IR 05000354/1986046
| ML20215M770 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/10/1986 |
| From: | Briggs L, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215M765 | List: |
| References | |
| 50-354-86-46, NUDOCS 8611030277 | |
| Download: ML20215M770 (13) | |
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b 9-U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-354/86-46 Docket No.
50-354 License No. NPF-57 Licensee: Public Service Electric and Gas Company 80 Park Plaza Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station, Unit 1 Inspection At: Hancocks Bridge, New Jersey Inspection Conducted: September 11-19, 1986 Inspectors:
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L. Wink, Reactor Engineer
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D. Florek, LeMleactor Engineer
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'date Test Programs Section, 08, DRS Inspection Summary:
Inspection on September 11-19, 1986 (Inspection Report No. 50-354/86-46)
Areas Inspected:
Routine, unannounced inspection of the overall power ascension test program including procedure reviews, test witnessing and test results evaluation, independent measurements and verifications, QA/QC inter-face, tours of the facility and follow-up of licensee action on previous inspection findings.
Results: No violations were identified.
Note:
For Acronyms not defined refer to NUREG 0544 " Handbook of Acronyms and Initialisms."
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Details 1.0 Persons Contacted Public Service Electric and Gas Company (PSE&G)
- R. Beckwith, Station Licensing Engineer
- G. Connor, Operations Manager G. Daves, Senior Engineer, Operations P. Dempsey, Shift Test Coordinator M. Dick, Startup Test Engineer
- M. Farshon, Power Ascension Manager -
B. Forward, Power Ascension Administrative Coordinator
- A. Giardino, Manager, Station QA
- R. Griffith Sr., Principal Engineer - QA R. Hovey, Senior Nuclear Shift Supervisor E. Riley, Senior Nuclear Shift Supervisor
- R. Salvesen, General Manager, Hope Creek Operations W. Schell, Power Ascension Technical Director U.S. Nuclear Regulatory Commission D.'Allsopp, Resident Inspector T. Koshy, Reactor Engineer
- R. Borchardt, Senior Resident Inspector
- Denotes those present at the exit meeting on September 19, 1986
The inspector also contacted other members of the licensee's staff including senior nuclear shift supervisors, reactor operators, test engineers, QA engineers and members of the technical staff.
2.0 Licensee Actions on Previous Inspection Findings (Closed) Violation (354/86-27-01) Use of unapproved temporary procedures to perform surveillance tests and failure to complete review and approval of on-the-spot changes (OSCs) to procedures within fourteen days.
The inspector reviewed Revision 5 to station procedure SA-AP.ZZ-032, Review and Approval of Station Procedures, which was issued to clarify the review and approvals required for temporary procedures and OSCs. The inspector also reviewed the Hope Creek Operations Required Reading Book and methods used by the operations department to track OSCs to verify satisfactory implementation.
In addition, the inspector reviewed the closure of Quality Assurance Corrective Action Request (HS-86-C011-0) which involved similar deficiencies identified by the station quality assurance organization and determined that overall the number of open corrective action requests had been significantly reduced.
The inspector concluded that the licensee's corrective actions for this violation were acceptabl.
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m (Closed) Unresolved Item (354/86-22-01), Licensee to sample Primary Containment Instrument Gas (PCIG) system to verify particle size of contaminants.
During Inspection 50-354/86-22 the inspector noted that PCIG samples had been taken at the air receiver drains instead of at the end of the feeder lines as suggested by Regulatory Guide 1.68.3.
The licensee stated that further samples would be taken and analyzed to determine particle size.
In addition, if particle size exceeded the 3.0 micron size criterion the licensee would request information from vendors of safely related equipment specifying maximum particle size that their equipment could tolerate without any adverse effects.
The inspector reviewed the Princeton Laboratory test results as stated in a May 21, 1986 letter. The results indicated that all PCIG feeder lines exceeded the 3.0 micron size with 0.0 percent detected in the 10 to 20 micron range in any line. Two lines had a small percentage (3 to 4%) of detected particles in the 20 to 50 micron range. No particles greater than 50 microns in size were identified. The inspector reviewed General Electric specifications and letters from Atwood & Morrill Company Incorporated which stated that particle sizos of 50 microns or less would not affect equipment operation.
3.0 Power Ascension Test Program (PATP)
3.1 References
Regulatory Guide 1.68, Revision 2, August 1978, " Initial Test Programs for Water-Cooled Nuclear Power Plants"
ANSI N18.7-1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants"
Hope Creek Generating Station (HCGS) Technical Specifications, Revision 0, April 11, 1986 HCGS Final Safety Analysis Report (FSAR), Chapter 14, " Initial Test Program"
HCGS Safety Evaluation Report (SER), Chapter 14, " Initial Test Program"
Station Administrative Procedure, SA-AP.ZZ-036, Revision 3,
" Phase III Startup Test Program"
Specification NEB 0 23A4137, Revision 0, " Hope Creek Startup Test Specification
HCGS Power Ascension Test Matrix, Revision 8
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3.2 Overall Power Ascension Test Program Discussion The inspector held discussions with various members of the PATP staff to assess the overall status of the test program and the test results review.
All test results for Test Condition 1, reviewed in a previous inspection report (50-354/86-43) have been formally accepted by licensee management.
The licensee completed Test Condition 2 (25%-45% rated power) on September 11, 1986 with the performance of TE-SU.ZZ-311, Loss of Offsite Power, and entered a maintenance outage to address previously identified problems and problems encounted during the loss of power test.
On September 16, 1986 the inspector attended the Station Operations Review Committee (SORC) meeting (86-237) called to review the results of the Test Condition 2 plateau. The power ascension staff provided an overview of the test condition including a summary of the results of all testing performed, the tests not performed and justification for deferral, the status of all open results deficiencies (RDFs) and unapproved on-the spot changes (OSCs). Conformance to Regulatory Guides and FSAR commitments was addressed and SORC recommended that test condition 2 be officially closed.
On September 18,1986 the inspector attended the Technical Review Board review of TE-SU.ZZ-311, Loss of Offsite Power, test results.
The review involved verification of conformance to acceptance criteria, review of RDFs and resolution of independent reviewer comments.
During the review, recommendations for procedure changes prior to the reperformance of the test were decided on and responsibilities assigned to resolve problems revealed during the first performance of this test.
Findings No unacceptable conditions were identified.
3.3 Power Ascension Test Procedure Review Scope The Power Ascension Test Procedures listed in Attachment A were reviewed for their conformance to the requirements and guidelines of the references in Paragraph 3.1 and for the attributes previously defined in Inspection Report 50-354/86-03.
Findings No unacceptable conditions were identified.
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3.4 Power Ascension Test Witnessing Scope The inspector witnessed the performance of the power ascension tests listed in Attachment B and discussed below.
The performance of these tests were witnessed to verify the attributes previously defined in Inspection Report 50-354/86-35.
Discussion TE-SU.CH-272.
This test was performed to demonstrate that, within the capacity of the turbine bypass valves (25% of rated steam flow), a main turbine generator trip will not cause a reactor scram. The demonstration was smoothly performed by both test and operations personnel and the resultant reactor transient was very mild with no challenges to the reactor protection system.
TE-SU.ZZ-311.
The performance of the loss of offsite power test was witnessed by two Regional inspectors and the Senior Resident Inspector (addition details may be found in Inspection Report 50-354/86-47).
The purpose of the test was to demonstrate that all safety related equipment necessary to achieve a safe shutdown condition would operate properly with only on-site power available.
The test also verifies that sufficient instrumentation and controls are available to monitor and maintain the plant in a safe shutdown condition.
The test was initiated from approximately 22% rated power with all systems in their normal alignments.
Prior to the start of the test, the inspector observed test preparations including the operations shift turnover meeting, on-shift briefing by both the shift test coordinator and senior nuclear shift supervisor and the verification of prerequisites.
Operations personnel were made aware of the scope of the test and their individual responsibilities.
Conditions were discussed which would cause the termination of testing and contingent emergency actions were reviewed for the unlikely possibility that insufficient on-site power would be available (less than three emergency diesel generators (EDG)).
At the command of the senior nuclear shift supervisor the test was initiated at 2006 by tripping the main turbine generator and simultaneously opening two breakers to de-energize the station ring bus.
A reactor scram occurred from loss of power to the RPS busses and a full MSIV closure was verified. All four EDGs started but the
"C" EDG output breaker failed to automatically close to provide power to its bus.
The failure of the
"C" EDG output breaker is a Level 1 test acceptance criteria failure and is documented as RDF #98.
In accordance with pre-test planning, the senior nuch ar shift supervisor elected to continue the test, for the purpose of data gathering, and instructed the control room operator to manually close the "C" EDG output breaker.
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During the next several minutes a number of systems were noted to be functioning improperly.
These included the Reacts Building Ventilation System (RBVS), the Reactor Auxiliaries Cooling System (RACS), the Acoustic
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Monitoring System and the Emergency Service Air System.
At 2010 the senior nuclear shift supervisor declared a test abort and directed control room personnel to restore normal off-site power.
During the test and subsequent recovery actions the inspector monitored reactor response and observed a very mild transient.
From an initial pressure of 930 psig, reactor pressure peaked at approximately 970 psig during the MSIV closure and reactor water level dropped to approximately
+5 inches wide range (166 inches above the top of the active fuel).
Subsequent to the scram, decay heat caused a slow, steady rise in reactor pressure to 1047 psig at which point the "H" SRV actuated in-the Lo-Lo set mode to control pressure as designed.
Following the SRV lift, reactor water level was observed to be approximately 5 inches lower. Following a second lift of the
"H" SRV, control room operators manually initiated RCIC to provide makeup to the vessel and as an alternate pressure control mechanism.
During the test an NRC inspector toured the reactor building to observe available lighting levels and look for any unexpected conditions. During the test, the lighting levels appeared adequate in the areas toured. The inspector observed that the RACS pumps were not running as required (see discussion of problems below). The inspector also observed water spilling onto the floor from overhead piping in the SACS room near panel 1AC281.
The licensee was investigating the source of this water and its possible impact.
The inspector observed and evaluated overall shift crew performance during the conduct of the test and subsequent recovery actions. The actions of both test and operations persor.nel were judged to be excellent.
All required actions were carried out in an expeditious and professional manner. The decision of the senior nuclear shift supervisor to abort the test was timely and prudent and the safety of the plant was never in jeopardy.
Following the test, the licensee initiated immediate actions to identify and correct problems and deficiencies revealed during its performance and the subsequent recovery.
Numerous items were identified for investiga-
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Reactor Building Ventilation System.
The RBVS supply and exhaust fans started but tripped on low flow when their dampers failed to function.
Investigation revealed that the power supply to the damper solenoids was from a non--1E source. This problem had been previously identified by the licensee and a design change package (DCP) was available to change the t
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solenoids from AC to DC to provide power from an uninterruptible source.
The licensee had failed to properly prioritize this DCP.
The DCP was completed and the licensee committed to review all DCPs to verify proper prioritization.
SRV Acoustic Monitor System.
The acoustic monitor lost pwer during the test.
Investigation revealed that is was not powered from an uninter-ruptable source. This monitor is a technical specification post-accident monitor and is required to be supplied from an uninterruptable power supply.
Reactor Auxiliaries Cooling System. _ The "A" pump started as designed but tripped on low head tank level. The low head tank level was caused by a system interaction with the chilled water system.
As planned, at 85 seconds following the loss of power, the RACS pump started, the RACS isolation valves to the drywell coolers began to open and the chilled water system isolation valves to the drywell coolers began to close.
Since both sets of isolation valves were open simultaneously the RACS pump transferred water to the chilled water system head tank from the RACS head tank.
The licensee issued a DCP to change the sequencer timing to insure that the chilled water system isolation valves are fully closed before the RACs isolation valves begin to open.
Emergency Service Air Compressor. The compressor started but tripped due to loss of RACS cooling.
The solution to the RACS system problem discussed above will solve this problem also.
"C" EDG Output Breaker.
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"C" EDG output breaker failed to automatic-ally close due to two failed input buffers to its Bailey logic module.
The logic module was replaced.
Following resolution of these problems, the licensee performed a loss of power test from a cold, shutdown condition to verify that all identified problems had been successfully resolved. A discussion of this retest is provided below.
TE-TE.ZZ-312.
The purpose of this test was to verify that all problems identified during the performance of TE-SU.ZZ-311, Loss of Offsite Power, had been successfully resolved.
Prior to the start of the test, tne inspector observed the on-shift briefing conducted by the shift test coordinator and senior nuclear shift supervisor. Operations personnel were made aware of the scope of the test and their individual responsibilities, conditions were discussed which would cause the termination of testing and contingent emergency actions were reviewed.
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At the command of the senior nuclear shift supervisor the test was initiated at 2150 by simultaneously opening two breakers to de-energize the station ring bus.
All four Ei?3s started and successfully loaded on to their respective busses.
Operations and test personnel monitored system performance and control room indications to assure that all equipment, especially those items modified following TE-SU.ZZ-311, were operating. properly. All problems noted during the previous loss of power test appeared to have been corrected.
Difficulty was experienced in placing the emergency service air compressor in service and it was also noted that a valve, which should have closed to isolate the service air header from the service air receivers, was indicating in the mid position.
These two problems caused the air operated (exhaust just prior to intake)
dampers in the reactor building ventilation system to close at approxi-mately 18 minutes into the test and resulted in a positive pressure in the reactor building. Operations personnel manually initiated the Filtration, Recirculation and Ventilation System (FRVS) to reestablish a negative secondary containment pressure. At approximately 21 minutes into the test the emergency service air compressor was returned to service, R8VS reestablished and FRVS secured, Following the completion of the required 30 minute run without off-site power, the senior nuclear shift supervisor directed the restoration of the normal electrical line-up and the test was completed.
The inspector observed and evaluated operations and test personnel performance during the conduct of the test. The actions of both test and operations personnel were judged to be excellent.
Following the test, the licensee initiated an immediate review of deficiencies disclosed during the test and recovery. A number of items were identified for investigation. The results of this investigation will be followed by the senior resident inspector.
Findings Within the scope of the inspection one unacceptable condition was identified involving failure to power the acoustic monitor from an uninterruptable source.
This issue will be dealt with in the resident inspector's report (see Inspection Report 50-354/86-47).
3.5 Power Ascension Test Results Evaluation Scope The power ascension test results listed in Attachment C were evaluated for the attributes identified in Inspection Report 50-354/86-24. A summary of
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significant test results and identified test results deficiencies is provided in the discussion below.
Discussion TE-SU.BB-223. All acceptance criteria for pressure regulator response and stability were satisfied.
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TE-SU.SV-281.
The hot shutdown acceptance criteria for this test were satisfied.
During the performance of the cold shutdown portion, at approximately 75 psig reactor pressure, SRV
"H" failed to reclose following a manual lift and forced the termination of testing.
RDF #85 documents this failure. A retest of the cold shutdown demonstration will be performed at a later time.
TE-SU.AB-341.
This test was run in conjunction with TE-SU.SV-281, Shutdown from Outside the Control Room. The test monitors the main steam lines inside the drywell to verify that transient loads caused by SRV lifts during major plant trips do not exceed established limits.
Since no SRVs lifted following the plant trip, meaningful data could not be obtained.
This test will be performed in conjunction with all planned trips in the power ascension program.
TE-SU.AB-344.
All measured loads on the SRV discharge piping were well within acceptance criteria limits.
TE-SU.GT-723.
All acceptance criteria on drywell undervessel area temperatures were met.
TE-SU.ZZ-005.
The inspector reviewed the test matrix for test condition two and verified that all planned testing had been completed or evaluated for deferral.
In addition, the inspector reviewed 24 open RDFs to verify that they could be safely carried forward to the next test condition.
TE-SU.SE-101. The level 1 acceptance criterion for overlap was satisfied for all detectors except IRM "E" (RDF #106) which was inoperable at the time of the test.
All IRMs failed to satisfy the level 2 acceptance criterion (RDF #92) and a complete reperformance of this test will be conducted following a high power APRM heat balance calibration in Test Condition 3.
TE-SU.SE-103. The Level 1 acceptance criterion for overlap was satisfied for all detectors.
IRM "E" failed the Level 2 acceptance criterion for a full decade overlap (RDF #91).
The other seven IRMs required gain adjustments to sati sfy the criterion.
The licensee plans to reperform this test in Test Condition 3 following a high power APRM heat balance calibration.
TE-SU.CH-274.
The capacity of the main turbine bypass valves was conservatively measured to be 25.2% of rated steam flow which satisfied the acceptance criterion.
TE-SU.ZZ-311.
This test was witnessed by the inspector and is discussed in detail in paragraph 3.4.
A Level 1 acceptance criteria failure (RDF
- 98) occurred when the "C" EDG output breaker did not automatically close onto its bus. Level 2 acceptance criteria failures occurred due to loss of power to the acoustic monitor, problems with rod position indication and failure of SRV tailpipe temperatures to return to within 10' F of their pre-test values. The licensee plans to reperform this test at the earliest opportunity in Test Condition 3.
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Findings No unacceptable conditions were identified.
4.0 Independent Measurements and Verifications The inspector performed multiple independent measurements and verifica-tions during the witnessing of power ascension testing (paragraph 3.4) and during the evaluation of test results (paragraph 3.5).
In all cases the inspector's measurements and verifications agreed with those of the licensee.
No unacceptable conditions were noted.
5.0 QA/QC Interface with the Power Ascension Test Program During the course of witnessing the loss of power tests, the inspector observed QA engineers performing surveillances of selected power ascension tests. The inspector also noted during etaluation of power ascension test results that the test results packages had been reviewed by QA engineers.
No unacceptable condition were moted.
6.0 Tours of the Facility During the performance of power ascension test TE-SU.ZZ-311, Loss of Offsite Power, the inspector toured various areas of the reactor building to assess lighting levels and to observe any unexpected or abnormal conditions.
Findings No unacceptable conditions were identified.
7.0 Exit Interview At the conclusion of the site inspection on September 19, 1986, an exit meeting was conducted with the licensee's senior site representative (denoted in paragraph 1.0).
At no time during the inspection was written material provided to the licensee by the inspector.
Based on the NRC Region I review of this report and discussions with licensee representatives during the inspection, it was determined that this report does not contain information subject to 10 CFR 2.790 restriction s
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Attachment A Power Ascension Test Procedures Reviewed TE-SU.ZZ-006 Test Plateau Matrix Test Procedure for Test Condition Three, Revision 0, approved September 13, 1986 TE-SU.ZZ-016 Chemical and Radiochemical 90 to 100% Power Tests and No-Reactor Water Cleanup Test, Revision 0, approved November 29, 1985 TE-SU.AE-233 Feedwater System Feedwater Pump Trip Test, Revision 0, approved October 15, 1985 TE-SU.AE-234 Feedwater Pump Runout Capability Test, Revision 0, approved October 15, 1985 TE-SU.CH-273 Full Power Turbine Trip Test, Revision 0, approved October 22, 1985 TE-SU.BB-291 Recirculation Flow Control System Local Manual Testing, Revision 0, approved November 29, 1985 TE-SU.BB-292 Recirculation Flow Control System Master Manual Mode Testing, Revision 0, approved November 29, 1985 TE-SU.BB-301 Recirculation System Single Pump Trip Test, Revision 0, approved October 28, 1985 TE-SU.BB-302 Recirculation System Two Pump Trip Test, Revision 1, approved August 8, 1986 TE-TE.ZZ-312 Loss of Offsite Power, Reactor in Cold Shutdown, Revision 0, approved September 17, 1986
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Attachment B
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Power Ascension Tests Witnessed
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TE-SU.CH-272 Turb'ine Trip Within Bypass Valve Capacity, performed i
September 11, 1986 i
t TE-SU.ZZ-311 Loss of Offsite Power, performed September 11, 1986
TE-TE.ZZ-312 Loss of Offsite Power, Reactor in Cold Shutdown, performed September 19, 1986
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Attachment C Power Ascension Test Results Evaluated Test condition One TE-SU.BB-223 Pressure Regulator Test-Bypass Valves Controlling, Revision S, completed August 20, 1986, results accepted September 8, 1986 TE-SU.SV-281 Shutdown From Outside the Control Room, Revision 2, completed August 23, 1986, results accepted September 11, 1986
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TE-SU.AB-341 NSSS Main Steam Piping Dynamic Response, Revision 3, completed August 23, 1986, results accepted September 11, 1986 TE-SU.AB-344 Main Steam Relief Valve Discharge Piping Dynamic Response, Revision 3, completed August 22, 1986, results accepted September 13, 1986 TE-SU.GT-723 Drywell Cooling System Post Trip Performance Test, Revision 2, completed August 22, 1986, results accepted
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September 13, 1986 l
l Test Condition Two
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TE-SU.ZZ-005 Test Plateau Matrix Test Procedure for Test Condition Two, Revision 3, completed September 16, 1986, results accepted September 16, 1986 TE-SU.SE-101 SRM/IRM Overlap Verification, Revision 6, completed September 6, 1986, results not yet ec:epted TE-SU.SE-103 IRM/APRM Overlap Verificatioc, i#vis on 4, completed August 31, 1986, results nc; ;v, a 4 pted TE-SU.CH-274 Bypass Valve Capacity Test, Revision 0, completed September 2, 1986, results accepted September 13, 1986 TE-SU.ZZ-311 Loss of Offsite Power, Revision 4, completed September 11, 1986, results not yet accepted
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