompt identification of these conditions by the radiation protection staff prevented the spread of contamination outside the RC (Section R 1.2)
PSE&G implemented an effective quality assurance program to review health physics program activities. The problem identification system was appropriately used to identify and track corrective actions for deficiencies. (Section R7)
The inspectors reviewed security activities, equipment, procedures, and records, and concluded that the security program performance was acceptable and met regulatory requirements and Security Plan commitments. (Section S1)
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TABLE OF CONTENTS
EXECUTIVE SUMMARY....................................................... ii
TABLE OF CONTENTS........................................................ iv I. Operations........................................................... 1
Conduct of Operations............................................. 1 0 General Comments......................................... 1 01.2 Unit 1 Reactor Trip.......................................... 2 01.3 Unit 2 Refueling Outage Activities.............................. 2 II. Maintenance................... ~ *........................................... 3 M 1 Conduct of Maintenance........................................... 3 M1.1 Unit 2 Refueling Outage Activities.............................. 3 M 1.2 Unit 1 Containment Fan Cooler Unit CCFCU) Service Water (SW) Leak
........................................... *.*............ 5 M2 Maintenance and Material Condition of Facilities and Equipment........... 6 M Failure of 21SJ17 Check Valve During High Head Injection Test..... 6 M2.2 lnservice Inspection...............................,......... 7 M2.3 Replacements in Progress.................................... 9 MB Miscellaneous Maintenance Issues.................................. 10 M (Closed) LER 50-272/99-002-00.............................. 10 Ill. Engineering....... _....................................................... 10 E2 Engineering Support of Facilities and Equipment....................... 10 E Unit2 OTDT Instruments Off-scale High........................ 10 E Fuel Manufacturing Issue Concerning Top Nozzle Spring Screws.... 11 EB Miscellaneous Engineering Issues.................................. 12 E.B.1 Year 2000 Program and Implementation....................... 12 IV. Plant Support............................................................ 12 R 1 Radiological Protection and Chemistry (RP&C) Controls................. 12 R Inadvertent Hot Particle Exposure............................. 12 R1.2 Occupational Exposure Controls During Unit 2 Refueling Outage.... 1 R7 Quality Assurance in RP&C Activities....................... :........ 15 RB Miscellaneous RP&C Issues....................................... 15 R (Closed) Violation 50-272.311/9B-05-12........................ 15 R (Closed) IFI 50-272.311/9B-05-13............................. 16 S1 Security....................................................... 16 V. Management Meetings...................................................... 16 X1 Exit Meeting Summary............................................ 16 iv
Report Details Summary of Plant Status Unit 1 began the period at 100% power. The unit remained at full power until May 20, 1999, when it automatically tripped following a dropped control rod. Operators restarted the unit on May 23, and it was synchronized to the grid on May 24, 1999: The unit achieved full power on May 24, where it remained for the rest of the report perio Unit 2 began the period in a de-fueled condition as part of the tenth refueling outage. Operators completed refueling activities on May 1, 1999, and the unit entered Mode 5 on May Operators completed a reactor startup of the unit on May 26, it commenced power operations on May 27, and was synchronized to the grid on May 28, 1999. The unit was at 45% power at the end of the perio I. Operations
Conduct of Operations 0 General Comments Inspection Scope (71707)
The inspectors conducted frequent observations of ongoing plant operations, including control room walkdowns, log reviews, and shift turnovers. The inspectors also observed portions of reactor startup operations on both units and operator response to an unanticipated transient during a malfunction of the automatic turbine bypass valves on Unit Observations and Findings Overall, the conduct of operations was professional and safety-conscious. The reactor start-ups of both units were well-controlled, characterized by effective supervision, good procedural usage, and good three-point communication The inspector was present in the control room on May 28, 1999, when the Unit 2 turbine bypass valve automatic controls (steam pressure mode) malfunctioned, and caused an increase in steam flow, a drop in plant temperature and a resultant 1 % increase in reactor power from 10%. Operators immediately identified the malfunction and took timely manual actions to stop the transient. The plant was in Mode 2 at the time of the event and the actual transient had minimal safety consequenc Conclusions The conduct of operations was professional and safety-conscious. Operators responded to an unanticipated transient during low power in a timely, appropriate manner.
,
- 0 Unit 1 Reactor Trip
- lnsp~ction Scope (71707. 62707. 37551. 92901)
At 9:37 p.m. on May 20, 1999, the Salem Unit 1 reactor automatically tripped from 100%
power due to a nuclear instrument (NI) negative rate trip signal. The trip signal resulted from the loss of stationary gripper coil voltage to a control rod causing it to drop into the core. The resultant power decrease was sensed by two of the four power range NI channels, which provided the necessary coincidence for a reactor protection system trip signal. The inspectors responded to the site to assess the nature of the event and to follow up on the subsequent plant recover Observations and Findings All plant safety systems functioned as designed following the trip, and operators implemented timely and appropriate actions in accordance with established procedure The plant was stabilized in Mode 3 (hot standby) with the reactor coolant system at normal operating temperature and pressure. Operators completed a 10 CFR 50.72 non-emergency event notification to the NRC within the required four hour PSE&G determined the causes of the trip and implemented reasonable corrective actions prior to restarting the unit. The loss of rod gripper voltage resulted frorn a blown fuse, due to a ground in the affected circuitry. Technicians isolated the ground to a wetted electrical junction box inside containment that contained* two faulty cables, which were replaced. PSE&G also inspected three other nearby junction boxes that contained rod control cables and found no indications of damage. The source of water was determined to be from a service water leak on a containment fan cooler unit (CFCU),
which was subsequently repaired (see Section M1.2). Engineering management concluded that there was reasonable assurance that these immediate corrective actions were sufficient to return the unit to power. This information was presented to the station operations review committee (SORC) during their post-trip review. The SORC members approved plant startup when all of the necessary repairs were complete Conclusions Operators responded appropriately following a Unit 1 reactor trip caused by a grounded control rod circuit. Technicians promptly identified and corrected the cause of the groun.3 Unit 2 Refueling Outage Activities Inspection Scope (71707)
. The inspectors observed portions of outage activities including refueling, and various surveillance and maintenance evolutions.
3 Observations and Findings PSE&G personnel effectively coordinated refueling operations in the containment and fuel handling building. Core reload was completed in a slow, deliberate manner and without incident. The new equipment hatch door maintained containment integrity during refueling operations. Personnel were stationed at the door to close penetration isolation valves in case of an emergenc Control room operators properly maintained core cooling via the residual heat removal (RHR) system and closely controlled mid-loop operations when the reactor coolant system was in a reduced inventory status. Inspectors observed activities associated with the 21 containment spray pump full-flow test. These were well-controlled with good pre-job briefings, procedural use and communications. PSE&G also stationed an equipment operator in the containment to coordinate maintenance activities, which was a good initiativ Conclusions Operators safely and effectively coordinated Unit 2 refueling outage activitie II. Maintenance
M1 Conduct of Maintenance (50001, 62707, 61726, 92902, & 40500)
M1.1 Unit 2 Refueling Outage Activities Inspection Scope (62707. 71707. 92902)
The inspectors followed up on some human performance errors which occurred during outage maintenance and surveillance testing activitie Observations and Findings Containment Fan Cooler Unit CCFCU) Maintenance During maintenance on the five CFCUs, the fan coolers were re-assembled using Duratough DL gasket sealant. PSE&G determined that maintenance procedure SC.MD-PM.CBV-0004(Q), revision 11, Containment Fan Coil Unit Heat Exchangers Internal Inspection, was not followed concerning the amountof sealant and its cure tim Specifically, $teps 5.4.2 and 5.4.3 concerning the correct amount of sealant (.007 inches)
was marked "not applicable (NIA)" and an excessive amount of sealant (318 inch) was applied. Likewise, step 5.4.8 concerning the sealant cure time was marked NIA to permit a shorter cure time than what was prescribed. In both instances, contractor supervision marked the procedure steps as NIA. This is a violation of Unit 2 technical specification (TS) 6.8.1.a in that a required maintenance procedure was not properly implemented.
This Severity Level IV violation is being treated as a non-cited violation consistent with Appendix C of the NRC enforcement policy. (NCV 50-311/99-04-01)
This condition was discovered during post-maintenance testing when a cooler leaked during a hydrostatic test. It resulted in excessive sealant material being forced into the tube sheet region of the heat exchangers, which blocked fan coil tubes and potentially reduced the heat removal capability of the CFCUs. It also showed ineffective maintenance supervision and resulted in extensive rework activities. PSE&G's corrective actions included the rework of all CFCUs which were repaired improperly and procedure revisions to ensure that only PSE&G supervisors be permitted to mark procedural steps as not applicabl Unplanned Technical Specification (TS) 3.0.3 Entry During the performance of procedure S2.0P-ST.SSP-0001 (Q), Manual Safety Injection -
SSPS, revision 18, step 5.4.4 directed operators to place the control switches for the 12 and 22 control room emergency air conditioning system (CREACS) supply fans (two per unit/train) in the stop position. This rendered both CREACS trains inoperable since TS 3.7.6.1.a specifies that both supply fans are required for an operable filtration train. This condition placed Unit 1, operating at 100% during the test, into TS 3.0.3 for a condition not covered by T A good questioning attitude by a shift operator identified this deficiency. The plant was not outside the design basis since one CREA CS fan provides 100% of required air flo Also, the fans were technically under administrative control during the test and could have been restored to operable status in a short time. The condition existed for ten minutes, well within the TS allowed outage time of one hour for each unit. The inspectors concluded that weak supervisory oversight combined with the inadequate surveillance procedure contributed to this event. This Severity Level IV violation of TS 3.7.6.1 is being treated as a non-cited violation consistent with Appendix C of the NRC Enforcement Policy. (NCV 50-311/99-04-02)
Miscellaneous Issues Two additional human performance errors of lesser significance occurred. Equipment operator inattention to detail resulted in the inadvertent isolation of control power to the 21 safety injection pump. There was no safety consequence since Unit 2 was in cold shutdown at the time and this equipment was not required. Also, inattention to detail by a maintenance supervisor and technician resulted in a minor electrical shock to the technician. This could have resulted in serious injury to the technician under different voltage conditions. The inspectors concluded that these events were minor violations which are not subject to formal enforcement action. However, these human performance errors demonstrated that continued management attention was warranted in this are These issues were documented in PSE&G's corrective action system and received appropriate management attention.
- Conclusions Human performance errors contributed to inadequate containment fan cooler maintenance and an unplanned technical specification 3.0.3 entry. Continued management attention to human performance was warrante M1.2 Unit 1 Containment Fan Cooler Unit (CFCU) Service Water (SW) Leaks Inspection Scope (61726. 62707. 40500) Following the investigation of an automatic Unit 1 reactor trip (see Section 01.2), PSE&G determined that a potential cause of a grounded rod control circuit was SW leaking from the 11 CFCU. The inspectors reviewed the events contributing to the leaks and PSE&G's corrective actions. Efforts to repair the leaks continued through the end of the inspection period, and led to the NRC's granting of a Notice of Enforcement Discretion (NOED) request on June 4, which authorized a one time extension of the technical specification (TS) allowed outage time (AOT) of seven day Observations and Findings PSE&G determined that the source of the water that potentially contributed to the ground was due to a May 18 SW leak from the 11 CFCU. This leak was apparently caused by surveillance testing using procedure S1.OP-ST.SW-001 O(Q), "lnservice Testing Containment Fan Cooler Unit Service Water and Control Air Valves." The operator performing the test first shut the CFCU SW outle~ and inlet valves. Next, he reopened the outlet valve, which partially drained the SW from the coolers, since the coolers are at a higher elevation than the SW outlet header. The pressure transient resulted when the operator reopened the inlet valve, causing the voided SW piping to become rapidly filled, damaging the gaskets on several CFCU coolers. The inspectors noted that the.
procedure was vague enough to allow this sequence of operation. PSE&G repaired the gasket leaks and restored the CFCU to service on May 22 within the seven-day AO Additionally, PSE&G concluded that the apparent cause of the pressure transient was the improper sequencing of the SW valves, and revised the noted procedure to include specific guidance on the proper sequence of valve operation for preventing a pressure transien The surveillance test was performed again on May 29 to synchronize the surveillance test interval with the on-line work planning schedule. Again, the 11 CFCU developed gasket leaks from its coolers. PSE&G's investigation determined that SW outlet valves were degraded and leaking by, allowing the draining and partial voiding of the coolers, and that improper valve operation sequence alone may not have caused this and the May 18 pressure transien Corrective actions from this second event included revising the test procedure to shut additional outlet valves to prevent voiding, and repairing the degraded valves and damaged gaskets. During post-maintenance testing, gasket leaks occurred again, and
- PSE&G determined that another factor contributing to the cooler leaks was the type of gasket being used. Since insufficient time remained to effect repairs before the expiration of the AOT, PSE&G requested a Notice of Enforcement Discretion (NOED) to extend the AOT _an additional five days to repair the coolers with different gaskets, which are identical to those being used in the Unit 2 CFCUs. These gaskets are of better design, in that they are thicker, less susceptible to installation problems, and installed with a more durable adhesiv Following thorough review of root causes, proposed actions, and risk information, the NRC granted the NOED on June 4, 1999. PSE&G personnel satisfactorily repaired the 11 CFCU and exited the TS action statement and the NOED on June 7, 199 The inspectors reviewed PSE&G's root cause analysis for the noted Unit 1 reactor trip, and learned that PSE&G ultimately concluded that no direct connection existed between the leaking CFCU and the grounded rod control circuit. However, the analysis.also stated that moisture, combined with existing insulation damage on a conductor in the affected rod control circuit, caused the ground fault and the resulting reactor trip. The sources of moisture described were containment humidity and leakage from the CFC The inspectors concluded that PSE&G's analysis inconsistently assessed the impact of the leaking CFCU to the reactor trip. Recognizing that damaged insulation alone does not necessarily result in a pathway to ground, the inspectors concluded that the SW leak from the CFCU more likely provided a pathway to ground than normal humidity in the containment building.
Conclusions PSE&G applied considerable operations, maintenance, and engineering efforts before determining and correcting all contributing causes of the pressure induced gasket leaks on a containment fan cooler unit (CFCU}. Although these efforts resulted in increased CFCU unavailability and a one time extension of the technical specification allowed outage time, PSE&G made reasonable attempts to correct the conditions. PSE&G's initial assessment of the leaking CFCU's contribution to a grounded rod control circuit, which resulted in an automatic reactor trip, had been incorrec M2 Maintenance and Material Condition of Facilities and Equipment M Failure of 21SJ17 Check Valve During High Head Injection Test Inspection Scope (92901. 92902. 92903)
On May 1, 1999, while operators were performing an 18-month high head injection test, there was no detectable flow in the 21 cold leg flowpath. Salem Unit 2 was in Mode 6 (refueling} at the time of the event and high head injection was not required to be operable. The inspectors followed up on this event to assess PSE&G's corrective actions.
7 Observations and Findings This surveillance test establishes proper high head injection throttle settings on each of the four flowpaths (21 to 24) to the reactor coolant system (RCS). No flow was detected in the 21 flowpath until flow was throttled to the other three. When flow initiated, operators heard a sound like that of a stuck valve unseating. Initial investigation revealed no problems with either the throttle or manual valve in the 21 cold le Engineers concluded that the 21SJ17 check valve must have stuck closed and prevented flow, possibly due to the presence of foreign material. This valve was replaced and a subsequent test was satisfactor Inspection of the 21SJ17 valve revealed no wear or defects, but internal valve clearances were at the low end of acceptability, making it susceptible to malfunction from system particulate. There was no indication that the valve had been in a degraded condition for any length of time. Inspectors concluded that PSE&G's approach to correct this problem was reasonable, but that no definite root cause was identifie PSE&G completed a 10 CFR 50.72 four-hour report to the NRC on May 3, 1999, for an event found while the reactor was shutdown that would have seriously degraded the plant or resulted in an unanalyzed condition. This report was prompt once the determination was made but it appeared that the determination was not timely based on the May 1 event date. However, subsequent engineering analysis showed that this condition did not place the plant outside the design basis and the report was retracte Additionally, PSE&G initiated a significance level 1 (highest) action request 990505096 to investigate untimely reporting of events to the NRG. The inspectors concluded that this issue posed no safety concern and that it received adequate management attention through the corrective action sy_ste Conclusions PSE&G's corrective actions for a high head injection check valve failure were reasonable. The 10 CFR 50. 72 four-hour report was made two days after the event, which was not timely. PSE&G subsequently determined a sufficient basis to retract this report and adequately addressed untimely reporting to the NRG through the corrective action syste M lnservice Inspection CISI) Inspection Scope (73753)
The inspector reviewed plans and schedules for the current ISi interval (second interval, tenth refueling outage) to verify compliance with the requirements of the ASME Code Section XI, 1986 edition, and 10 CFR 50.55a(g). Areas inspected included ASME Section XI ISi program coverage, qualifications and certifications of the nondestructive examination (NOE) personnel, NOE procedures, results of NOE, and oversight of NOE contractors. In addition, the inspector observed selected NOE activities, including ultrasonic {UT) and penetrant (PT) examinations of a pipe to elbow weld on the safety
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injection system. The inspector also observed the remote eddy current testing (ET) of steam generator tubin Observations and Findings The ISi procedures had been approved by the ISi contractor, PSE&G and the authorized nuclear inspector and were in accordance with the ASME Code requirements. The work performed was found to be thorough and of sufficient extent to determine the integrity of the components inspected. The inspector reviewed portions of the UT, PT and ET procedures used by NOE personnel and found them to be adequate for the NOE tasks performed. PSE&G had NOE contractors perform the ISi examinations and provided oversight, which included a review and approval of personnel and procedure qualifications, monitoring of activities and a review and acceptance of test results. The inspector found that ISi implementation was consistent with the approved PSE&G procedures. NOE personnel qualification records complied with ASME Code requirement The inspector evaluated PSE&G's oversight of contractor NOE and component replacement activities by reviewing the oversight plans, reports and weld travelers which documented appropriate PSE&G involvement in the NOE and welding activities performed. PSE&G's oversight plan was thorough with sufficient concentration on field activities to determine ISi performance in the areas of program compliance and technical adequacy.
The inspector observed PSE&G's training regarding the robotic ET equipment in the steam generator primary (bottom) head mock-up and discussed the activity with training personnel and craftsmen. The training was thorough, conducted professionally, and reflected conditions and obstructions that would be encountered in the field. Instructors emphasized personal radiation protection and safet The inspector evaluated steam generator tube ET through procedure review, examination of test equipment, and observation of operator actions, equipment function, data collection and data evaluation and resolution. The inspector discussed the ET activity with PSE&G staff, test equipment operators, data analysts and resolution analysts and concluded that the process was well-planned and controlled at a level that would provide an accurate and reliable assessment of steam generator tube integrit The inspector reviewed the results of radiographic inspections of main feedwater flow nozzles for two steam generators. This ISi activity.revealed the presence of linear indications on the inside diameter of the nozzle-to-fitting weld which was later confirmed by UT examination. PSE&G characterized and sized the indication(s) using the UT process and initiated the appropriate actions to disposition these indications. These indications exceeded the acceptance criteria of ASME Section XI and repair/replacement was required. The inspector concluded that PSE&G took appropriate action to identify, characterize, size and capture this data for evaluation.
- The inspector verified that examination data was in accordance with the ISi procedures and ASME Code requirements. NOE personnel performing inspections properly identified and explored indications to determine their relevance. The tracking of ISi examination results indicated that PSE&G's ISi program was in compliance with the ASME Code Section X Conclusions For the selected areas observed, PSE&G performed acceptable inservice inspections (ISi) which included adequate ASME Code program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and contractor oversight. The inspections performed were thorough and of sufficient extent to determine the integrity of the components inspected. Nonconforming
.
conditions were properly documented and resolved in accordance with established requirement M Replacements in Progress Inspection Scope The inspector reviewed PSE&G's procedure for documenting and controlling repairs, replacements, and modifications to nuclear ASME Code items. The procedure delineated responsibilities, sources, methods and means to assure that processes such as welding, nondestructive examination, heat treating and testing were performed in accordance with applicable requirements. The inspector reviewed one code job package (CJP) from its inception to completion of the replacement activity. The CJP involved the replacement of two ASME Code Class I, seismic class 1 valves and associated piping in a safety injection system drain lin Observations and Findings The inspector reviewed the CJP which documented the replacement history of the two valves. The inspector examined the equivalency evaluation, material certifications, weld filler metal certifications, weld procedure, welders identification, and assembly drawings in the CJP and found this documentation to be in compliance with the applicable code requirements. Because the original valves were obsolete, PSE&G had to perform an equivalency evaluation for the replacement valves. The inspector examined the fabrication of the two replacement valves and the associated piping and pipe ca PSE&G had replaced the valves in a well-controlled manner and in accordance with specified requirements. Documentation and tracking mechanisms were in place with hold points established at critical steps in the process. Nondestructive tests and acceptance criteria were specified for base material weld zones and the completed welds. The fabrication was acceptable as completed, and no repairs or rework were required. QA involvement provided adequate oversight and verification that the specified acceptance criteria of the pipe to valve welds met the ASME Code requirement *
10 Conclusions PSE&G's replacement of two ASME Code class 1 valves was well-planned, controlled and coordinated. and included appropriate QA involvemen MB Miscellaneous Maintenance Issues M (Closed) LER 50-272/99-002-00: Auxiliary Building Ventilation Found Outside of Design E2 E The inspectors conducted an onsite review of the licensee event report (LER) and verified selected corrective actions using inspection procedures 92700, 92901, 9290 This issue was previously reviewed by the NRC in Inspection Report 50-272 & 311/99-02 (section M1.3), and resulted in three separate non-cited violations. PSE&G's corrective actions for the issue were adequate. The LER did not contain explicit corrective actions for the untimely notification of a condition outside of the plant's design basis; however, PSE&G appropriately addressed this issue in their corrective action program (AR 990505096). Additionally, PSE&G determined after further investigation that the auxiliary building low differentiai pressure alarm did not function during the event. The alarm is currently scheduled for repair in July 199 The inspectors concluded that the immediate and planned corrective actions associated with the event were adequate and that this LER was closed.
Ill. Engineering Engineering Support of Facilities and Equipment Unit 2 OTDT Instruments Off-scale High Inspection Scope (37551. 62707)
Three out of four Unit 2 OTDT instruments were reading off-scale high at normal operating temperature and pressure when procedural guidance indicated that the instruments should have been on-scale. This prevented an accurate instrument channel check, which was required for a reactor startup. The inspectors followed up on the reason for the occurrence and PSE&G's corrective action Observations and Findings PSE&G's investigation revealed that the reactor engineering procedure for Unit 2 had not been revised to reflect the new values resulting from the fuel upgrade margin recovery program (FUMRP). The new values caused the instruments to be off-scale high in a
.. certain temperature range. Additionally, the fourth OTDT instrument was on-scale due to the fact that its associated pressurizer pressure instrument, which provides one input to OTDT, had not been re-scaled to the new FUMRP value due to a scheduling erro * E * PSE&G's corrective actions included the revision of the applicable reactor engineering procedure to reflect the new FUMRP values, re-scaling of the deficient pressurizer pressure instrument, and a design change to ensure that the OTDT instruments were on-scale throughout their temperature range. PSE&G also initiated a significance level 1 action request 990522146 to investigate the root cause of engineering data not being updated in procedures in a timely manner. Inspectors concluded that these actions were reasonable and that there was no safety issue since the situation was corrected before the Unit 2 reactor startup. PSE&G had previously revised the Unit 1 procedures to reflect the new.data and intended to complete the design change to Unit 1 OTDT instruments as wel Conclusions Inattention to detail resulted in no update to some Unit 2 reactor engineering procedures and no re-scaling of a pressurizer pressure instrument, which were identified by operators prior to the Unit 2 startup. PSE&G's corrective actions for these issues were reasonable and timel Fuel Manufacturing Issue Concerning Top Nozzle Spring Screws Inspection Scope (37551, 92903)
PSE&G performed a visual inspection of certain fuel assemblies in response to operating experience at two other Westinghouse plants concerning the stress corrosion cracking of spring screws on the assemblies. The inspectors followed up on this issue to determine the safety impact for Unit 2 operation Observations and Findings PSE&G's visual inspection of suspected fuel assemblies revealed a slight gap between the top nozzle block and the spring hold-down block. However, this gap was within manufacturing tolerance, and there were no indications of spring screw failures. The failures observed in the industry have occurred during the second cycle of operation and the fuel assemblies in question have only been through one cycl Westinghouse provided a safety assessment which demonstrated that there would be no impact on safe operation from spring screw failures. This was approved by the station operations review committee (SORC) as part of the Unit 2 cycle 11 safety evaluatio The inspectors reviewed this analysis, and concluded that it was thorough and provided reasonable assurance for safe operation. PSE&G also intended to perform a review of the Westinghouse root cause analysis when available for the spring screw failures, and to use this information for further corrective actions at both units. The issue was appropriately documented in PSE&G's corrective action syste * Conclusions PSE&G took appropriate actions in response to industry operational experience concerning potential stress corrosion cracking of fuel assembly components. A Westinghouse safety assessment provided reasonable assurance for safe plant operatio ES Miscellaneous Engineering Issues E. Year 2000 Program and Implementation A review was conducted of Salem Year 2000 (Y2K) activities using NRC Temporary Instruction (Tl) 2515/141, "Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants." The review included aspects of PSE&G's Y2K management planning, assessment, documentation, and remediation activities. PSE&G's Y2K testing and validation, notification activities, and contingency plans were also reviewed. Th NRC reviewers used NEl/NUSMG 97-07, "Nuclear Utility Year 2000 Readiness," and NEl/NUSMG 98-07, "Nuclear Utility Year 2000 Readiness Contingency Planning," as the primary references for this review. The detailed results of this review will be combined with similar reviews of Y2K programs at other U.S. commercial nuclear power plants and summarized in a report to be issued by the NRC staff by July 31, 199 *
- 1v. Plant Support
R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Inadvertent Hot Particle Exposure Inspection Scope (83724) A reactive health physics inspection was conducted to review the circumstances surrounding an unplanned hot particle exposure to a worker on April 27, 1999. This included a review of the: (1) identification of the hot particle; (2) calculations for exposure to the skin of the worker; (3) documentation of exposure; (4) documentation of the event; and, (5) proposed and completed corrective action Observations and Findings On April 27, 1999, with Unit 2 in a refueling outage, a radiation worker entered the containment to operate the polar crane. The worker entered the containment at approximately 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br />, took a break at approximately 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />, re-entered the containment at 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br />, and exited containment at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on April 28, 1999. At*
the time of this exit, the worker alarmed the portal monitors located at the radiologically controlled area (RCA) exit. PSE&G radiation protection staff promptly responded to this alarm which rapidly led to the identification of a hot particle located on the upper right knee of the worker. Preliminary calculations by the radiation protection staff based on
- dose rates measured by an open windowed model R0-2 survey meter resulted in a skin dose.estimate of 48 rem. Based on more detailed analysis, PSE&G assigned an exposure to the skin at the knee of 21.1 rem from this event. (The regulatory limit is 50 rem per 10 CFR.20.1201 (a)(2)(ii)).
Personnel contamination event (PCE) report #99-073 was immediately initiated, and subsequently documented as a level 2 action request, #990428095. The particle, which was readily removed from the worker's knee, was analyzed by gamma spectroscopy and identified as cobalt-60. The results from this analysis were entered into a dose calculation code (VARSKIN). Additional investigation by PSE&G staff identified that the worker had worked near the refueling cavity around midnight, kneeling on the floor to perform some work. This area was adjacent to a highly congested area where there was a step-off pad used by workers exiting the reactor cavity after performing flange seal inspection/cleaning. The flange seal area is a common location for finding hot particle Since this was the first time the worker knelt while working during this shift, PSE&G assumed it was the most likely location and time for the hot particle contaminatio PSE&G's subsequent actions included a detailed survey of the refueling floor as well as most other areas of the containment. No additional hot particles were identified during this survey. Proposed long term corrective actions included a review and revision, as necessary, of the procedures and practices for establishing boundaries around suspected hot particle areas. This included ensuring that adequate space is available for removing protective clothing upon exiting these areas.
Conclusions The hot particle exposure event of April 27, 1999 did not result in a skin overexposure to a worker. PSE&G's response to the event was appropriate and timely, and the event was appropriately entered into the corrective action progra R1.2 Occupational Exposure Controls During Unit 2 Refueling Outage Inspection Scope (83750>
The inspectors conducted a focused health physics practices assessment during the Unit 2 refueling outage, which began in April 1999. Areas of inspection focus were based on the following regulatory requirements from 10 CFR 20:
20.1101 20.1601 20.1602 20.1902 20.1904 20.2103 Radiation protection program Control of access to high radiation areas Control of access to very high radiation areas Posting requirements Labeling containers Records of surveys Direct observation of outage work in progress, including steam generator eddy current testing and secondary side steam generator cleaning, was conducted throughout the
inspection. Additional inspection focus included direct observation of in-process work in the radiologically controlled areas (RCAs), review of pertinent documents including surveys, radiation work permits (RWPs) and as low as is reasonably achievable (ALARA) reviews, and discussions with cognizant personne Observations and Findings PSE&G established an occupational exposure goal of 160 person-rem for the Unit 2 refueling outage (RF-10). This goal appeared to be based on appropriate plannin Through the first two and one-half weeks of the outage, exposures tracked below estimated projection The inspectors reviewed controls for work conducted in high radiation areas, especially inside the bioshield, and determined that they met the applicable requirements of 10 CFR 20. In addition to required postings and briefings given to workers entering this area, a number of informational postings identifying areas of special significance for maintaining occupational exposures ALARA were in place. Although the area inside the bioshield was highly congested due to simultaneous work involving eddy current testing, pump seal replacement and secondary side steam generator cleaning, PSE&G established effective work controls to maintain exposures ALAR Control of work conducted near or inside the primary side of the steam generators involved a number of additional controls to minimize personnel exposure and to aid in minimizing heat stress conditions. Workers on the steam generator platforms maintained constant communications, both audio and visual, with a designated control station established at the site. Continuous health physics staff coverage through this closed circuit system, augmented with periodic direct coverage by roving technicians inside the bioshield, provided appropriate radiological work control PSE&G identified. a number of low level clothing contamination events during the first 20 days of the outage. In general, these events were the result of poor radiological practices exhibited by radiological workers. Actions taken by management appeared to be reasonable and included the issuance of written communications to the plant staff stressing the importance of radworker adherence to health physics requirements, and a heightened level of first line supervisory oversight of workers performing activities in the RCA. The inspectors verified that none of the contamination events resulted in measurable levels of radiation exposur Conclusions PSE&G implemented an effective radiation protection program for the Unit 2 refueling outage. Control of work in high radiation areas, especially work inside the bioshield involving steam generator activities, was appropriately planned and implemented, which minimized occupational exposures. Some minor radworker practice issues resulted in a number of clothing contaminations. Prompt identification of these conditions by the radiation protection staff prevented the spread of contamination outside the RC *
R7 Quality Assurance in RP&C Activities
- Inspection Scope (83750)
The inspectors examined PSE&G's quality assurance (QA) program in the health physics area, specifically as it related to outage activities. The inspector reviewed audits, surveillances and appraisals previously completed or currently ongoing, and discussed the results of these reviews with cognizant personne * Observations and Findings Beginning in January 1999, PSE&G modified the audits and surveillances program to be a continuous assessment program. Audits and surveillances were performed on a continuing basis, with the results "rolled up" quarterly. A formal matrix of areas to be examined in the health physics area was established, using a three year planning cycl In this way, a QA review of each major program functional area would be performed on at least a triennial basis. Additionally, QA performed periodic monitoring of specific activities or program areas, typically performing these in a day or less. The scope and depth of these activities appeared sufficient to independently identify adverse conditions within the health physics progra All QA identified deficiencies were entered into the issue tracking system as a corrective action request (AR) or condition resolution (CR), depending on the severity of the deficiency. All ARs and CRs identified by QA were assigned a condition resolution quality verification (CRQV) to aid in tracking QA identified issues. A review of outstanding CRQVs indicated that 10 open items existed at the time of this inspection, none of which were beyond their response due date. None of the issues involved issues of safety significanc Conclusions PSE&G implemented an effective quality assurance program to review health physics program activities. The problem identification system was appropriately used to identify and track corrective actions for deficiencie *
RS Miscellaneous RP&C Issues R (Closed) Violation 50-272.311/98-05-12: Failure to Establish Environmental Sampling Procedures to Collect Drinking Water. Fish and Invertebrate Samples The inspectors conducted an in-office review of PSE&G's violation response, as indicated in their letter dated August 3, 1998, using inspection procedure 84750. On June 9, 1998, the inspectors determined that PSE&G had not established written vendor procedures for sampling potable (drinking) water and that the procedure for aquatic media (fish and invertebrates) lacked sample collection guidance. In response, PSE&G implemented corrective actions that included establishment of a vendor procedure for sampling potable water, and a revision to the existing vendor's procedure for sampling
- fish and invertebrates. These procedures contained a description of sample collection methodology in sufficient detail without being prescriptive. PSE&G incorporated standard collection methodologies from the American Society for Testing and Materials (ASTM) for each sample type into the procedures. The inspectors determined these corrective actions to be acceptable. This item is close R (Closed) IFI 50-272.311/98-05-13: Channel Functional Test for Meteorological Sensors The inspector opened this item to track PSE&G's evaluation of their calibration procedure for wind speed sensors to determine the appropriateness of including a channel functional test. PSE&G revised the procedure to include a channel function test as part of the periodic channel calibration. The functional test included spinning the wind speed transmitter at a known rotation rate and verifying the channel output display. The procedure revision also included a specific definition of a channel calibration and a channel functional test. Although no prescriptive requirement existed regarding channel functional test methodology, information is documented in the American Society for Testing and Materials (ASTM) Code. Based on PSE&G's amendment to the channel calibration methodology to include a channei functional test for meteorological senors, this item is close S1 Security From April 19-22, 1999, the inspectors reviewed the security program using inspection procedure 81700. The inspection covered both Hope Creek and Salem and the details are described in Hope Creek Inspection Report 50-354/99-0 Overall Security Program Conclusions:
The inspectors reviewed security activities, equipment, procedures, and records, and concluded that the security program performance was acceptable and met regulatory requirements and Security Plan commitment V. Management Meetings X1 Exit Meeting Summary On June 4, 1999, the inspectors presented their overall findings and conclusions to members of PSE&G management led by Mr. Lou Storz, the Senior Vice President, Nuclear Operations. The inspectors held separate exit meetings on April 22, 1999, to present the results of the occupational exposure control inspection and on April 23, 1999, for the physical security and inservice inspections. PSE&G management acknowledged the findings presented and did not contest any of the inspector's conclusions. Additionally, they stated that none of the information reviewed by the inspectors was considered proprietary.
- IP 37551:
IP 40500:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 73753:
IP 81700:
IP 83724:
IP 83750:
IP 84750:
IP 90712:
IP 92700:
IP 92901:
IP 92902:
IP 92903:
IP 92904:
IP 93702:
INSPECTION PROCEDURES USED Onsite Engineering Effectiveness of Licensee Controls in lderitifying, Resolving, and Preventing Problems Surveillance Observations Maintenance Observations Plant Operations Plant Support Activities lnservice Inspection Physical Security Program for Power Reactors Occupational Exposure Control and Personnel Dosimetry Occupation Radiation Exposure Environmental Monitoring lnoffice Review of Written Reports of Nonroutine Events at Power Reactor Facilities Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities Plant Operations Followup Maintenance Followup Engineering Followup Plant Support Followup Event Followup ITEMS OPENED AND CLOSED Opened/Closed 50-311199-04-01 50-311199-04-02 Closed 50-272&311/98-05-1 &311 /98-05-13 50~272/99-002-00 NCV Containment fan cooler unit maintenance. (Section M 1.1)
NCV Unplanned technical specification 3.0.3 entry. (Section M1.1)
VIO Failure to establish environmental sampling procedures to collect drinking water, fish, and invertebrate sample (Section R8.1)
IFI Determine the appropriateness of performing a channel functional test for meteorological sensors. (Section R8.2)
LER Auxiliary building ventilation found outside of desig (Section M8.1)
A LARA AOT AR ASTM CAS CCTV CFCU CFR CJP CR CREA CS CRQV ET FUMRP ISi LER N/A NCVs NOE NI NOED NRC OTDT PA PCE PDR PSE&G PT QA RCA RCS RHR RP&C RWP SAS SFM SORC SW T&Q the Plan TS UT
LIST OF ACRONYMS USED As Low As Is Reasonably Achievable Allowed Outage Time Action Request American Society for Testing and Materials Central Alarm System Closed Circuit Television Containment Fan Cooler Unit Code of Federal Regulations Code Job Package Condition Resolution Control Room Emergency Air Conditioning System Condition Resolution Quality Verification Eddy Current Testing Fuel Upgrade Margin Recovery Program lnservice Inspection Licensee Event Report Not Applicable Non-Cited Violations Nondestructive Examination Nuclear Instrument Notice of Enforcement Discretion Nuclear Regulatory Commission Over-temperature Delta Temperature Protected Area Personnel Contamination Event Public Document Room Public Service Electric and Gas Penetrant Quality Assurance Radiologically Controlled Area Reactor Coolant System Residual Heat Removal Radiological Protection and Chemistry Radiation Work Permit Secondary Alarm System Security Force Member Station Operations Review Committee Service Water Training and Qualification NRG-Approved Physical Security Plan Technical Specification Ultrasonic l