IR 05000354/1987011
| ML20214S729 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/28/1987 |
| From: | Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214S708 | List: |
| References | |
| 50-354-87-11, NUDOCS 8706090389 | |
| Download: ML20214S729 (18) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
050354-870420 Report No.
50-354/87-11 Docket 50-354 License NPF-57 Licensee:
Public Service Electric and Gas Company Facility:
Hope Creek Generating Station Conducted:
April 14, 1987 - May 11, 1987 Inspectors:
R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector R. R. Brady, Reactor Engineer r/o t f s"7 Approved:
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L.(orrholm, Chief,ProfectsSection28 Dat(e
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Inspection Summary:
Inspection on April 14, 1987 - May 11, 1987 (Inspection Report Number 50-354/87-11)
Areas Inspected:
Routine onsite resident inspection of the following areas:
followup on outstanding inspection items, operational safety verification, surveillance testing, maintenance activities, engineered safety feature system walkdown, security inspection, part 21 response, design change package review, and a core thermal power evaluation.
This inspection involved 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> by the inspectors.
Results: While housekeeping and cleanliness standards in general have been satisfactory, there have been several lapses to below satisfactory levels.
Section 9.0 of this report describes construction materials discovered in IE battery rooms which potentially threatened safety related equipment. Additional management attention is needed in this area to ensure cleanliness standards are maintained.
8706090389 S7052 PDR ADOCK 05000 54 G
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Details 1.
Persons Contacted Within this report period, interviews and discussions were conducted with Mr. R. Salvesen and members of the licensee management and staff and various contractor personnel as necessary to support inspection activity.
2.
Followup on Outstanding Inspection Items (Closed)
Inspector Follow Item (87-04-01); The analyses from the licensee and Brookhaven National Laboratory were completed and a comparison was made.
Results were in agreement.
This item is closed.
(Closed) Inspector Follow Item (87-08-05); During the previous inspection period, a deficiency in the licensee's strike contingency plan was identified relating to the control room shift rotation schedule and the number of hours an individual is allowed to work in a 7 day period.
The licensee has subsequently revised the contingency plan by adding a third shift of control room operators to the shift rotation.
This new rotation corrects the identified deficiency.
The inspector has no further questions and this item is closed.
3.
Operational Safety Verification 3.1 Inspection Activities On a daily basis throughout the report period, inspections were con-ducted to verify that the facility was operated safely and in con-formance with regulatory requirements.
The licensee's management control system was evaluated by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation, and review of facility records.
These evaluations included backshift inspections conducted on April 30 (3:00-6:00 a.m.), May 1 (12:01-1:05 a.m.) and Saturday, May 9 (12:40-4:40 p.m.).
The licensee's adherence to the radiological protection and security programs was also verified on a periodic basis.
These inspcction activities were conducted in accordance with NRC inspection procedures 71707, 71709 and 71881.
3.2 Inspection Findings and Significant plant Events The unit entered this report period at 100% power (not limited by the transmission network stability curves generated after the damage to the Keeney 500 KV transmission lines since Salem Unit 2 was off line for repairs).
The unit remained at the maximum
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allowable power levels throughout this report period except for short power reductions in order to perform maintenance or surveillance activities.
On April 14 at 6:12 p.m., the "A" reactor recirculation pump tripped while the reactor was operating at 100% power. A redundant reactivity control system surveillance test was in progress at the time.
Test meter leads were accidentally pulled from their proper position, resulting in the generation of the recirculation pump trip signal.
The "B" recirculation pump was manually run back and rods inserted to maintain the core within operating limits. The "A" recirculation pump was restarted and-the unit returned to 100% power later in the evening.
On April 17, the unit reduced power to approximately 90% power to stay within the transmission network stability curves as Salem Unit 2 was returned to operation.
By letter dated April 20, 1987, from NRC Region I to PSE&G, the licensee was granted a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> discretionary enforcement of Action Statement 80 (a) of Technical Specification Table 3.3.7.5-1, Primary Containment Hydrogen /0xygen Concentration Analyzer and Monitor. The action statemut required that the subject analyzer / monitor be made operable within 7 days or the plant be placed in hot shutdown and subsequently in cold shutdown. The seven day action period ter-minated at 9:37 a.m. on April 20, 1987. The relief request (from PSE&G to NRC Region I dated April 20,1987) became necessary due_to a series of problems encountered during routine surveillance on the subject instrumentation, and during the ensuing repairs which required emergency acquisition of replacement parts. A redundant channel of H2/02 concentration monitoring and post accident sampling system capabilities were available during the extended inoperability of the "B" channel. The affected instrumentation was declared operable at 3:50 a.m. on April 21, 1987.
On April 21 at 2:20 p.m., while placing the "B" reactor water cleanup system (RWCU) filter demineralizer (F/D) in hold and shifting to the "A" F/D, a sudden insurge of water stemming from the transition caused the RWCU system to isolate on high differential flow. The RWCU isolation valves shut in response to the NSSSS isolation signal. The RWCU pumps tripped on low flow. All systems responded as expected.
The isolation was reset and the RWCU system was restored to service at 9:55 p.m.
On April 29 at 2:00 p.m., the licensee shifted guard force contractors from Yoh to Wackenhut.
The transition occurred without inciden,
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On April 29, the IBEW union and Public Service Electric and Gas reached agreement on all terms in the proposed contract between the union and the company for the next two years. The proposal was voted on and accepted by the rank and file on April 30.
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On May 8 at 10:30 a.m., the "A" reactor protection system tripped which caused a reactor water cleanup isolation, recirculation sample line isolation, main steam line drain isolation, and a half scram signal. The "A" RPS bus tripped when its electronic protection assembly (EPA) sensed an undervoltage condition during a surveillance
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test which transferred infeed breakers on the "A" 4160 volt vital bus.
The "A" RPS bus was more sensitive than normal to electrical
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transients as the bus was being powered from its alternate power supply, 108491, due to maintenance on the "A" RPS MG set. All systems were returned to service.
No violations were identified.
4.
Surveillance Testing 4.1 Inspection Activity-
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The inspector performed detailed technical procedure reviews, witnessed in progress surveillance testing, and reviewed completed surveillance
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packages. The inspector verified that the surveillance tests were performed in accordance with Technical Specifications, licensee approved procedures, and NRC regulations. These inspection activities were conducted in accordance with NRC inspection procedure 61726.
4.2 Inspection Findings The following surveillance tests were reviewed, with portions witnessed by the inspector:
IC-FT.AB-033 Functional Test of SRV Acoustic Monitors
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IC-FT.SA-004 Functional Test of Redundant Reactivity
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Control System Channel 8 OP-IS.BJ-001 HPCI Inservice Test
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IC-FT.SE-014 Functional Test of APRM Channel B
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No violations were identified.
5.
Maintenance Activities 5.1 Inspection Activity The inspector observed selected maintenance activities on safety
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related equipment to ascertain that these activities were conducted
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in accordance with approved procedures, Technical Specifications, and appropriate industrial codes and standards. These inspections were conducted in accordance with NRC inspection procedure 62703, 5.2 Inspection Findings Portions of the following activities were observed by the inspector:
Work Order Procedure Description 87-03-27-017-3 MD.DC.ZZ.001
"D" Diesel Water Jacket Expansion Tank Level Calibration 87-03-27-022-0 MD.DC-ZZ.001 Calibrate "D" Diesel "A" Air Compressor Start /Stop Switch 87-03-27-023-8 MD.DC-ZZ.001 Calibrate "D" Diesel "B" Air Compressor Start /Stop Switch 87-02-25-005-5 IC-DC.ZZ-240 Temperature Indicating Device Calibration IKJTSHA 422A 87-04-14-224-1 IC.DC.ZZ.201 Diesel Generator A Lube Oil Strainer D/P Gage Calibration No violations were identified.
6.
Engineered Safety Feature (ESF) System Walkdown 6.1 Inspection Activity The inspectors independently verified the operability of selected ESF systems by performing a walkdown of accessible portions of the system to confirm that system lineup procedures match plant drawings and the as-built configuration.
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 appropriate.
This inspection was conducted in accordance with NRC inspection procedure 7171.
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6.2 Inspection Findings The low pressure coolant injection (LPCI) portion of the residual heat removal system (RHR) was inspected. The LPCI system is an emergency core cooling system that utilizes four electrically and physically independent loops to inject water from the suppression pool to the reactor vessel upon receipt of reactor vessel low level or drywell high pressure signals. A walkdown of the system did not identify any conditions adversely affecting system operability nor were any minor discrepancies identified that were not previously licensee identified. The inspectors verified that corrective actions were in progress for the minor discrepancies observed.
A review of the RHR system operating experience identified only one significant equipment failure.
During November, 1986 the
"B" RhR pump required replacement due to bearing failure caused by a series of minor water hammer / cavitation events over a period of time.
Additional details of this failure can be found in Inspection Report 354/87-01.
No violations were identified.
7.
Special Report Followup The licensee submitted the following special report during the inspection period.
This report was reyjewed for accuracy and timely submission.
Special Eeport 87-003 Containment Atmosphere H202 Analyzer / Monitor Inoperable for More Than 7 Days Special Report 87-003 describes the events which resulted in the licensee requesting discretionary enforcement to continue troubleshooting the primary containment H2/02 analyzer. Details of this troubleshooting activity are discussed in section 3.2 of this inspection report.
No violations were identified.
8.
Part 21 Report The licensee has been informed, by a Part 21 notification, of a potential problem concerning the diesel generator exciter units manufactured by Basler Electric.
The AC shutdown contactor component of the exciter unit may fail due to a defective 0-ring in the latch mechanism. The AC shutdown contactor vendor (Sprecher and Schuh) recom-mended solution is to replace the 0-rings.
The new 0-ring material has been tested by the Basler Electric qualification lab, and results indicate a life expectancy exceeding 40 years for the given design service conditions.
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The licensee has been in contact with the vendor and has ordered a new replacement AC shutdown contactor for each diesel generator.
The inspector reviewed the Part 21 report, and discussed the issue with the system engineer.
The inspector had no questions, however this item will be tracked as an inspector follow item to review the installation of the replacement AC shutdown contactors.
(87-11-01)
No violations were identified.
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9.
Design Change Package Review
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The inspector verified adequate implementation of five design change packages (DCP) identified during the independent design verification program conducted by Sargent & Lundy.
These five DCPs numbered 652, 653, 656, 658, and 7031 were initiated to direct post LOCA leakage from non-safety grade water systems away from safety related components. The inspector reviewed each DCP, verified the inplant installation, and was satisfied with the DCP implementation. During the verification inspection, the inspector located and reported to the senior nuclear shift supervisor stray sheet metal in a drip tray in the overhead of a IE battery room.
The operating shift investigated and determined that the identified drip tray contained seven pieces of scrap sheet metal. The shift made an extensive search of other 1E battery rooms and located 23 additional pieces of scrap sheet metal, a six foot 1" x 10" board, a white shirt, and a string of lights 20 feet long. All this extraneous material was removed. The inspector conducted spot check inspections throughout the plant and had no further questions.
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No violations were identified.
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10.
Security Inspection The inspector conducted an inspection to verify the control of locks, keys, and combinations used to control access to protected areas, vital areas, material access areas, and controlled access areas. Adequate protection for locks, keys, and combinations was evaluated due to the recent change in guard force contractor. The inspector reviewed the lock and key control system, the key cards, and held discussions with the lock and key custodian.
No unacceptable conditions were identified.
No violations were identified.
11. Core Thermal Power Evaluation An inspection was conducted to verify that the licensee's calculation used to determine core thermal power was correct and the procedure was technically adequate.
The inspector reviewed the following documents:
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RE-RA.ZZ " Core Thermal Power Evaluation"
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Core Performance Data Printouts (00-3)
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The inspector discussed the procedure with the reactor engineering group and was satisfied with the calculation results.
No violations were identified.
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12. Management Meeting
.e On April 24, 1987, a management meeting was held between NRC and PSE&G to discuss the results of the Engineering Department's review of engineering performance and plans for improvements. This meeting was held in response to the Hope Creek and Salem SALP board recommenda-tions in the outage (Engineering Support) functional area. The list of attendees and the presentation outline handout provided by the licensee are attached as enclosure 1 to this report.
13.
Exit Interview The inspectors met with Mr. C. Conner and other licensee personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspection activities.
Based on Region I review and discussions with the licer.see, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.
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Enclosure 1 Management Meeting on April 14, 1987
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List of Attendees Name Organization W. F. Kane NRC W. V. Johnston NRC R. M. Gallo NRC S. J. Collins NRC L. H. Bettenhausen NRC R. W. Borchardt NRC R. J. Summers NRC T. J. Kenny HRC C. A. McNeill PSE&G
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S. E. Miltenberger PSE&G
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T.' Boettger PSE&G R. A. Burricelli PSE&G C. W. Churchman PSE&G l
W. Cristali NJ D.E.P.
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ENCLOSURE 1
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I NRC REGION 1/PSE&G MEETING i
APRIL 24,1987
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t ENGINEERING PERFORMANCE AND RESPONSIVENESS
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AGENDA
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1. Opening Remarks 11. Opoortunities for improvement 111. PSE&G Engineering Effectiveness Improvement Assessment IV. Concluding Remarks
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ENCLOSURE 1
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NRC REGION 1/PSE&G MEETING 4/24/87
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OPPORTUNITIES FOR IMPROVEMENT
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WITHIN E&PB ORGANIZATION
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- Accountability and Responsibility
- Effectiveness and Productivity
- Cumbersome Processes
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- Past NRC Observations Selection of A.T. Kearney to Study E&PB
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ENCLOSURE 1 NRC REGION 1/PSE&G MEETING 4/24/87
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NRC OBSERVATIONS I
SALP Year - 1986
- The Engineering Staff Has Not Consistently Displayed the Initiative to identify and Resolve Technical Problems in the Field
- The Engineering Department Needs to improve Performance and Responsiveness
- The Licensee Demonstrated a Lack of Effective Followup to NRC Concerns
- The Licensee's Response Was incomplete... Records Were Not Complete Nor Properly Maintained SALP Year - 1985
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- The Licensee Has Not Exhibited... Initiative in Response to Concerns and Unresolved items identified by the NRC The Engineering Support Group Seems to Have a Different Set
of Priorities Than... the Station
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ENCLOSURE 1
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NRC REGION 1/PSE&G MEETING 4/24/87 ENGINEERING AND PLANT BETTERMENT DEPARTMENT JUNE 1986 ORGANIZATION CHART GENERAL MANAGER ENGINEERING AND PLANT BETTERMENT I
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ASSISTANT ASSISTANT GENERAL MANAGER GENERAL MANAGER PROJECT SERVICES PROJECT ENGINEERING I
MANAGER MANAGER MANAGER MANAGER MANAGER MANAGER MANAGER ENGINEERING AND NUCLEAR NUCLEAR NUCLEAR MANAGER PROJECT PLANT
'NSTALLATION PLANT BETTERMENT SYSTEMS ENGINEERING SYSTEMS ENGINEERING ENGINEERING ENGINEERING ENGINEERING MANAGEMENT
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CONTROLS SALEM HOPE CREEK DESIGN
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ENCLOSURE I
PSE&G NUCLEAR DEPARTMENT Engineering and Plant Betterment Finalreport
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A.T. Kearney September,1986
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ENCLOSURE 1
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NRC REGION 1/PSE&G MEETING 4/24/87 A.T. KEARNEY 9/86 REPORT RECOMMENDED THAT A DEDICATED PSE&G TEAM BE
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ESTABLISHED
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l lMPLEMENTATION TEAM LEADER
IORITY SENG
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PERFORMANCE DCR AND PROJECT ORGANIZATION MEASUREMENT TASK FORCE T9ACKING EFFECTIVENESS TASK FORCE TASK FORCE
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ENCLOSURE 1 NRC REGION 1/PSE&G MEETING 4/24/87 TECHNICAL PROGRAMMATIC ISSUES
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Department Programmatic sponsors have been functionally identified in the organizational structure to provide better accountability for the following technical issues:
- Environmental Design Criteria
- Stress Analysis and Support Evaluation Program
- Separation Criteria Program
- Radwaste Systems
- Seismic II/I Program
- ALARA/ Shielding Design Criteria PVORT Program e
- Set Point Program
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ENCLOSURE 1 NRC REGION 1/PSE&G MEETING 4/24/87
- Design Change Process Improvements
- Project Controls improvements
- Performance Measurement / Indicators
- Additional Features for Improvement
- Design Basis Documentation
- Formulated Technical Standards,
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Procedures, Writer's Guide
- Training
- Engineering Records Management
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