IR 05000354/1986024

Insp Rept 50-354/86-24 on 860428-0509.No Violations Noted. Major Areas Inspected:Preoperational & Tech Spec Surveillance Witnessing,Power Ascension Test Program,Loss of Offsite Power Event,Qa/Qc Interfaces & Tours of Facility

ML20205T321
Person / Time
Site:	Hope Creek
Issue date:	05/30/1986
From:	Eselgroth P, Marilyn Evans, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20205T305	List: IR 05000354/1986024 ML20205T303
References
50-354-86-24, NUDOCS 8606130105
Download: ML20205T321 (13)

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U.S. NUCLEAR REGULATORY COMMISSION ,

REGION I

Repor,t N /86-24 ,

Docket N License No. CPPR-i20 Category B Licensee: Public Service Electric and Gas Company

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80 Park Plaza - 17C Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station, Unit 1 Inspection At: Hancocks Bri'dge, New Jersey _

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Inspection Conducted: April 28, 1986 - May 9, 1986 -

Inspectors: b bt>d/RtL/ ..

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6 26 M. E a eac or Engineer da e LV AA197 L." W ' k h

e. tor Engineer /

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Approved by: .

1 M P.EselM,I.h, Chief,TestPrograms I da.te Section, 08. DRS Inspection Summary: Inspection on Apri_1,28,1986 - May 9,1986 (Inspection Report No. 30-354/86-24)

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Areas Inspected: ' Routine unanno m 3 4' .pection by two region-based inspectors of preoperational test c t- .ing, preoperational test results evaluation / review, technical spi... fit u <n surveillance witnessing, power ascension test program, review of loss of offsite power event, QA/QC interfaces, independent measurements and evaluations, and tours of the facility.' ,

Results: No violations were identifie NOTE: For acronyms not defin' de refer to NUREG-0544 " Handbook of Acronyms and Initialisms."

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8606130105 860606 PDR ADOCK C5000354 G PDR

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DETAILS 1.0. Persons Contacted Public Service Electric and Gas (PSE&G) Personnel and Contractors

1. J. Carter, Startup Manager

• G. Chew, Power Ascension Results Coordinator

1. R. Donges, Lead Quality Assurance (QA) Engineer

1. J. Duffy, Supervisory Engineer

• 1. M. Farschon, Power Ascension Manager B. Forward, Power Ascension Administrative Coordinator

1. C. Fuhrmeister, Lead Engineer

• A. Giardino, Manager - Station Quality Assurance (QA)

• R. Griffith, Principal QA Engineer

1. C. Jaffee, Startup Engineer T. Hersum, Power Ascension Engineer

• D. Hosmer, Lead Shift Test Coordinator

1. P. Krishna, Assistant to General Manager - HC0

• 1. S. LaBruna, Assistant General Manager - HC0

1. M. LaVechia, Principal Engineer

• M. Metcalf, Principal QA Engineer D. Moon, Shift Test Coordinator W. Ott, Power Ascension Enginee M. Parsons, Power Ascension Engineer

• R. Salvesen, General Manager - HC0 R. Schmidt, Senior Reactor Supervisor

1. S. Singh, Principal Engineer C. Sullivan, Power Ascension Engineer U.S. Nuclear Regulatory Commission D. Allsopp, Resident Inspector-R. Borchardt, Senior Resident Inspector The inspector also contacted other members of the Licensee Staff including Senior Nuclear Shift Supervisors, Reactor Operators, Test Engineers and members of the Technical Staf * Denotes those present at the interim exit interview on May 2, 198 # Denotes those present at the exit interview on May 9, 198 .0. Preoperational-Test Witnessing 2.1. Scope Testing witnessed by the inspector included the observations of overall crew performance stated in paragraph 3.0 of Inspection Report No. 50-354/85-1 . ._ _

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2.2. Discussion On May 8, the insp2ctor witnessed several portions of PTP-BB-3 (part A), Standby Diesel Generator Loading, Retest #4. The inspector observed the performance of section 8.15, D/G IDG400 Hot Automatic Start and Load Sequencing. The inspector noted that all major safety related components performed satisfactorily. All testing observed satisfied the criteria abov .3. Findings No unacceptable conditions were observed within the scope of this revie .0. Preoperational Test Results Evaluation Review 3.1. Discussion The inspector reviewed the licensee's resolution of 46 of 111 test exceptions identified during previous NRC review of test result Attachment A lists the remaining open test exceptions and-collectively constitutes unresolved item 354/86-24-01. Unresolved item 354/86-23-01 is close Findings No violations were identifie .0. Technical Specification Surveillance Witnessing 4.1. Discussion Two licensee surveillance activities were monitored by the inspector to insure that technical specification surveillance requirements were being adequately me The following surveillance tests were witnessed by the inspector:

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OP-IS.BD-101, RCIC System - Power Operated Valves -

Inservice Test

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OP-IS.EA-101,. Station Service Water System Valve Inservice Test 4.2. Findings No violations were identified within the scope of this review.

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5.0. Power Ascension Test Program (PATP)

.5.1. 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 Assurence for the Operational Phrase 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 2,

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" Phase III Startup Test Program"

Specification NEB 0 23A4137, Revision 0, " Hope Creek Startup Test Specification" HCGS Power Ascension Test Matrix, Revision 3 5.2. Overall Power Ascension Test Program The inspector held discussions with the Power Ascension Manager, Power Ascension Administrative Coordinator and Power Ascension Results Coordinator to verify the review and approval cycle for complete power ascension test results. The initial evaluation of test results is performed by an independent technical reviewer who insures that the test package is complete, acceptance criteria are satisfied or results deficiencies genera ped to track problem Following this review, copies are made and distributed to appropriate reviewers such as the NSSS vendor representative, Architect-Engineer Representative and Quality Assurance. Following these reviews the test package is submitted to the Technical Review Board for final review and comment resolution. The final test results package is then submitted to the Station Operations Review Committee (SORC) who must recommend approval to the Power Ascension Manager for final test results acceptanc .3. Power Ascension Test Procedure Review 5. Scope The power ascension test procedures listed in Attachment B were reviewed for conformance to the requirements and guidelines of

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the references listed in paragraph 5.1 and for the attributes previously defined in Inspection Report'No. 50-354/86-0 . Discussion The inspector had a number of questions concerning various aspects of the listed procedures which were successfully resolved during discussions with various members of the Power Ascension Group. The.following items will be followed by the inspector in a subsequent routine inspection:

TE-SU.SE-121, performs a preliminary calibration of the APRM by means of a constant heatup rate heat balance. The inspector noted that the heat balance used was a closed system, single phase heat balance using the specific heat of water at constant pressure. For this heat balance to be valid the vessel must be isolated (No Feed Flow or Steam Flow) and the coolant temperature must be maintained below 212 F for the entire test. The inspector noted that the initial test conditions specified in the test were not sufficiently clear to insure that these conditions would be me TE-SU.ZZ-162, verifies that the reactor water level instrumentation calibration is correct. The procedure performs calibration calculations using as-built elevation data for the instruments and actual, measured operating environment temperatures. The inspector noted however that the procedure did not compare these results to the actual plant calibration records to assess the adequacy of the initial calibratio TE-SU.BC-713, verifies the design heat exchange capacity of the RHR heat exchangers in the suppression pool cooling mod To correct the measured heat exchange capacity to the design conditions specified on the process diagram a log mean temperature difference (LMTD) method is used. The inspector noted that the LMTD method employed is valid only for true counter flow heat exchangers while the RHR heat exchangers are of a more complex design and would require a slightly different correction facto . Findings Within the scope of the inspection, no violations were identifie .4. Power Ascension Test Results Evaluation 5. Scope The power ascension test results listed in Attachment C were

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reviewed to verify that:

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The test was performed in accordance with a current, approved procedur Test changes were approved in accordance with

' administrative procedure Test changes are annotated in the procedure and completed, if appropriat The basic objectives of the test were me Test deficiencies are documente Test deficiencies were resolved, accepted by management and retests completed, if require System or process changes necessitated by test deficiencies have been properly documented and reviewe Test deficiencies which constitute reportable occurrence have been properly reporte Required test data has been obtained and is within tolerance Individual test steps and data sheets have been properly initialed and date An engineering evaluation of the test data has been performe Test results are evaluated against established acceptance criteri Review and acceptance of test results has been documente Tests results have been reviewed by Quality Assurance or other Independent Organizatio Test results have been approved by appropriate managemen .4.2. Discussion

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TE-SU.KE-032, Fuel Loading The inspector independently evaluated the test results including the 1/M plots maintained during the loading and the partial core shutdown margin test. Fuel Loading began on April 15, 1986 and the last bundle (764) was loaded

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on April 27, 1986. No test deficiencies were issued. The inspector also reviewed the occurrence list to verify proper documentation of difficulties encountered during test performanc TE-SU.KE-033, Full Core Verification The inspector evaluated the test results and verified the Core Loading pattern as documented in RE-FR.ZZ-008, Core Loading Verification. The inspector also reviewed the video tape made during the Core Verification. No test deficiencies were issue TE-SU.BF-051, Control Rod Drive Functional Testing The inspector reviewed both performances of this procedure (Pre and Post Fuel Load). An independent evaluation was made of Control Rod Drive System performance as documented in these tests. No test deficiencies were issue The occurrence list was also reviewe TE-SU.BF-052, Scram Testing of Selected Control Rod The inspector held discussions with the appropriate test engineer to determine the basis for the selection of the four-control rods for this tes It was determined that since no control rods had exhibited any unusual operating characteristics, the rod selection was based only on slow scram times. The selection was further limited by the requirement that these rods would have to be fully withdrawn during the course of a normal startup prior to exceeding 600 psig vessel pressure to permit testing at that condition. This requirement limited the selection to group 1 and 2 rods in Sequence A. When tested the selected rods exhibited acceptable scram times to notch 05 as summarized below:

Rod Time (SEC) Limit (SEC)

26-07 1.664 <7 34-07 1.620 27 34-23 1.606 27 34-47 1.592 37 ,

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TE-SU.BF-053, Control Rod Drive System Friction and Scram Testing at zero reactor pressur The inspector independently evaluated the friction traces obtained for each control rod. The average scram times

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. calculated for all control rods was:

Notch Average Time to Notch (SEC) Limit (SEC)

-45 .205 < .43 39 .384 2 .86 25 .822 E 1.93 05 1.522 33.49 One test deficiency was issued for control rod 38-43 (RDF#004) but had not been reviewed or approved at the time of this review. The occurrence list was also reviewe TE-SU.SE-061, SRM signal-to-noise ratio and minimum count rate determinatio No test deficiencies were issue The minimum count rates were found to be:

SRM Count Rate (CPS) Limit (CPS)

A 69 > .7 8 55 5 .7 C 71 5 .7 0 23 The signal-to-noise (S/N) Ratios were calculated to be:

SRM

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S/N Ratio Limit A 689 >2 B 549 32 C 472 52 D 229 E2 5.4.3. Findings Within the scope of the review, no violations were identifie All test acceptance criteria were satisfied. With the exception of TE-SU.BF-051, Control Rod Drive Functional Testing (Pre Fuel Load) the test results packages had not completed the review and approval process. Engineering evaluations were still on going in most cases and the results had not been reviewed by Quality Assurance or accepted by management. The completion of these reviews and approvals will be verified in a subsequent routine inspectio The inspector noted during the reviews of the occurrence lists included with the test packages that it did not appear that all station administrative requirements were being consistently and uniformly interpreted and applied by the test engineers. The inspector met with the Power Ascension Manager on several

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occasions to discuss these concerns. Pending the completion of the test results review and approval process and subsequent NRC review of the approved test packages, the performance of power ascension tests in conformance with appropriate station admini-stration requirements will remain unresolved (354/86-24-02).

6.0. Review of Loss of Offsite Power Event 6.1. Discussion On May 2, 1986 the licensee experienced a loss of offsite power when, with one switchyard line out of service (Keeney line), a relay tech-nician error resulted in the loss of the two remaining 500KV switch-yard lines (Salem and New Freedom). Due to the abnormal lineup a loss of offsite power occurred. Emergency Diesel Generators (EDG)

A, B and C started, but only EDGs A and C properly loaded. EDG B was out for maintenance and EDG D was undergoing preoperational testing following refurbishment. An alert was declared since three EDGs did not start and properly load as require During the period May 6-8, the inspector reviewed the available in-formation concerning the loss of offsite power event. The inspector reviewed the licensee's Incident Report, the Nuclear Control Operator and the Nuclear Shift Supervisor narratives, and Operating Procedure (0P)-AB.ZZ-135(Q), Revision 1, " Loss of Offsite Power". Data from the alarm chronolog printer was not available for the inspector to verify proper sequencing of safety-related loads because the sequence of events recording function of the process computer was not in ser-vice at the time. Therefore the inspector reviewed Technical Speci-fication surveillance OP-ST.KJ-005(Q) and OP-ST.KJ-007(Q), Integrated EDG's IAG400 and ICG400 Tests - 18 months. These tests were performed on March 31, 1986 and April 7, 1986 respectfull .2. Findings Based on the above review and discussions with operations personnel, at the time of the event it cppears that all available safety-related equipment functioned properl .0. QA/QC Interfaces The inspector reviewed 14 QA surveillance reports covering activities as-sociated with the initial fuel loading. It was noted that the surveil-lances included a preformulated checksheet which detailed critical attri-butes to be monitore The QA/QC involvement in the initial fuel loading activities was acceptabl .

8.0. Independent Measurements and Evaluations The inspector made independent measurements of control rod drive piston differential pressures from photographs of oscilloscope traces made during continuous insert friction testing as part of TE-SU.BF-053. The inspector also evaluated the control rod drive data obtained during TE-SU.BF-051 and TE-SU.BF-053 to verify acceptable control rod drive performanc The inspector's measurements and evaluations agreed with the licensee's 9.0. Tours of the Facility The inspector made several tours of various areas of the facility to ob-serve work in progress, housekeeping, and cleanliness control No unacceptable conditions were note . 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. Unresolved items disclosed during the inspection are discussed in paragraphs 3.1 and 5. . Exit Interview A management meeting was held at the conclusion of the inspection on May 9, 1986 to discuss the scope and findings as detailed in this report (see Paragraph I for attendees). In addition, an interim exit meeting was held with the licensee on May 2, 1986. No written information was provided to the licensee at any time during the inspection. The licensee did not in-dicate that proprietary information was contained within the scope of this inspectio ..

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ATTACHMENT A Procedure N Short Title SDR N PTP-BB-3 (Part A) S/B DG Loading GJ-1353, 1417, 1450 .

PTP-BB-3 (Part B) ECCS integrated KJ-1516 PTP-BB-4 RPV Internal Vibration B8-0220, 0867, 0600, 094 PTP-EA-1 SSW EA-0513, 0528, 0553, 0612, 0652, 0669, 0703, 0706, 0716, 0728, 0747, 0750, 0751,-

0752, 0753, 0760, 0763, ZC-00 PTP-GU-1 FRVS GU-528, 529, 558, 574, 576, 572, 530, 575, 573, 577, 574, 568, 556 and 581; PTP-SV-1 Remote Shutdown Panel BB-1019; BC-1046 and 1080; BD-411 and 496; EG-655 and 666; FC-17; GJ-129, 185 and 195; RL-950; SV-36, 43, 47, 48, 49, 50, 51, 52, 55 and 57; ZZ-99 .

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ATTACHMENT B Power Ascension Test Procedures TE-SU.ZZ-012, Chemical and Radiochemical heatup test, Revision 0, TE-SU.SE-101, Source Range Monitor / Intermediate Range Monitor Overlap Verification, Revision 2, TE-SU.SE-111, Local Power Range Monitor Flux Response Verification, Revision 1, TE-SU.SE-121, APRM Calibration During Heatup, Revision 2, TE-SU.BD-141, Reactor Core Isolation Cooling System Condensate Storage Tank Injection, Revision 2, TE-SU.BJ-154, HPCI Survefilance Test Demonstration, Revision 1 TE-SU.ZZ-162, Water Level Measurement Test, Revision 0, TE-SU.BB-261, Relief Valve Test, Revision 1, TE-SU.ZZ-311, Loss of Offsite Power, Revision 1, TE-SU.ZZ-334, Reactor Core Isolation Cooling Piping Vibration, Revision:2, TE-SU.FD-346, High Pressure Coolant Injection Turbine Steam Supply Piping Dynamic Response, Revision 3, TE-SU.BG-701, Reactor Water Cleanup System Normal Mode Performance Test, Revision 2, and TE-SU.BC-713, Residual Heat Removal System Suppression Pool Cooling Mode Test, Revision g o-ATTACHMENT C Power Ascension Test Results

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.TE-SU.KE-032, Fuel loading, preliminary TE-SU.KE-033, Full core verification, preliminary TE-SU.BF-051, Control rod drive functional testing (Pre fuel load), results approved April 30, 1986 TE-SU.BF-051, Control rod drive functional testing (Post fuel load),

preliminary TE-SU.BF-052, -Scram testing of selected control rods, preliminary TE-SU.BF-053, Control rod drive system friction and scram testing at zero reactor p' essure, preliminary TE-SU.SE-061, SRM signal-to-noise ratio and minimum count rate determination, preliminary

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