IR 05000272/1987028
| ML18093A518 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/02/1987 |
| From: | Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18093A517 | List: |
| References | |
| 50-272-87-28, 50-311-87-30, GL-87-12, IEB-87-001, IEB-87-1, NUDOCS 8712080124 | |
| Download: ML18093A518 (62) | |
Text
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Report No Docket No License No Licensee:
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
50-272/87-28 50-311/87-30 50-272 50-311 DPR-70 DPR-75 050272-870525 050272-870917 050311-870930 050311-871002 Public Service Electric and Gas Company 80 Park Plaza Newark, New Jersey 07101 Facility Name:
Salem Nuclear Generating Station - Units 1 and 2 Inspection At:
Hancocks Bridge, New Jersey Inspection Conducted:
September 29, 1987 - November 2, 1987 Inspectors:
Approved by:
T. J. Kenny, Senior Resident Inspector K. Halvey Gibson, Resident Inspector
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P. D. Swetland, Chief, Reactor Projects Section No. 2B, Projects Branch No. 2, DRP Inspection Summary:
/~~7 date Inspections on September 29, 1987 - November 2, 1987 (Combined Report Numbers 50-272/87-28 and 50-311/87-30)
Areas Inspected:
Routine inspection~ of plant operations including:
followup on outstanding inspection items, operational safety verification, maintenance, surveillance, review of special reports, licensee event followup, Appendix R followup, Generic Letter 87-12,Bulletin 87-01, and regional meetings with licensee manigemen The inspection involved 226 inspector hours by the resident NRC inspectors in.eluding backshift inspections on September 30, October 8 and 9 and weekend inspections on October 4 and 1 Results:
This inspection covered the licensee's progress during a refueling shutdown on Unit 1 and the voluntary shutdown of Unit 2 as a result of the licensee's inability to produce a justification for continued operation regarding inadequate electric breaker coordinatio The report also documents licensee-identified violations related to unauthorized entries into vital areas, a missed surveillance test and a failure to retest an emergency diesel generator output breaker (Sections 3 & 7).
8712080124 871203 PDR ADOCK 05000272 G
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DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspection activit.
Followup on Outstanding Inspection Items (Closed)
Inspector Follow Item (50-272/83-36-01); This item was opened to follow the installation of a design change that would prevent the inadvertent blocking of the isolation signal for containment ventilation (necessary for containment isolation) from radiatio~ monitoring channel R4 The inspector verified that design change lEC-1790 has been completed and an alarm is present in the control room when R41 has failed or is placed in bypass for functional testin The inspector considers this item close (Closed)
Unresolved Item (50-272/85-12-03); This item concerned potential seismic interactions between flux mapping system and seal tabl The concern was addressed and closed in Combined Inspection Report 50-272/86-19; 50-311/86-1 (Closed)
Unresolved Item (50-272/85-28-01; 311/85-30-01); This item was opened to review the development of the design change proces The inspector reviewed selected Unit 1 seventh refueling outage design change packages and draft engineering procedure GM8-EMP-99 This draft design change procedure is planned to be implemented by the end of 1987 and will supersede the present design change procedure GM8-EMP-00 The inspector noted that organization, control and content of design change pqckages have improved since this item was opened and the licensee's procedures have undergone several revisions which have affected the improvement As a result of the recent Engineering and Plant Betterment reorganization, the design change process was evaluated and restructure GM8-EMP-999 will consist of standardized workbooks for development of design change package The inspectors are following development and implementation of this procedur This item is close (Closed)
Unresolved Item (50-272/85-28-02; 311/85-30-02); This item was opened to review safety evaluation summarie The inspector reviewed selected Salem Units 1 and 2 monthly operating reports for the period January, 1985 through September, 1987, and noted that the author of the 10 CFR 50.59 safety evaluation summaries had changed in September, 1986, and that since then detail and quality of the summaries has improve This item is close *
(Closed)
(Closed)
Inspector Follow Item (50-272, 311/86-09-01); This item was opened because casualty situations were addressed by both an Emergency Operating Procedure (EDP) and an Emergency Instruction (EI) and the lower procedure (EI) did not reference the higher procedure (EDP).
The licensee has rewritten EOP 1 s and has revised the EI 1 s to reference the EOP 1 The licensee is currently converting the EI 1 s to Abnormal Operating Procedure The inspector considers this item close Inspector Follow Item (50-272, 311/86-11-02)~ This item involved licensee identification and correction of environmental qualification deficiencies and the need for an NRC followup inspectio Subsequently an NRC inspection was performed (Combined Inspection 272/86-23; 311/86-23) which identified and opened further item This item is closed based on subsequent closure of these inspection 86-23 item (Closed)
Inspector Follow Item (50-311/85-15-01); This item was opened due to the steam generator level channel, channel check surveillance not meeting the specified acceptance criteri Subsequently, the licensee performed a venting procedure on the level channel transmitte The inspector verified that the steam generator level channels are now functioning as designe This item is close (Clos~d) Unr*esolved Item (50-311/86-29-01); Determine the failure mechanism of transformer LCT 2 The inspector reviewed engineering evaluation S-2-E130-EEE-0157 11 Failure Analysis of ITE 4160 delta/240Y/139 volt transformer (LCT-2F)
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The results of the evaluation are:
- No evidence of a cause of failure was attributable to the design or construction of the transforme Two likely possibilities may have caused the failur. Physical damage to the coil since the failure was concentrated on the outside surfac. An over-voltage transient or impulse failure since the failure appeared to be concentrated near a line termina The licensee concurred with the contractor's analysis which concluded that the failure is not generic to this transformer desig The inspector considers this item closed.
I
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(Closed)
Unresolved Item (50-311/86-29-02); Determine the failure mechanism of transformer SPT 2 The inspector reviewed engineering eva 1 uat ion S-C-E170-ESE-0676, 11 Failure Analysis of Station Transformer No. 22 11 *
The results of the evaluation ar It was determined that the transformer failed as a result of a turn-to-turn fault on the high voltage 11A 11 phase windin Two possible reasons are listed belo. A deterioration in turn insulation due to stresses experienced by the transformer during its operation over tim. A localized weak spot in the turn insulation which could not sustain either thermal or mechanical stresses induced when the LCT 2F transformer faile The licensee concurred with the contractor's analysis which concluded that the fault was not due to inadequate design or misapplication of the transforme The inspector considers this item close (Closed)
Unresolved Item (50-311/86-29-03); This item was open pending NRC review of the engineering evaluation of the transformer failure even The inspector reviewed the licensee's analyses performed, with the aid of expert contractors, and concluded the following:
- No generic problems were identified and therefore no design changes were recommende The licensee has increased their inspection and sampling requirements for all oil filled transformers, both active and spare Procedure MP2. 4, 11 Mi see 11 aneous Dry Type Power Transformer Maintenance 11 has been modified to clarify inspection requirements and include vacuum cleaning on a refueling basi The inspector considers this item close (Closed)
Unresolved Item (50-311/86-34-01); This item was open pending NRC Licensing and Region I review of weld overlay repairs for containment fan coil unit service water pipin A meeting between NRC and the licensee was held on December 4, 1986 to resolve the issu The licensee provided a written response to NRC questions resulting from this meetin NRC Licensing subsequently approved the weld overlay process as a temporary repai The licensee has
committed to replace the affected pipe at the next refueling outag A Region I specialist inspector observed the weld overlay process (see Combined Inspection 272/86-32; 311/86-36) and no further concerns were identifie The service water pipe replacement will be reviewed during routine NRC followup of outage activitie This item is close.
Operational Safety Verification 3.1 Documents Reviewed Selected Operators* Logs Senior Shift Supervisor 1s (SSS) Log Jumper Log Radioactive Waste Release Permits (liquid & gaseous)
Selected Radiation Work Permits (RWP)
Selected Chemistry Logs Selected Tagouts Health Physics Watch Log Refueling Procedure Firewatch and Security Post Orders 3.2 The inspector conducted routine entries into the protected areas of the plants, including the control rooms, Auxiliary Building, fuel buildings, and containments (when access is possible).
During the inspection activities, di~cussions were held with operators, technicians (HP & I&C), mechanics, firewatch and security personnel, supervisors, and plant managemen The inspections were conducted in accordance with NRC Inspection Procedures 71707, 71709, 71710, and 71881 and affirmed the licensee 1 s commitments and compliance with 10 CFR Technical Specifications, License Conditions, and Administrative Procedure.2.l Engineered Safety Feature (ESF) System Walkdown:
The inspectors verified the operability of the Unit 2 diesel generators by performing a walkdown of accessible portions of the system to confirm that system lineup procedures match plant drawings and the as-built configuratio This ESF system walkdown was also conducted to identify equipment conditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked, as appropriat No violations were identified.
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Plant Operational Activities:
The inspector observed/reviewed selected phases of the units
operation to determine compliance with the NRC 1 s regulation The inspector determined that the areas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personne The following are noteworthy areas the inspector researched in depth:
3. Unit 1 On October 2, 1987, Unit 1 was shutd6wn for a 63-day refueling outage and 10 year inservice inspectio During the plant shutdown, with rod insertion in progress and the reactor subcritical, a reactor trip occurred when the source range nuclear instrumentation automatically reinstated at P-6, as designe Source range channel N-31 failed high when the detector high voltage was energized causing a 1 of 2 source range tri The plant responded as designe The source range detector housing was found to have condensation built up in it causing breakdown of the polyethylene insulation surrounding the detecto The insulation breakdown resulted in an increase in leakage current between the detector and its housin The detector was replaced with a newer design which has a drainage port to alleviate buildup of water in the hou~ing. The failed channel was subsequently tested and placed in servic The inspector discussed this issue with the system engineer and I&C supervisor to ascertain the condition of the other source range detectors in the Salem unit Records indicate that the other detectors are of the old design and N-31 is the first to be installed with the new detector desig The licensee has two spare detectors of the new design on site and plans to replace the old detectors when necessar The*
inspector also reviewed the testing and verification performed on the new detector prior to its return to service, and attended two station operations review committee (SORC) meetings in which this event was discusse The inspector has no further question On October 9, 1987, the licensee experienced a spill of approximately 900 gallons from No. 13 steam generator primary side manwa The spill occurred when workers were removing the manway diaphrag As the workers pulled the diaphragm they heard an inrush of air, began to exit the area, and then observed an outrush of wate Eight personnel experienced various levels of contamination and were subsequently successfully decontaminate *
The licensee had recently changed the operating procedures involving draining the reactor vessel to the mid-loop level in accordance with a recent Westinghouse evaluation prompted by NRC Generic Letter 87-1 See Section 9 of this report for followup to the Generic Letter and this inciden During the containment cleanup effort a contractor radiation control technician picked up a hot particle from contact with a mop handl The exposure is currently under evaluation and was limited to the upper arm and shoulder are NRC radiation specialists were onsite the week of October 19 (Combined Inspection 50-272/87-30; 50-311/87-31) to followup on the radiological aspects of the spil On October 9, 1987, the licensee identified a security
.event involving a site maintenance supervisor who tailgated through a vital area door into the auxiliary buildin The individual's destination was the Unit 1 containment where he was performing work on an approved work orde The licensee determined that the individual also gained access to the auxiliary building and Unit 1 containment on the previous night (October 8) by tailgatin The individual 1 s vital access authorization had been administratively downgraded to protected area acceis in April, 1987 by his department management, and was not subsequently upgraded for the present Unit 1 outag The inspector discussed this event with the senior security supervisor and was informed that the individual 1 s site access was terminated by security pending resolution by the individual 1 s managemen The individual was counseled by his management concerning his responsibilities in the security area and was required to complete General Employee Training (GET) and Radiation Worker Training prior to his access being reinstate The inspectors had no further question On October 16, 1987, while filling the reactor cavity, the licensee identified an unacceptable cavity leak rate (approximately 5 gpm at 9 ft., and 10 gpm at 23 ft.).
The inspectors observed portions of the licensee's troubleshooting effort, held discussions with licensee management and staff, and attended licensee meetings concerning identification and resolution of the lea The licensee's investigation included: removing and resealing the sand plug and nuclear instrumentation covers, verifying reactor cavity seal leak off line integrity and isolation, in-place inspection of the inflatable seal and sealing
surfaces, and inspection of the cavity line Leak rates were also measured while varying the inflatable seal internal pressur The results indicated that the leak was at the seal because at a low internal pressure, the cavity leak rate increased and a higher internal pressure decreased but did not stop the cavity leak.. The licensee then removed the seal and performed a detailed inspection of the seal, cavity and vessel sealing surface The licensee determined that a carbon steel band welded to the top portion of the cavity liner (serving as a concrete form during construction) had begun to separate from the liner, resulting in an uneven sealing surface. This defor-mation could not be seen with the seal in plac A new seal was installed and pressurized and RTV sealant was applied to both sides of the sealing seam The cavity was then flooded to the 23 ft. level with no leakage detecte The use of RTV was evaluated and approved as a temporary repair for this refueling outag The licensee is evaluating permanent repairs to the carbon steel ban Also, Unit 2 will be inspected to determine the condition of the carbon steel band during the present outag The inspectors will follow the licensee's evaluation and repair of the band(s).
The inspectors had no further questions at this tim On October 26, 1987, core unload was complete The inspector observed fuel.movements in both the reactor cavity and spent fuel pi The inspector verified that approved procedures were in place and the required refueling controls were in effec On October 30, 1987, the lower internals were removed from the reactor vessel and placed on the support stand to facilitate programmed and remote (PAR) inspection of the reactor vessel as part of the 10 year inservice inspection (ISI).
The inspector witnessed fit-up and latching of the lower internals lifting rig via an underwater camera and remote TV monito Eddy current testing of No. 13 steam generator tubes was completed during the report perio The testing identified 3 defective tubes (Category C2), and one tube with a leaking plug which will require repai The licensee also identified degradation of the feedwater ring 11J 11 nozzles in nos. 12 and 14 steam generators and has decided to replace all J-tubes in the four steam generator The present J-tubes are carbon stee The replacement J-tubes will be inconel with a carbon steel colla An NRC Region I inspector
- was onsite during the week of October 26, 1987 to review licensee actions with regard to steam generator ISI results, repairs, sludge lancing, et Refer to NRC Inspection Report 50-272/87-29 for detail No violations were identifie.3.2 Unit 2 Unit 2 began this inspection period operating at 100%
reactor power and maintained steady state condition On October 22, 1987, the licensee declared all three Unit 2 diesel generators inoperable due to lack of breaker coordination causing inoperability of one or more of the diesels under certain postulated accident condition This resulted from an apparent improper setting of the diesel generator output breaker overcurrent protection relay The licensee entered Technical Specification (T.S.) 3.0.3, requiring correction of the problem within one hour or be in hot standby within the next six hour The resident inspectors were notified and the licensee made an ENS cal The licensee reset the diesel breaker relays to the proper setting, performed the relay verification test and diesel operability run, and exited T.S. 3. after two diesels were declared operabl The third diesel was subsequently declared operable following relay resetting and testin The resident inspector reviewed minor design change 25M-0487 which delineated the reason for the breaker relay setpoint change, the safety evaluation, and the method for incorporating the relay setting change The inspector then witnessed the changing of the relays in all three diesel generator The inspector also reviewed surveillance procedure SP(0)4.8.l.l.2, 11 Electrical Power Systems - Emergency Diesels" and witnessed the testing of 2A and 2B diesel generator The diesels were operated during the surveillance at 2700 amps which is 50 amps above the maximum diesel loading during the most severe accident condition Breaker coordination concerns were previously identified during an NRC Appendix R inspection in September, 198 The licensee had committed to provide written justification for continued operation of Unit 2 on October 23, 198 Following the emergency diesel generator 4 KV switchgear problems on October 22, the licensee notified NRC Region I on October 23, 1987 that additional uncertainties existed regarding the onsite
electrical distribution system design, which prevented reaching a comprehensive justification for continued operatio The problems included breaker coordination, seismic interaction and outstanding design basis recovery evaluation The licensee commenced a voluntary shutdown of Unit 2 on October 2 A four to six week outage was estimated to complete planned electrical analyses and projected modification The licensee has committed to provide to NRC an analysis of the onsite ~lectrical distribution system including proper electrical breaker coordination prior to restart of either uni During the shutdown, the licensee plans to (1) make repairs to the No. 24 steam generator, which had a 50 gallon per day primary to secondary leak prior to the plant shutdown; (2) replace the control rod drive mechanism (CROM) vent fans; and (3) other selected maintenanc The inspectors had no further questions at this tim During the tagging of diesel generator breakers for the above mentioned modifications, a licensee-identified incident occurred which resulted in the failure to retest the lA diesel output breake Therefore, lA diesel operability was not verified, resulting in less than the required number of diesels being operable during fuel movemen An equipment operator (EO) was sent to tag out 2A diesel (Unit 2) in order to reset the overcurrent rela He went to lA diesel (Unit 1) and racked-out the breaker, thinking he was at 2A diese When the EO began the tag out an alarm sounded in the Unit 1 control roo The spare control room operator (CO)
went to the breaker room and, upon meeting the EO and discussing the alarm, the EO realized that an error was made and returned the breaker to the operable positio The EO told his supervisor, the Unit 2 Senior Reactor Operator (SRO), about the inciden The SRO verified that two other diesels were still operable on Unit He thought that the resetting of the relays and the fact that Unit 2 was in an LCO took precedence over testing of the lA diesel, and he continued to address the relay setting evolutio In the meantime, the Unit 1 control room operators thought their supervisor had been informed and also recognized that two diesels were still operable (the minimum number for fuel movement).
However, the subsequent shift turnover
- failed to address the lA diesel breaker not being tested after being racked down and racked back i The next shift tagged out lB diesel generator for planned maintenance and unknowingly made two diesels inoperable (lA and lB) which is a violation of Technical Specifications for the refuelin operations in progress at the time:
This condition remained in effect until the CO, who was involved in the lA breaker mistake, reported for work the next day and asked how the lA breaker was tested prior to lB diesel being taken out of servic The scenario was then identified and the NRC was informe Subsequent testing of lA breaker was satisfactory, indicating that if needed, the lA diesel would have performed as designe As a result of the licensee's investigation of this event, several actions are being addresse.
Wrong train - wrong unit has been identified to training to reiterate the steps in place at this time to prevent this type of proble.
All personnel directly involved have been counseled and a disciplinary letter has been
- placed in their fil.
The seriousness of the matter has been discussed in the site newsletter to all operator The retesting of safety-related equipment has not been an identified problem at Salem Station, however the wrong train - wrong unit incidents have occurred in the pas The last such occurrence was three years ag The inspector concluded that this event constitutes a licensee-identified violation of Technical Specifications, for which adequate corrective action has been take (NV 272/87-28-01)
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At the end of this report period the Unit was in Mode 5 with the reactor coolant system drained and No. 24 steam generator opened to facilitate inspection and repair of the primary to secondary lea.3.3 Both Units On October 13, 1987, the licensee submitted to the NRC a request for discretionary enforcement relative to removing one of two offsite 500 KV power sources from service to enable the degassing of the No. 1 station power transformer oi The licensee requested an
additional three day period beyond the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> technical specification action statement (TSAS) to restore both offsite power sources to operable statu However, with Unit 1 in a refueling outage and the licensee's removal of Unit 2 from service to verify electrical configuration control, the TSAS does not apply and the discretionary enforcement was not neede The degassing of the transformer oil is scheduled to begin on November Four occasions of firewatches being found asleep occurred during the report perio Three were licensee identified and one was NRC identifie In each case, the firewatch was immediately relieved of his duties, was subjected to fitness for duty testing, and employment was terminate These firewatches were established in response to breaker coordination concerns identified during a previous NRC fire protection inspectio The inspector discussed the circumstances of each occurrence and license corrective actions with the senior fire protection superviso In each case, it was established that the sleeping watches had not missed any required round Nevertheless, corrective actions included putting more supervisors in the field, retraining firewatches, having security guards verify firewatch alertness during their rounds, and relieving the firewatches more ofte The licensee is in the process of installing additional detection in the auxiliary building to reduce the number of firewatches neede The inspector has no further questions at this tim On October 26, 1987, a fireman at the Vineland, New Jersey Fire Company, about 25 miles from the Salem plants, received a telephone call from a male who stated "there is a bomb on the Salem Plant.
The fireman notified the police department in Vineland, who, after an unsuccessful attempt to contact the licensee, notified the NRC Operations Center in Bethesda, Marylan The Headquarter's Operations Officer (HOO) notified the licensee's Senior Shift Supervisor (SSS) of the threa The SSS declared an unusual event in accordance with emergency plan procedure The senior security supervisor directed a search of the vital areas of each plant, even though the threat was assessed to have low credibilit The search proved negativ The resident inspector was notified of the bomb threat by the licensee and followed licensee action to the conclusion of the unusual even The resident inspector had discussions with licensee management concerning the police
- departments difficulty in contacting the licensee directl The licensee is inquiring into the matter and will provide the resident inspector with a response when their inquiry is complet No violations were identifie.
Maintenance Observations In conjunction with Unit 1 refueling activities the inspectors reviewed portions of the following outage design change packages (including the safety evaluations), attended selected SORC and planning meetings relative to these items, and witnessed portions of continuing work in the fiel The inspectors verified that the activities were performed in accordance with approved procedures and in compliance with NRC regulations and recognized codes and standard The inspectors also verified that the replacement parts and quality control utilized on the repairs were in compliance with the licensee 1 s quality assurance (QA) progra The inspectors plan to follow the below listed design changes through testing and final acceptance by the license Design Change Package lEC-2230 lEC-1890 Description Removal of resistance temperature detector (RTD) bypass loops and installation of Combustion Engineering in-line narrow range reactor coolant system (RCS) temperature measurement system, including installation of electronics to the reactor protection syste Status:
The bypass loop piping for all four loops has been remove Piping preparations for installation of new thermowells and RTD 1 s in each loop is in progres Appendix R modification to install wide range temperature indication for RCS loops nos. 12 and 13 on the hot shutdown pane Status:
Cabling installation is complete and termination is in progres Post-modification testing is scheduled to be complete by November lEC-2173 lEC-2261 lEC-2232 lEC-1076
Install ATWS Mitigation System Actuation Circuitry (AMSAC)
Status:
Hardware installation is complet Acceptance testing is in progres The inspector observed field wiring and continuity testing of circuit board components while installation was in proces Modifications to 64 safety related pipe support U-bolt anchor assemblies 3 11 diameter and abov Status:
Installation of U-bolt hangers is approximately 60% complet The inspectors witnessed in-progress field welding and NOE testing in the containment and auxiliary buildin Remove existing flux thimbles and top mounted thermocouples, install bottom mounted ther-mocouples and associated electronic Status:
The old flux thimbles are remove Removal of thermocouple columns is complete and preparations for welding the caps is in progres The inspector witnessed portions of the flux thimble removal, the cutting evolution and disposal of the cutup wast Upgrade 210 pipe supports of various systems throughout the plan Status:
Installation is approximately 60% complet On October 20, a quality control stop work order was issued due to poor work practices by the contracto More supervisors were placed on the job and the work was changed from two shifts to one shif On October 21 the resident inspectors received an allegation by two con-tractor foremen that someone was damaging system
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components at their worksite in order to get the contractor removed from sit The NRC requested license~ followup of the allegation and the inspectors are monitoring the licensee's investigatio No violations were identifie.
Surveillance Observations During this inspection period, the inspector reviewed in-progress surveillance testing as well as completed surveillance package The inspector verified that the surveillances were performed in accordance with licensee approved procedures and NRC regulation The inspector also verified that the instruments used were within calibration tolerances and that qualified technicians performed the surveillance The following surveillances were reviewed:
Unit 1 Procedure lPD-16.4.032 lPD-16.4.010 lPD-16.4.007 lPD-16.4.005 lPD-16.4.011 lPD-16.4.006 SP(0)4. Unit 2 (OP)II-17. SP(0)4.8.1. Description Source Range Channel 1N31 Discriminator and Plateau Voltage Adjustments Detector Installation Procedure Source and Intermediate Range Detector Electrical Tests Triaxial Cable Tests Detector Removal Procedure Detector Pre-Installation Checks Refueling Operation - Containment Isolation Auxiliary Building Ventilation Operation Electrical Power Systems - Emergency Diesel Generators No violations were identified.
- 16 Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special report The review included the following:
inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report informatio The following periodic reports were reviewed:
Unit 1 Monthly Operating Report - September 1987 Unit 2 Monthly Operating Report - September_ 1987 In addition, the inspector reviewed:
Unit 1 Special Report 87-7; This report delineates the licensee's actions after a fire post indicator valve (1FP246) was broken when hit by a pile driver cran The fire pumps were turned off and a watch was stationed to restore them if neede The pumps were turned off to limit the loss of water until the post indicator valve was repaire After the valve was repaired the system was returned to autom~tic. The inspector had no further questions about this repor Special Report 87-8; This report delineates the licensee's actions with regard to a service water leak in No. 11 fan cooler uni The leak was discovered during a hydrostatic test being performed under maintenance procedure M9-1HP-SW1 The leak was subsequently repaired using approved welding procedures and standard The inspector had no further questions about this repor Unit 2 Special Report 87-4; This report delineates the licensee's actions regarding the inadvertent lifting of the pressurizer overpressure protection system (POPS) while in Mode The inspector noted that the event was due to personnel error when a control room operator failed to control plant pressure during reactor coolant system fill and vent operation The tolerance band for operation is very close 325-375 psi Although no safety parameters were exceeded during this event, the operator was counseled and a letter from operations management was issued to all licensed operators regarding the seriousness of the even The inspector had no further questions on the repor No violations were identified.
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Licensee Event Report (LER) Followup The inspector reviewed the following LERs to determine that reportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification Unit 1 87-006-01 87-011 Unit 2 87-012 87-013 Both Trains of High Head SI Declared Inoperable -.0.5 entere This LER involved both trains of high head SI being inoperable due to an unattached spring charging motor on No. 11 centrifugal charging pump (CCP) 4 KV circuit breaker with the redundant No. 12 CCP being tagged out for maintenanc The original LER was discussed in Combined Inspection 272/87-19; 311/87-2 This supplement identifies changes made to Maintenance Procedure M3D 11 4KV and 13 KV Magne-Blast Circuit Breaker Inspection and Test
to preclude the mounting bolts from loosening and to facilitate inspection of the bolt The inspector had no further questions in this area.
Potentially Inadequate Breaker Coordinatio This issue was discussed in Combined Inspection 272/87-25; 311/87-27 and resident, regional, and headquarter 1 s inspectors are continuing to follow the licensee's progres RHR Pump Room Flood Curb Missing Due to Personnel Error This LER discusses a flood curb installed at the entrance of the auxiliary building sump tank room to prevent flooding of both RHR pump rooms in the event of a moderate energy line brea The curb was removed at sometime in the past and was not replace The exact time and reason for the removal could not be identifie The licensee has replaced the curb and labeled it 11 Do Not Remove - Flood Barrier.
The inspector attended two SORC meetings where this item was discussed, reviewed the licensee's corrective actions and had no further question Surveillance for Diesel Generator Fuel Oil Transfer Pump Missed due to Inadequate Procedural Control
L
This LER identifies a missed Technical Specification (TS)
surveillance due to a work order discrepancy involving the component identification (I.D.) for only one of two pumps being specified, but the work order delineating completion of a surveillance procedure which tests both pump Operations performed the test on the No. 21 DIG fuel oil transfer pump assuming that a separate work order would be issued for No. 22 pum Partial completion of operations procedures is permitted to facilitate testing if only one redundant component needs testin The planning department did not issue a second work order thinking that operations would complete the specified procedure on both pump To prevent recurrence of this problem, the
"component I.D.
11 section of work orders has been changed to "Functional Equipment Group", allowing for more than one component to be liste The licensee reviewed surveillance records to determine whether this discrepancy in the work order form could have resulted in other surveillances being missed and concluded that this was an isolated cas The inspector attended the station operations review committee (SORC) meeting where this item was reviewed, discussed the problem with operations and planning personnel, and reviewed the licensee 1 s corrective action The inspector concluded that this event constituted a licensee-identified violation of TS surveillance requirements for which adequate corrective action was take (NV 311/87-30-01)
The following relates to LER 1s that were written five and six years ag Recent reviews have determined that the issues related to these LER 1 s have not been formally close The inspector reviewed the licensee 1 s findings with regard to the LER 1 s and concludes the followin LER 81-71, 81-79, 82-19, 82-31, 82-63,82-132 and 82-147 were written because of spurious safety injection signals being generated by the safeguards equipment control (SEC) system under certain condition The licensee finally contacted the vendor who suggested adding resistors to boost the voltage levels within the SE The inspector reviewed DCR 1 s lEC-1377 and 2EC-1378 for Units 1 and 2, respectively, which added the resistor Since the addition of the resistors no similar instances have been identifie The inspector considers this matter close LER 82-63 and 82-6 The licensee performed an evaluation (S-l-R600-CEE-0125-RO) and determined that the problem (differential readings on two pressurizer level channels) was of a hydraulic nature rather than electrica Subsequently it was
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discovered that hydraulic lines 787-HD and 785-HD were installed incorrectl The licensee generated deficiency reports IC-82-1622 and IC-82-1623 and the lines were reinstalled correctl Since the correction no similar events have been observe The inspector considers this item close Supplemental LER 86-009-0 This supplemental LER delineates more detail about a reactor trip on Unit 2 (from 75%), due to the failure of station transformer The LER gives the results of the engineering evaluations (S-C-El70-ESE-0676 and S-C-E-130-EEE-0157) concerning the failure mechanism of the failed transformer The trip event was documented in NRC combined inspection report 50-272/86-24 and 50-311/86-29 and special inspection report 50-311/86-2 All previously opened items regarding this event were closed in paragraph 2 of this repor The inspector considers this LER close.
Appendix R Followup As a result of the NRC Appendix R inspection performed the week of September 14 (50-311/87-29) which identified fire protection (FP)
program deficiencies, the licensee committed, via letter dated September 18, 1987, to establish compensatory measure The resident inspectors verified the implementation of fire watches and the installation of smoke detectors in the common ventilation ducting for both units' chemical and volume control system holdup tank (CVCS HUT)
room The inspectors are also continuing to monitor the fire protection department's progress in reducing outstanding work orders for fire protection deficiencie Except for the fire watch discrepancies detailed in paragraph 3.3.3 of this report, the inspectors identified no further problems in this are.
Generic Letter 87-12, (Dated July 9, 1987)
11 Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Fi 11 ed This letter was sent to all licensees of operating PWR's and its intent was to solicit the licensee's responses to nine considerations dealing with the loss of RHR when the RCS is partially fille The inspector has reviewed the licensee's response (dated September 23, 1987) to this letter and has determined the followin All nine considerations were addressed in detail and the licensee made several changes to their methodology of operation due to a Westinghouse report dealing with this issue and the licensee's assessment of their current method of operatio In the licensee's response they delineated the below listed changes to operation The inspector verified the licensee's timely response to the generic issue, and the implementation of the licensee's committed action NRC Licensing will review the technical content of the licensee's response at a later dat,_-
- Revise the existing operating procedure for draindown to provide more information to operators while draining the RCS to a partially filled conditio The procedure is also to address an administrative control prohibiting draindown of the RCS until the I-131 concentration is less than 0.1 micro Ci/m The inspector has confirmed this procedure chang Develop an operating procedure to provide guidance during operation with the RCS partially fille Revise minimum required level in RCS hot leg to prevent vortexing and air entrainment to be consistent with the level required to prevent spilling due to removal of the steam generator manwa The inspector had verified the existence of this procedure, however as detailed below, problems with the adequacy of this procedure were noted following a reactor coolant spill on October 9,198 Develop an abnormal operating procedure to provide improved guidance during a loss of RHR event while in the partially filled RCS condition, including contingency for isolation of containment if a release of RCS inventory to containment is encountere The inspector has verified the existence of this procedure.
Provide guidance for installation, operation and maintenance of tygon tubing level indicatio The inspector has verified such guidanc Address training on potential for RCS perturbations due to actions performed whiJe the plant is in mid loop operatio The inspector has verified that all operators have been trained and has attended one of the training session Instruct control room personnel - the inspector has verified written correspondence pertaining to mid loop operation has been disseminated in the control roo Incorporate into operation training - scheduled for April of 198 Instructions provided to other affected department The inspector has verified that supervisors from other station departments which may be affected by mid loop operation have been instructe On October 9, 1987, the licensee drained the Unit 1 reactor to the mid-loop level for work in the steam generators (SIG) which was to be performed during the refueling outag However, a reactor coolant spill occurred when the diaphragm on No. 13 SIG was removed (See operations Section 3 for more detail).
Subsequent evaluations of the incident identified the following problems:
1 The operating procedure instructed the operators to maintain level in the loop between 97 1 9 11 and 98 1 *
The procedure also contained a statement to maintain 98 1+/- 3 11 for tolerance purpose Because of the concern of the generic letter and the instructions the operators were given, they elected to stay higher in the tolerance band which was 98 1 3 11 *
This level does not fully drain the hot leg and introduce air into the loop to allow water to drain from the S/G 11 U11 tube The result was a partial draining of the loops and the spill of reactor coolant upon opening the S/G manwa There was a two inch discrepancy between the control room indication and the local tygon tube attached to one of the reactor coolant loop The operators opted to use the most conservative (in the high direction) reading which was the control roo Subsequent investigation showed that the tygon tubing was more accurat As a result of the spill the licensee changed the procedures to reflect the lessons learne The new procedures and instructions to operators were utilized to drain down Unit 2 on October 27, 1987 with no identified problem The licensee also cataloged the volume of water drained from the system for future reference during reactor drain-down operations.
The inspector had no further questions on this subjec Bulletin 87-01, Thinning of Pipe Walls in Nuclear Power Plants The NRC issued this bulletin to all operating nuclear power plants to solicit information concerning their programs for monitoring the thickness of pipe walls in high-energy single-phase and two-phase carbon steel piping system The licensee was asked to submit their reply under oath or affirmation within 60 days of the receipt of the bulleti The bulletin was dated July 9, 198 The licensee replied to the bulletin September 14, 198 The inspector confirmed that all the questions asked in the bulletin were answered by the license The licensee 1 s submittal was for the Salem and Hope Creek generating station After a review of the licensee 1 s submittal, the inspector concluded that the licensee responded in a timely manner and conformed with the bulletin 1 s reques This Bulletin is considered close.
Regional Meeting With Licensee Management 11.l On September 29, 1987, the licensee management and staff (Attendees included as Attachment A to this report) traveled to the Region I office and presented their current assessments and conclusions regarding the Unit 2 reactor vessel head leak and their proposed schedule for the repairs and replacement of service water piping at the Salem nuclear facilit **
The Unit 2 head leak was identified on August 7, 1987, and was discussed in NRC Inspection Report 50-311/87-2 The licensee's presentation material on this matter is included as Attachment B to this repor During this meeting several questions regarding the licensee's interface with the Westinghouse owners group, the testing of the failed reactor head penetration weld, and residual boric acid in the recessed areas of the head were proposed to the license The licensee proposed that when the answers to these questions are available another meeting with Region I management would be scheduled to discuss the result The discussions regarding the licensee's approach to the management of service water system leaks has been. an ongoing dialogue with licensee managemen The licensee presented a comprehensive program (details are included as Attachment C to this report) to deal with the service water pipe corrosion/
erosion problem NRC will continue to monitor the progress of this program through routine inspection and review of licensee report.2 Another meeting was held at Region I (Attendees included as Attachment D to this report) on November 3,1987, to discuss the licensee's program for review, evaluation and correction of potential problems related to electrical distribution system coordinatio As discussed in paragraph 3 of this report, the licensee voluntarily shutdown Salem Unit 2 on October 23, 1987, because of these electrical system uncertaintie Further, the licensee committed to resolve these issues to NRC's satisfaction prior to restart of either Salem Uni The licensee's presentation is included as Attachment E to this repor NRC will continue to closely follow the licensee's recovery activities during subsequent inspections, and a followup management meeting will be scheduled upon completion of the recovery progra.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and finding An exit interview was held with licensee management at the end of the reporting perio The licensee did not identify 2.790 materia NRC:
ATTACHMENT A LIST OF ATTENDEES SALEM MANAGEMENT MEETING - SEPTEMBER 29, 1987 W. Kane, Director, Division of Reactor Projects (DR)
W. Johnston, Director, Division of Reactor Safety (DRS)
E. Wenzinger, Chief, Projects Branch No. 2, DRP P. Swetland, Chief, Reactor Projects Section 2B, DRP T. Kenny, SRI, Salem R. Summers, Project Engineer, Reactor Projects Section 2B, DRP J. Durr, Acting Deputy Director, DRS J. Strosnitjer, Chief, Materials & Processes Section, DRS N. Blumberg, Chief, Operational Programs Section, DRS H. Gray, Senior Reactor Engineer, DRS H. Kaplan, Reactor Engineer, DRS E. Brown, Senior Mechanical Engineer, AEOD Licensee:
S. Miltengerger, Vice President Nuclear Operations J. Zupko, Jr., General Manager - Salem Operations J. Ronafalvy, Technical Department Manager -
Sal~m Operations F. Cielo, Project Manager - Service Water J. Rowey, Project Engineer - Service Water J. Owens, Project Engineer - Service Water C. Timm, Salem System Engineer G. Reggio, Station Licensing Engineer - Salem
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I ATTACHMENT B PSEG UNIT 2 - REACTOR VESSEL NO. 5 T/C COLUMN LEAK by and Carl Timm
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John P. Ronafalvy Tech. Mg Salem System Enginee
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AGENDA
- Overview of the Event
- Events Leading to Discovery
- Description of Project Team
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Organization
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Charter
- Details of the Deficiencies Found and Resolutions Used by Carl Timm
- Future Actions
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OVERVIEW On Friday, Aygust 7, 1987, Unit 2 was removed from service for a main power transformer replacemen During the following 19 days a leak on No. 5 thermocouple (T/C) column was discovered which caused boric acid solution contamination of carbon steel component The wet boric acid interacted with the carbon steel and removed metal from the closure head causing pitted area The T/C column was repaired with a split canopy ring arrangement, the boric acid was cleaned up, the closure head was repaired and 17 studs were removed, cleaned and reinstalle.5 man/rem was attributed to these activitie Unit 2 was returned to service on August 26, 1987 and has been on line since.
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EVENTS LEADING TO DISCOVERY Precursors Prior to Shutdown
- Airborne Activity Was Elevated in Containment
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Nobel Gas Activity at*9K cpm
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Particulate Activity Approximately 7*0K *cpm
- RCS Humidity; No Abnormal Trends
- RCS Leak Rate; No Abnormal Trends
.9
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SAL9J1UNIT2 LEAK RATES
Containment Sump
- Highest at.47 GPM
-- Identified Leakage
- Highest at.59 GPM
Unidentified Leakage - Highest at.29 GPM 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 July/ August 1987
- PR.OGRAM Operatio~s perform cc::>titai~rne~t i~spectic::>~
~he~ U~it is i~ Mode D~e to persiste~t airbc::>r~e activ-ity i~ the co~tairirnerit, a
team of erigirieers ~ere '\\.l.tilized to make q'\\.l.adrarit by quadrarit irispectiori of coritairirne~t duri~g this outage-O~ Sat'\\.l.rda.y, 8/8/87 the ir.i.spectiori ~as completed a.rid some boric acid ~as fourid ori CRDM shroud-The shroud a.rid irisula.tio~
~ere disassembled to rev-eal three piri hole leaks ori the #5 T/C col~rn~
lo~er cariopy seal ~eld, a~d the e~terit of the boric acid corita.rniria.tiori ori ca.rbc::>ri steel cornpc::>rierits-Boric acid had corit.a.rniria.ted.:
CRDM a.rid T/C col'\\.l.rnr. *
Cl49sure head
Studs a.rid riuts
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Clos'\\.l.re head a.rid v-essel gap area
CRDM v-eritila.tiori duct
Various s'\\.l.pport structures
CRDM V"e~t shro~d
Closure head iris'\\.l.la.tiori r
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PROJEC'T
'I'E~~M ORGANIZATION
Pro.:j ect Ger:i.eral team ~as formed by Manager of Salem_
Tech role Dept -
assigr:i.e in pro.:ject -
lead Other departmer:i.ts members
~ere matri~ed ir:i.to the pro.:ject team:
Mainter:i.ar:i.ce
Planning/Sched~ling
Radiatior:i. Protection
Corporate Er:i.gineering
Welding Er:i.gir:i.ee::r
ISI Gro~p-
{NDE services)
Westir:i.gho~se ~eldir: e~pe::rt a.rid head closure/vessel desigr: enginee::rir: e~pe::rt.:joir:i.ed tea QA and Of fsite Safety Revie~ provided dedicated persor:i.ne1 to the p::ro~ect to provide o~ersite perspective du::rir: activities
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9Y9TEM FNGINF.:Fn~
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I TECHNICAL J MANAGER
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I SCHEDULING/ l L~ANN~-~:_J RADIAlION PROTECTION l I L_ _ --
~ J OFFBITE SAFETY REVIEW SITE SEl'IVICES I
CORPORATE ENBINEERIN9 PROJECT TEAM ORBANIZATIDN
MATRIXEO N LINE
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TEAM CHARTER
- Develop a Recovery Plan
- Implement Plan
- Obtain All Available Industry Experience
- Control Internal Information Flow (Single Point of Contact)
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- Interface with Regulator and Other External Organizations as Single Point of Contact (e INPO and Vendors)
- Develop Future Action Plan
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SPLIT j /
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,l l\\CK SCREWS
- l\\pply NF:OUIRF. to 111'<1!'1 " Cont.act
~urfaceg l--.A I
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F~LE FIAt<<;E MARMON CLl\\MP
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.--- Apply Nl!OWBE to All Clmie Contact Surf aces *
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123, 24 & 25 were in region of worst crud build up on a flange.. Leaky Canopy and head
- 31 found clean
- 21 had worst deposit on stud shank because ran under washer and nut-~~c:~
- 17 found clean Cooling Duct Support Ring
-~--132 and #33 were not pulled because of inter ferente-betwee1 rN-ui::tb:~"-:tens ioner and h~'ad H 1'~ ~-~ u "'1 34 found clean Lifting Lug
- 54 checked as I reference and found clea X * Studs that were removed. cleaned and re-installe STATUS OF CLOSURE STUDS AFFECTED BY THE BORIC ACID
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CLEAN-UP OPERATIONS
- All Cleaning Done with Stainless Steel Brushes, Scotch Brite and Damp Rags
- Head and Flange Area Vacuumed
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DEPRESSIONS
- 11 Depressions Found
- Deepest 15/32"
- All in Area of Boric Acid Pile
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Location of pits and grooves
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RESOLUTION
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- Depression Cleaned
- Magnetic Particle Test
- Ultrasonic Test RESULTS
- UT Showed a Thickness of 7.6" in Unaffected Region
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- Evaluation Showed That Y.:l" Thickness Would Be *Required to_ Meet ASry1E Code
- Depressions Were Shown to Be Acceptable
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E-4
LEAKING THERMOCOUPLE COLUMN COUNTERBORE Thermocouple Column #5 Suffered Wastage in Its Counterbore Region
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RESOLUTION
- Counter Bore Cleaned
- Ultrasonic Test
- Evaluation RESULTS
- Same Transients as Applied to the Head Were Applied to the Counterbore Region
- Evaluation Showed Counterbore Region Acceptable
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REACTOR VESSEL STUDS
- Boric Acid Ran Down the Side of Reactor Head into the Flange Gap Area
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Gap nfE:P.MOCOUPLE__
COLti"MN I~
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RESOLUTION
- Clean and Fiberoptic Reactor Flange Gap Area
- Continue to Remove Studs Until One Stud on Each Side of Boric Acid Pile**-
ls Found to Be Free of Boric Acid
- Remove One Stud at a Time
- Stud Must Be Completely Free of Boric Acid After Cleaning
- Magnetic Particle Test Studs in Area of Pile RESULTS
- Reactor Vessel Flange Area Verified Free of Boric Acid
- A Total of 17 Studs Removed
- Mag Particle Test Performed on 5 Studs and Revealed No Reportable Indications
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CROM SHROUD AND DUCTWORK
- One Section of Shroud Had Boric Acid Deposits
- No. 23 CROM Ductwork Had Deposits on the Inside *
RESOLUTION
- Clean Shroud and Ductwork
- Inspect for Wastage from Boric Acid RESULTS
- Shroud Free of Boric Acid
- Ductwork Free of Boric Acid and No Evidence of Wastage on Shroud or Ductwork
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- WELD REPAIR
. * Split Canopy
- ASM E Code Section 111
- New Seal Pressure Boundary for RCS
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RETESTS
- Dye Penetrate Successful
- Hydrostatic Test to 2300 psig Successful
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FINAL INSPECTION The Following Areas Were Inspected and Were Free of Boric Acid
- CROM Counterbore in Head
- CROM Boric Acid Residue
- Closure Head
- Depression Area on Head
- Flange - Top/Between Studs/Nuts/ *** ~,..
Washers/Side Face
- CROM Vent Shroud and Internals
- CROM Vent Pipe Supports
- Cavity Seal Ledge
- CROM Duct Work at Stud #19
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~ Groove Between Flanges
- Shank and Threads of Studs Removed
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POTENTIAL FAILURE MECHANISMS
- Only a Postulation Can Be Made at This:
Time as to the Root Cause of the Canopy Seal Weld Failure and Subsequent Leaks
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- This Postulation Is Based en the Failure Analysis of Two Identical Welds from Another Plant t---
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Leaks Are Usually Associated with Purge Holes Which Are Used to Introduce Inert Gas to the Back Side of Weld
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WELD REPAIR FAILURE
- Occasionally Purge Holes Blow Out
- Purge Hole Weld Is Not Fused Properly When Wel*ded Over
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More Accurate Root Cause Analysis of #5 Thermocouple Column Seal Weld Failure Will Be Determined by Westinghouse After Core Exit Thermocouple Columns Are Removed During the Unit #2 4th Refueling Outage, April, 1988
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FUTURE ACTION * Install Viewing Ports on Unit 1 and 2 CROM Ventilation Sh.rouds
- PSE&G Will Seek.Westir:i_gt}o,~se O~ners Group Involvement *
- J
- PSE&G Will Remove All T/C Cano-Seal Connections
- Affected TIC Column Shall Receive Detailed Root-Cause-Analysis by Westinghouse
- Improve Administrative Controls
I oPIPIHG PROJECT ENGINEER J.ROWEY J. r.oHEZ I.OWENS A.OAKES O COMl'OiENT 5 o PROJECT PLAN 0 INTERFACE WITH s*w ENGllEERING O IOENTIFICATlllN OF STATION REOUIAEMENTS oSEilVICE WATER DR RESOLUTION o RESOLUTION OF SERVICE WATER TECHNICAL ISSUES OSON: INTERFACE lo£CHANICAL l-
- !£SIGNER I
I A.SMELTZER I
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WELDING
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C.FREW
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COATING I
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A. FRANC1£TTI I
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I l-MATERIALS I
I S.LICUO I
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COMPONENTS LATER PL~ING &
SC1£DULING II.EDWARDS SERVICE WATER PROJECT TEAM.
PROJECT MANAGER F.V. CIELO o PROJECT EMECuTION o CONTRACT AOHINISTRATION OPROJECT PLAN OEVELOPHENT o BUor.£ T OEVELOPHENT INSTALLATION ENGINEER J.OOATHER oRESOURCE PLANNING oCONFLICT RESOLUTION OPROJECT STATUS REPORTING PROJECT MAINTENANCE J. WEBSTER PROJECT LICENSING G. ROGGIO
ATTACHHENT I PROJECT.
TECH. SERVICE W.STRAUSIU.LER oPROJECT PLAN OEVELOl'HENT o SCHEOU..ING oCONSTRUCTABILITY REVIEWS o PROBLEM RESOLUTION oPROJECT PLAN OIDENTIFICATIOH OF PH REQUIREMENTS o IOENTJFICA TIOH OF LICENSIGN ISSUES OHAHAGE LICENSING c~s o SPECIAL PROJECTS OPllQJECT ST"TUS REPORTS OINTEllFACE WITH CONTRACTOR Ol'LIHollNG O INTEGRATION WITH OUTAGE PL-.alNG ODIRECT CONTRACTOR INSTALLATION ACTIVITIES OllEVELOP, IMPLEMENT AND EVALUATE TESTING PROGRAMS oRESOLV[ TESTING PROBl.EH5 o PERFORM FUNC TlllNAL AND SYSTEM TESTING o IDENTIFICATION OF MAINTAINABILITY CONSIDERATIONS OIHPUT TO REPLACEMENT PRIORITIZA TlllN OPROJECT PLAN STATION PLN.NING COST ENGlt£ERING OUALITY ASSUW.:E R. BE!fllNG K. QJZMA OOUTAGE PLANNING oESTIHATING oOUTAGE SCHEDULING oCOST REPORTING C.OOMANOSKY ooc REOUIREHENT o DA REOUIREMENT oPAQJECT PLAN oPRDJECT PLAN MATERIAL EXPEDITOR LATER HOPE REEK INTERFACE A. MOUOGILL OINTERFACE o LESSONS LEARNED
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L. LEITZ STRESS ANALYSIS K. HATl' PRDJECTIZED REPORllNG
MIURIX REPORTlt<
IEVEUP SNA.E (Q.A *. 11---a1 CRITERIA llEJITIFY IELDS IELD OVERLAYS r~----- -- - ---
1111 SNFl..E IELDS V/R I YR. ISi HYmOTEST MALVZE DATA
,...........,
DO
! lllJTff 116 j
.UNIT NO. I C.F.C.U. 3 316 S.S. SPOOL INSPECTIONVWELD OVERLAY
- -
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<U. T. H6EJ llDTIFY NE LOS SPOIL MTERIAL llSTALL ID
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NO SNIU NE LOS V/U. YR. ISi HYmOTEST MALVZE
- DATA r...........,
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UNIT NO. I C.FCU. 3 c/L C.S. SPOOL INSPECT ION/CHANGE-Cl.IT
/
NOTIFY ENGINEERING ENTER TSAS SERVICE WATER SYSTEM LEAK MANAGEMENT NO NOTIFY PLANNING PLAN/
SCHEDULE REPAIRS NO NO IMPLEMENT REPAIRS*
SAFETY RELATED SYSTEM LEAK ISOLATE LEAK ENTER TSAS PREPARE W/R NO
- UNDER APPROPRIATE MODES
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ATTACHMENT D LIST OF ATTENDEES SALEM MANAGEMENT MEETING NOVEMBER 3, 1987 NRC W. Kane, Director, Division of Reactor Projects (DRP)
W. Johnston, Director, Division of Reactor Safety (DRS)
S. Collins, Deputy Director, DRP J. Durr, Acting Deputy Director, DRS E. Wenzinger, Chief, Projects Branch No. 2, DRP P. Swetland, Chief, Reactor Projects Section 2B, DRP T. Kenny, SRI, Salem R. Summers, Project Engineer, Reator Projects Section 2B, DRP C. Anderson, Acting Chief, Engineering Branch, DRS T. Koshy, Reactor Engineer, DRS 0. Chopre, NRR D. Tondi, NRR Licensee C. McNeill, Jr., Senior Vice President - Nuclear J. Boettger, Assistant Vice President - Nuclear Engineering J. Zupko, Jr., General Manager - Salem Operations C. Johnson, General Manager - Nuclear Quality Assurance L. Reiter, General Manager - Licensing and Reliability L. Miller, Manager, Nuclear Engineering Services C. Lambert, Sciences Manager B. Preston, Manager, Licensing and Reliabililty State of New Jersey (Department of Environmental Protection)
D. Scott, Bureau Chief W. Cristafi, Nuclear Engineer
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ATTACHMENT E REGION I NRC MEETING TUESDAY, NOVEMBER 3, 1987 INTRODUCTION/SHUTDOWN STATUS I BREAKER RELAY COORDINATION PROGRAM II ELECTRICAL DESIGN REVIEW PROGRAM*
I SUMMARY c. A. MCNEILL (SR. VP-NUCLEAR)
L. K. MILLER (MANAGER-NUCLEAR ENGRG SERVICES)
L. K. MILLER c. A. MCNEILL
APP7
BREAKER RELAY COORDINATION OBJECT IVIE:
- ACHIEVE FULL BREAKER RELAY COORDINATION FOR IE EQUIPMENT ACHIEVED BY:
- MODIFICATIONS ANO ADJUSTMENTS
- FIELD VERIFICATION <WALKDOWN>
- INDEPENDENT VERIFICATION
- JUSTIFIABLE EXCEPTIONS
APP6
ELECTRICAL DESIGN REVIEW OBJECTIVE:
- COMPLETE THE DESIGN REVIEW OF SALEM'S ELECTRICAL SYSTEM TO ASSURE THAT ANY SIGNIFICANT SAFETY ISSUES ARE IDENTIFIED AND ADDRESSED ACHIEVED BY:
- PROCEDURE AND DOCUMENTATION REVIEW
- INDEPENDENT VERIFICATION OF THE ELECTRICAL SYSTEM STUDY <PHASE I>
- EVALUATION OF SYSTEM ANALYSIS <PHASE II>
- CONSIDERATION OF ADDITIONAL ELECTRICAL AREAS
APP5 ELECTRICAL CONFIGURATION CONTROL OBJECTIVE:
- ESTABLISH ELECTRICAL CONFIGURATION CONTROL FOR RELAY AND BREAKER SETTINGS ACHIEVED BY:
- ANALYSIS OF CURRENT ORGANIZATIONAL INTERFACES
- IMPLEMENTATION OF NEW CONFIGURATION PROCESS
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SALEM UNIT 2 RESTART ACTION PLAN ELECTRICAL DESIGN REVIEW SUMMARY PLAN ACTJVITY DESCRIPTION NOVEMBER 1 q97 I
DECEMBER 1q97 I
1qaa q
23
7
21 28 IJIFIMIAIMIJ I
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I l. ELECTRICAL OESJGN REVIEW
- PROCEDURE AND OOClHNTATION--------------------
- INJEPENDENT VERIFICATION--------------------------
- FIELD VERIFICATION WALKDOWN lPHASE lJ ----------
- RESOLVE BREAKER COCRJINATJ(J4 --------------------.
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- lDSIDERATION OF ADOITIOOAL ELECTRICAL--------
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AREAS FOR REVIEW
- SYSTEM ANALYSIS EVALUATION lPHASE lD ---------
t tR REVIEW t
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RNIJE OF RESTART RST<J