IR 05000354/1986023
| ML20211D720 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/04/1986 |
| From: | Aolla J, Briggs L, Eselgroth P, Marilyn Evans, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20211D700 | List: |
| References | |
| 50-354-86-23, NUDOCS 8606130136 | |
| Download: ML20211D720 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-354/86-23 Docket No.
50-354 License No.
CPPR-120 Category B Licensee:
Public Service Electric & 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: April 14-25, 1986 Inspector -
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1%Le nds eactor Engineer dat
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L. Wine, Reactor Engineer date -
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Evans, Reactor E ineer date
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Approved by: '(g e
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P. Eselgro Chief, Test Programs Section, date 08, DRS Inspection Summary:
Inspection on April 14-25, 1986 (Inspection Report No.
50-354/86-23.
Areas Inspected:
Routine, unannounced inspection by five region-based inspec-tors of followup of licensee actions on previous inspection findings, initial fuel load witnessing and plant tours, preoperational and detailed test results review evaluation, QA/QC interfaces and independent verification.
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 & Gas Company
- J. Carter, Startup Manager
- G. Connor, Operations Manager M. Dick, Startup Test Engineer
- M. Farschon, Power Ascension Manager
- A. Giardino, Manager, Station QA, Hope Creek J. Gibson, Startup Test Coordinator
- R. Griffith, Principal QAE F. Hughes, Senior Nuclear Shift Supervisor
- C. Jaffee, Startup Engineer D. Kelly, Startup Test Engineer
- P. Krishna, Assistant to General Manager, Hope Creek Operations
- P. Kudless, Maintenance Manager
- S. LaBruna, Assistant General Manager, Hope Creek Operations
- M. Metcalf, Principal QAE
- G. Moulton, Project QAE W. Ott, Startup Test Engineer E. Riley, Senior Nuclear Shift Supervisor
- R.Salvesen, General Manager, Hope Creek Operations W. Schell, Power Ascension Technical Director W. Thomas, Startup Test Coordinator Other NRC Personnel
- R. Borchardt, Senior Resident Inspector
- J.'Lyash, Resident Inspector
- L. Norrholm, Chief, Reactor Projects Section 2B
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The inspectors also contacted other members of the' licensee's operating and QA/QC staff.
- Denotes those present at the interim exit meeting conducted on April 18, 1986.
- Denotes those present at the exit meeting conducted on April 25, 1986.
2.0 Licensee Actions on Previous Inspection Findings (Closed) Violation (354/86-03-03) Deficiencies in the startup test proce-dure for initial criticality and full core shutdown margin determination.
During a previous inspection (50-354/86-22) the inspector had reviewed the revised startup test procedure, TE-SU.ZZ-041, Full Core Shutdown Margin Demonstration, Revision 1 in detail and verified that the
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specific identified deficiencies had been corrected.
In addition, the inspector has observed on multiple occasions the functioning of the Tech-nical Review Board and verified that all startup test procedures are being reviewed in detail to insure technical adequacy and conformance to regula-tory requirements. The inspector had no further questions concerning the corrective steps taken.
(Closed) Violation (354/86-10-01) Failure to test acoustic monitors in ac-cordance with Preoperational Test Procedure (PTP) SN-1, Automatic Depres-surization System. The licensee has issued work order (WO) 86-03-029-035-5 to verify proper installation and to functionally test the acoustic monitors.
This work order will not be performed until just before Drywell closure to prevent physical damage to the monitors and to ensure operability before initial criticality. The performance of the work order is being followed under unresolved item 354/86-10-03 which will be closed after NRC verifica-tion of proper installation and testing of acoustic monitors.
(Closed) Violation (354/86-10-02) Inadequate review of PTP SN-1 and GT-1.
Subsequent to corrective action inplemented by the licensee, as stated in Inspection Report 50-354/86-12, the inspector observed a large reduction in the number of items that were not identified by the licensee during the results review cycle.
The inspector found the results of licensee action taken to be acceptable.
(Closed) Unresolved Item (354/86-12-02) Licensee to perform QA restoration verification of FRVS ventilation fans low flow switenes.
The inspectors reviewed QA startup surveillance report SR 9038 which verified that switch-es 1GU-FSL-9426 Al and 81 were connected in accordance with appropriate documents. The inspectors reviewed drawings M84, Reactor Building Exhaust Control Diagram and E-0472-0, Reactor Building Exhaust FRVS Ventilation Fans to verify that the correct componerts had been checked.
Corrective action was acceptable.
3.0 Initial Fuel Load Witnessing and Plant Tours 3.1 Scope NRC inspectors began onsite shift coverage on April 12, 1986 in anti-cipation of fuel loading.
Initial fuel loading commenced on April 15 at 6:40 P.M.
Onsite shift coverage was maintained through the comple-tion of the partial core shutdown margin test on April 18, 1986.
Fuel loading and associated testing activities were observed in the Control Room, on the refueling floor, on,the refueling bridge, and in the reactor building near the control rod drive hydraulic control units.
In addition to observations of fuel loading and testing activities, the inspectors performed general plant tour _
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The Senior Resident Inspector and Resident Inspector also participat-ed in the inspection of fuel loading activities.
The details of their inspection activities are included in Inspection Report 50-354/86-20.
Fuel load activities were reviewed to verify that:
The current, approved revisions of the fuel handling and test
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procedures were available and being followed.
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The crew requirements, as defined in the approved procedures, were being met in the control room and on the refuel floor and that technical specification minimum staffing requirements were being satisfied.
Continuous communications were established and maintained
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between the refueling bridge and the control room during core alterations.
Proper access controls, housekeeping requirements and radiologi-
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cal protective measures were in effect on the refuel floor.
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Refueling status boards were maintained current in the control room and on the refueling floor.
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Required nuclear instrumentation was available and operating properly.
The following Test procedures were being performed by the licensee:
TE-SU.KE-032, Fuel Loading
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TE-SU.BF-051, CRD Functional Tests
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TE-SU.BF-053, CRD Friction and Scram Tests
The following specific activities were observed:
General operation of the refueling bridge.
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Movement of new fuel from the spent fuel pool to the reactor vessel.
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Overall test control and coordination from the control room.
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Nuclear instrumentation response checks (source bugging).
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Performance of subcriticality checks.
Maintenance of inverse multiplication plots.
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Control rod functional checks.
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Control rod friction and scram tests.
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Shift turnovers.
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Partial core shutdown margin test with 144 fuel bundles loaded.
3.2 Discussion Prior to fuel movement into the core, the inspector verified that the shorting links were removed and witnessed source checks of SRM A and C.
The inspector also witnessed portions of the refueling bridge interlock testing. By review of the I and C completed surveillance log the inspector also verified that surveillances were completed for other nuclear instrumentation. The inspector also witnessed the con-duct of the shift test briefing conducted prior to movement of fuel and observed that the briefing was comprehensive and thorough and covered the fuel movement as well as performance of the control rod testing.
Fuel loading began at 6:40 PM on April 15, 1986. At the time of fuel loading began, one of the four source range monitors (SRM "B") was considered inoperable due to intermittent electric failures.
Techni-cal specifications and the fuel loading pattern chosen by the licensee allow fuel loading to begin and up to 60 bundles to be loaded before SRM "B" (or a substitute fuel loading chamber) must be operable in the "B" core quadrant.
During the loading of the first 16 bundles the inspector noted that the core map tag board in the main control room indicated that bundle LYB 758 was loaded into the core rather than LYB 785 as indicated on
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the rod loading sequence.
Fuel loading was suspended until locations were verified and when corrected fuel loading was core map resumed.
Following the successful loading of the first 16 bundles, the first subcriticality check was performed by notch-wise withdrawal of control rod 18-47. When the operator attempted to re-insert this control rod by depressing the " Continuous Insert" pushbutton on the Reactor Manual Control System Panel the rod failed to respond.
The operator then attempted to insert the rod using the normal " Insert" pushbutton and the rod responded properly.
Fuel loading was suspended and trouble shooting revealed that oxidation on the silver switch contacts of the
" Continuous Insert" pushbuttons had prevented proper electrical con-tact.
The contacts were cleaned to resolve the problem. At 11:25 PM t
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a reactor protective system (RPS) trip (non-coincident mode) occurred due to a momentary spike on IRM "F".
Since all rods were fully in-serted at the time, no control rod motion occurred. The system was observed to function properly and the investigation revealed that a
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technician working on the SRM "B" had bumped a cable to IRM "F" caus-ing the spike.
Proper and timely ENS notifications were made.
Fol-lowing resetting of the scram, functional, friction and scram testing were performed on control rod 18-47 and fuel loading was resumed at 1:53 AM on April 16. Four additional bundles had been loaded when a
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second RPS trip (non-coincident mode) was received due to APRM "F" upscale. Again no control rod motion occurred, since all rods were fully inserted. The investigaton revealed that a single LPRM in the APRM channel had electrically failed upscale.
Since gains for both the LPRMS and APRMs are initially set conservatively high, this was sufficient to cause the APRM to exceed its upscale trip setpoint.
Following ENS notifications, bypassing of the failed LPRM and scram reset, fuel loading was again resumed.
Fuel loading was again halted when 60 bundles had been-loaded to allow the installation, calibration and testing of a fuel loading chamber (FLC) in the "B" core quadrant to serve in place of the inoperable SRM "B".
This FLC was electrically connected to the SRM "C" electro-nics to allow trouble shooting and repairs to continue on SRM "B".
The inspector witnessed portions of the SORC meeting which reviewed the procedures necessary to accomplish the above.
Following the in-stallation of the FLC, nuclear instrumentation response checks were performed, prerequisites for fuel loading were re-verified and fuel loading resumed at 2:26 AM on April 17. At 12:10 AM on April 18, 144 bundles had been loaded and preparations were begun for the partial core shutdown margin test.
Prior to beginning this test, operations and test personnel had to resolve a problem noted during the scram testing of control rod 38-43, one of the nine rods that would be witn-drawn during the demonstration.
This control rod had exhibited unus-ual scram times as calculated by the GETARS scram timing program.
Although this control rod satisfied all technical specifications and test criteria, the test personnel had elected to conservatively treat it as a test failure until the anomaly could be resolved and had placed a test hold on further use of this rod. Detailed examination of the raw data recorded during the scram test of this rod indicated that the GETARS scram timing program had misinterpreted a spurious signal pulse which occurred approximately 1.7 seconds before the actual scram initiation signal and had added this time to the actual scram times of the control rod.
To confirm this interpretation, operations and test personnel decided to re perform the scram test on this rod. The re-test demonstrated the satisfactory performance of the control rod and the hold on its use was lifted, permitting the performance of the partial core shutdown margin test.
The partial core shutdown margin test commenced at 4:26 AM on April 18 and at 5:36 AM all nine control ro'ds had been fully withdrawn and the partial core verified to be
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subcritical.
By 6:00 AM all control rods had been returned to their fully inserted positions and preparations were made to resume fuel loading.
The inspector also reviewed the licensee analysis to con-firm that the GETARS values of scram time, were consistent with strip chart traces. A March 6, 1986 document reviewed compared 7 control rod strip chart traces with GETARS results. GETARS provided equiva-lent results.
This satified inspector concerns.
3.3 Findings No violations or deficiencies were identified.
4.0 Preoperational and Detailed Test Results Evaluation Review 4.1 Scope The completed test procedures listed in attachment A were reviewed during this inspection to verify that adequate testing had been con-ducted to satisfy regulatory guidance, licensee commitments and FSAR requirements and to verify that uniform criteria were being applied for evaluation of completed test results in order to assure technical and administrative adequacy.
The inspector reviewed the test results and verified the licensee's evaluation of test results by review of test changes, test exceptions,-
test deficiencies, "As-Run" copy of the test procedures, acceptance criteria, performance verification, recording conduct of test, QC inspection records, restoration of system to normal after test, inde-pendent verification of critical steps or parameters, identification of personnel conducting and evaluating test data, and verification that the test results have been approved.
4.1.1 Discussion
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PTP-GJ-1 During review of PTP-GJ-1, the inspector noted that the completed procedure was very confusing with numerous test exceptions, procedure change notices (CN) and on-the-spot (OTS) changes.
The inspector also noted that during the licensee's reviewing process, quality assurance had recom-mended a complete retest of the PTP.
This recommendation was based on the difficulty of determining whether the intent of the test had been met.
Subsequently 17' retests were performed. The inspector reviewed these completed and approved retests as well as the original PTP and found that all acceptance criteria appeared to be satisfied.
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PTP-GP-1
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The Hope Creek generating station preoperational contain-ment integrated leak rate test (CILRT) was concluded on January 3, 1986. The purpose of the test was to demon-strate the acceptability of the primary containment leakage rate at the calculated design basis accident pressure of 48.1 psig.
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The inspector reviewed the summary technical report sub-mitted to the NRC by the licensee entitled " Primary Reactor Containment Preoperational Integrated Leakage Rate Test Final Report" dated March 1986.
The results therein indi-cate a successful (CILRT). The calculated leakage rate based on the mass point method of analysis is 0.176 wt.%/
day of containment air. At the 95% upper confidence limit (UCL) the leakage is 0.181 wt.%. day. With the addition of local leakage rate from valves out of LOCA alignment during the test, the corrected total leakage rate is 0.193 wt.%/
day at 95% UCL. This is below the allowable leakage rate of 0.375 wt.%/ day.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CILRT was followed by a successful 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> veri-fication leak rate test (VLRT).
This test is designed to check the accuracy of the CILRT instrumentation.
The supe-rimposed leak for the VLRT was 0.470 wt.%/ day. The measured composite leakage rate was 0.571 wt.%/ day which fell within the acceptance interval of 0.507 wt.%/ day to 0.763 wt.%/ day.
The inspector had several minor questions during the review of the remaining DTPs listed in Attachment A.
All questions were satisfactorily answered by the licensee.
4.1.2 Findings No violations were identified during the above review.
4.2 Test Exceptions 4.2.1 Discussion No new test exceptions were generated during the above re-view of test results.
During this inspection the inspec-tors reviewed the licensee's resolution of 99 of 210 test exceptions identified during previous NRC review of test results. Attachment B lists the remaining open test excep-tions and collectively constitute unresolved item 354/
86-23-01. Unresolved item 354/86-21-02 is closed.
4.2.2 Findings No violations were identifie.
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5.0 QA/QC Interfaces The startup section of the station Quality Assurance organization has com-mitted to provide coverage of approximately 25*4 of all power ascension tests, which have been preselected based on their significance. Spot coverage will be provided for other tests.
Power Ascension Test TE-SU.KE-032, Fuel Loading, had been selected for full coverage and during this evolution QA inspectors were observed performing their functions on all shifts. No un-acceptable conditions were noted.
6.0 Independent Measurements, Calculations and Verifications During this inspection, the inspectors performed the following independent measurements, calculations and verifications:
On a sampling basis, control rod withdrawal and insert times and ob-
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servation of appropriate indications during the performance of TE-SU.BF-051, CRD Functional Tests.
On a sampling basis, evaluation of control rod friction traces and
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scram time data during the performance of TE-SU.BF-053, CRD Friction and Scram Tests.
Calculations of expected nuclear instrumentation response during con-
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trol rod withdrawals performed for the partial core shutdown margin demonstration based on reactivity data provided in TE-SU.KE-032, Fuel Loading.
7.0 Plant Tours Plant tours conducted during this inspection are discussed in Paragraph 3.
8.0 Unresolved Items Unresolved items are matters about which more information is required in order to determine whether they are acceptable, an item of noncompliance or a deviation.
The resolved item identified during this inspection is discussed in Paragraph 4.2 of this report.
9.0 Exit Interview At the conclusion of the site inspection on April 25, 1986, an exit inter-view was conducted with the licensee's senior site representatives (denoted in Section 1).
The findings were identified and previous inspection items were discussed. At no time during this inspection was written material provided to the licensee by the inspector.
Based on the NRC Region I review of this report and discussions held with licensee representatives during this inspection, it was determined that this report does not contain in-formation subject to 10 CFR 2.790 restriction, _ _
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ATTACHMENT A Preoperational and Detailed Test Procedures Reviewed PTP-GJ-1, Auxiliary Building Control Area Chilled Water System, Revision
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0, results PORC approved April 7, 1986 PTP-GP-1, Primary Containment Integrated Leak Rate Test, Revision 0,
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results PORC approved February 28, 1986 DTP-SB-0004, RRCS (ATWS) Recirculation Pump Breaker Response Time Test,
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Revision 0, results PORC approved March 7, 1986 OTP-SB-0006, Drywell High Pressure Scram Response Time Test, Revision 0,
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results PORC approved March 7, 1986 DTP-SB-0007, Reactor Vessel Low Level Scram Response Time Test, Revision
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0, results PORC approved March 12, 1986
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DTP-SB-0008, Reactor Vessel High Steam Dome Pressure Scram Response Time
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Test, Revision 0, results PORC approved March 7, 1986 DTP-SB-0009, Main Steam Line Radiation Monitors D11-K610A-D and RPS Trip
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Logic Response Time, Revision 0, results PORC approved March 14, 1986 DTP-SB-0010, Scram Discharge Volume High Level, MSIV closure /RPS Trip
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Logic Response Time, Revision 0, results PORC approved March 14, 1986 DTP-SB-0011, RPS/ Intermediate Range Monitor Response Time, Revision 0,
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results PORC approved March 12, 1986 DTP-SB-0012, RPS/ Average Power Range Monitor Response Time, Revision 0,
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results approved March 25, 1986.
Retest No.1 (Thermal Power Constant)
results PORC approved April 2, 1986 DTP-SB-0014, Main Steam Reactor Level 1 Response Time Test, Revision 0,
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results PORC approved March 12, 1986
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ATTACHMENT B Open Test Exceptions Procedure No.
Short Title SDR No.
PTP-88-4 RPV Interval Vibration B8-0220, 0867, 0600 and 0942 PTP-SV-1 Remote Shutdown Panel BB-1011 and 1019; BC-1046, 1080, 1141 and 1142; B0-411, 482 and 496; EG-562, 577, 665 and 666; FC-17; GJ-129, 185 and 195; RL-942, 944 and 950; SV-36, 39, 43, 45, 46, 47, 48, 49, 50, 51, 52, 53, 54,55 and 57; ZZ-996
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PTP-GU-1 FRVS GV-528, 529
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558, 574, 576, 572, 530, 575,
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573, 577, 574, 568, 556 and
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581 PTP-EG-1 STACS EG-6 37, 709,
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722, 724, 727, 781, 783.
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Attachment B
PTP-88-3 S/B DG Loading KJ-1353, 1417, (PARTA)
and 1450 PTP-BB-3 ECCS Integrated KJ-1516.
(PART B)
PTP-EA-1 SSW EA-0471, 0506, 0513, 0528, 0529, 0553, 0572, 0574, 0612, 0636, 0637, 0652, 0656, 0659, 0669, 0683, 0685, 0687, 0692, 0686, 0703, 0704, 0706, 0708, 0716, 0728, 0742, 0743, 0744, 0746, 0747, 0748, 0749, 0750, 0751, 0752, 0753, 0757, 0758, 0760, 0761, 0762, 0763, 0764, ZC-003.
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