IR 05000354/1986010

Insp Rept 50-354/86-10 on 860127-0207.Violations Noted:Two Mandatory Witness Points Not Signed During Procedure CN-16 & Adequate Review of Test Results Not Performed,Causing Failure to Identify Deficiencies

ML20140H202
Person / Time
Site:	Hope Creek
Issue date:	03/07/1986
From:	Briggs L, Eselgroth P, Marilyn Evans, Florek D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140H177	List: IR 05000354/1986010 ML20140H171 ML20140H189
References
50-354-86-10, NUDOCS 8604040074
Download: ML20140H202 (14)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. ' 50-354/86-10 Docket No.

50-354 License No. CPPR-120 Priority Category B_

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Licensee: Public Service Electric and Gas Company 80 Park Plaza - 17C Newark, New Jersey 07101'

Facility Name: Hope Creek Generating Station, Unit 1 Inspection At: Hancocks Bridge, New Jersey Inspection Conducted: January 27, 1986 - February 7,1986 3 /7/N

' Inspector L. Briggs ead Reactor Engineer

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D. FloFek, Lead Reactor Engineer

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W % uu 31blw M. Evans, Reactor Engineer

' date Approved by:

d U Eselgrot Chief, Test Programs Section, da'te 08, DRS Inspection Summary:

Inspection on January 27, 1986 - February 7, 1986 (Inspection Report Number 50-354/86-10)

Areas Inspected: Routine, unannounced inspection (145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br />) by three region-based inspectors of licensee actions on previous inspection findings, preoperational test results evaluation review, preoperational test witnessing, power ascension test program, fuel channel shims, QA/QC interface with preoperational test program, plant tours and independent verification.

Results: Two violations were identified.

(paragraph 3.3).

NOTE: For acronyms not defined refer to NUREG-0544 " Handbook of Acronyms and Initialims."

9604040074 e60327 PDR ADOCK 05000354 G

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DETAILS 1.0 Persons Contacted l

_#A. Barnabei, Principal Quality Assurance Engineer

• E. Barclay, Lead Quality Control (QC) Engineer

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• V. Blenx, Assistant Project Manager G. Chew, Power Ascension Technical Support

-#*R. Donges, Lead Quality Assurance-(QA) Engineer

1. • J. Duf fy, Site Engineering Supervisor

• N. Dyck, RCT Chairman

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1. • M. Farschon, Power Ascension Manager

• J. Fisher, QC Supervisor

1. • A. Giardino, Manager, Station QA

1. • W. Goebel, Qa Engineer

1. R. Griffith, Principal QA Engineer l

1. A. Indigo, Startup Group l

1. • C. Jaffee, Startup Engineer

• P. Kudless, Maintenance Manager

• 0. Lynn, Safeteam Manager I

• M. Metcalf, Principal QA Engineer

1. G. Moulton, Principal QA Engineer

1. J. Nichols, Technical Manager L

• 1. R. Salvesen, General Manager i

• W. Schell, Power Ascension Test Director L. Zull, Lead STD & A Engineer l

1. Indicates those present at interim exit meeting on January 31, 1986.

• Indicates those present at final exit meeting on February 7, 1986,

i Other NRC Personnel

1. • D. Allsopp, Resident Inspector

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1. • R. Borchardt, Senior Resident Inspector

1. • J. Lyash, Resident Inspector

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• J. Strosnider, Section Chief, Division of Reactor Projects, Region I

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l The inspectors also contacted other personnel of the licensee's i

operating, technical and QA/QC staff.

2.0 Licensee Actions on Previous Inspection Findings (0 pen) Unresolved Item (354/86-03-02) This item deals with having practical BWR reactor'engin ering experience in the reactor engineering staff.

The inspector held discussions with the technical manager who informed the inspector that an experienced BWR reactor engineer was being hired and will be on board in mid-February. A review of the resume for t

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the individual indicated substantial practical BWR reactor engineering experience. The Technical Manager indicated that the duties and.

responsibilities of this individual would be as a technical advisor to the reactor engineering staff. This satisfied the inspectors concerns but the_ item will remain open pending the actual' employment of the individual on site.

(0 pen) Circular (354/80-CI-21) This circular deals with the duties and responsibilities of_ Senior Reactor Operator during refueling operations.

The inspector reviewed SA-AP.ZZ-049(Q) " Conduct of Fuel Handling and Core Alterations," Revision 0 dated January 21, 1986.

Refueling crew responsi-bilities were defined for all the members of the crew except the Senior Reactor Operator. Pending addition of the duties and responsibilities of the Senior Reactor Operator during refueling to the procedure, this item-will remain open.

(Closed) Unresolved Item (354/85-41-01), Licensee to perform speed control checks for hoists and trolleys during preventive maintenance of the polar crane. The licensee issued Revision 1 to procedure MD-PM.KF-004(Q), Polar Crane Preventive Maintenance and Check Out, on January 24, 1986, to test main and auxiliary trolley and hoist speed controls. The inspector reviewed the revision to the procedure and verified that trolley and hoist speed controls are adequately checked.

This item is closed.

3.0 Preoperational Test Results Evaluation Review 3.1 Scope The completed test procedures listed below were reviewed during this inspection to verify that adequate testing had been conducted 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 procedure, acceptance criteria, performance verification, recording conduct of test, QC inspection records, restoration of system to normal after test, independent verification of critical steps or parameters, identification of personnel conducting and evaluating test data, and verification that the test results have been approved.

PTP-BB-4, Reactor Pressure Vessel Internal Vibration Test,

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Revision 0, Results Approved November 20, 198.

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PTP-GT-1, Drywell Ventilation, Revision 1,'Results Approved

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January 17, 1986.

PTP-PN-1, Class IE 120 VAC Power and Non-Class IE, Revision 1,

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.Results Approved January 10, 1986.

PTP-SC-1, Loose Parts Monitor, Revision 0, Results Approved

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January 25, 1985.

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PTP-SE-1, Source Range Neutron Monitoring System, Revision 1, Results Approved December 20, 1985.

PTP, SN-1, Automatic Depressurization System, Revision 0, Results

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Approved January.4, 1986.

DTP-SB-0008, Reactor Vessel High Steam Dome Pressure Scram Response

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Time Test, Revision 0, Preliminary Results Review.

3.2 Discussion PTP GT-1, Drywell Ventilation

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During review of PTP GT-1 the inspector had several questions concerriing how testing was conducted. The completed procedure was very confusing with numerous test exceptions, procedure change notices (CN) and on-the-spot (OTS) changes.

Several examples of problems observed were:

(A) OTS 11, 12 and 13 issued in January 1986 to cover retests performed during June 1985 (licensee identified and correct-ed) to comply with Startup Administrative Procedure (SAP) 24.

(B) Two mandatory witness points (MWP) not signed during per-formance of CN-16 (Paragraphs 8.17.73 and 8.17.75). This was identified during QA review of the completed procedure (prior to preoperational review committee (PORC) approval).

The QA-inspector required issuance of OTS 18 which deleted all previous CN-16 testing and retested the portions of the system previously tested by CN-16. MWP's were completed during OTS-18. The inspector could not easily determine the above without discussion with the QA inspector and review of QA Surveillance Report (QASR) 3136 since CN-16 did not indicate that it had been replaced by OTS-1.

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(C) One additional MWP was not observed or signed by QA during the performance of the test (step 8.1.20).

This was not identified and corrected or retested in accordance with SAP 24.

PTP-SN-1, Automatic Depressurization

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During review of PTP-SN-1, Automatic Depressurization System, the inspector had several. questions and observations concerning the conduct of testing as follows:

(A) Test exception 53 issued on January 2, 1986, stated that the acoustic monitors had not been installed on the SRV's.

However, section 8.24 of PTP-SN-1, in which each acoustic monitor is mechanically excited (by hitting the pipe with a hammer or using a shaker table) was signed off as being successfully completed on November 5, 1985. After a dis-cussion with the system test engineer, the inspector deter-mined that each acoustic monitor had not been tested in Section 8.24. Instead, the test engineer used one acoustic monitor and connected it to each of the 14 acoustic monitor.

cables. The above was accomplished without a change' notice being issued to' revise the method of testing.

In addition, SDR-SN-0083 was issued on December 27, 1985 to verify installation of the acoustic monitors. The inspector noted that the SDR did not require a retest to verify proper operation of the monitors following installation.

SDR-SN-0083 has not been closed and subsequent to NRC questioning the. licensee is reevaluating the need for retesting to en-sure proper functioning of the acoustic monitors following installation.

(B) A MWP was deleted without QA approval during retesting on December 17, 1985, of step 8.26.2.10 of CN-017, ADS Accumu-lator Inlet Check Valve Leakrate Test.

(C) The inspector noted that pages 17, 18 and 19 of OTS-020 were missing from PTP-SN-1. Af ter questioning the licensee, the inspector determined that QA during review of the completed procedure (prior to PORC approval) had identified the miss-ing pages. However, no subsequent action was taken to either retest or locate and replace the missing pages.

DTP-SB-0008, High Steam Dome Pressure Scram

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The inspector conducted a preliminary review of DTP-SB-0008, which has been completed, but not yet reviewed and approved by the licensee. The inspector independently verified the response

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times for various relay logic actuations by reviewing strip-chart recordings. Several of the response times did not appear to meet the G.E. acceptance criteria of 50 milliseconds.

fhe test engineer issued an SDR requesting G.E. to evaluate the response times. During a subsequent inspection, the inspector will review the approved test results and G.E.'s evaluation to assure overall system response times are met.

3.3 Findings Items (A) and (B) identified during review of SN-1 and item (C)

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identified during review of GT-1 constitute.a violation of 10 CFR 50 Appendix B, Criterion V and SAP 24, Preoperational Test Procedure, Format and Instructions (354/86-10-01).

The above items and item (C) of PTP-$N-1 review constitute an

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additional violation of 10 CFR 50 Appendix B, Criterion XI, when the licensee did not perform an adequate review of test-results which resulted in failure to identify the procedur_e deficiencies (354/86-10-02).

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Retesting of acoustic monitor sensors being installed per SDR

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SN-0083 (item A of SN-1 review) is an unresolved item pending licensee determination of required retesting and subsequent results review by NRC (354/86-10-03).

In addition to the above, several-open test exceptions require

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resolution by the licensee. The test exceptions listed below collectively constitute unresolved item 354/86-10-0.

Procedure No.

Short Title SDR No.

PTP-PJ-1 250 VDC Class 1E PJ-0026, 0033 RJ-0129.

PTP-GT-1 Drywell Ventilation GT-0233, 0231, 0237, 0255.

PTP-PN-1 Class IE 120 VAC PN-0182, 0184, 0188.

PTP-BB-4 RPV Internal Vibration 88-0220, 0867 0600, 0942.

PTP-SC-1 Loose Parts Monitoring SC-0012, 0030, 0032, 0033, 0034, 0035, 0040, 0039.

PTP-SE-1 SRM SE-0149, 0079, 0092, 0201.

PTP-SN-1 ADS SN-0043, 0085, 0072, 0070, 0079, 0080, 0083, 0084, 0086, 0087, 0088.

The inspector also reviewed open test exceptions noted in unresolved item 354/86-03-01, and closed out four open test exceptions. Unre-

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solved item 354/86-03-01 is closed and the remaining test exceptions

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are included in unresolved item 354/86-10-04 above.

4.0 Preoperational Test Witnessing 4.1 Scope Testing witnessed by the inspector included the observations'of overall crew performance stated in Paragraph 3.0 of Inspection Report 50-354/85-18.

4.1.1 PTP KP-1 MSIV Sealing During the time of observance the test engineer was in the process of initiating a procedure change notice to facilitate testing.

The inspector reviewed the completed portions of the procedure and all test exceptions. No discrepancies were observe _

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PTP SF-1B Rod Sequence Control System (RSCS)

The inspector witnessed a portion of the RSCS rod block logic checks and discussed open test exceptions with the test engineer. Testing was being conducted in accordance with the attributes of Paragraph 3.0 of Inspection Report 50-354/85-18.

Only three minor test exceptions were identified by the test'

engineer.

4.2 Findings No unacceptable conditions were identified.

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5.0 Power Ascension Test Program (PATP).

5.1 References Regulatory Guide 1.68, Revision 2, August 1978, " Initial Test

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Programs for Water-Cooled Nuclear Power Plants" i

ANSI N18.7-1976, " Administrative Controls and Quality. Assurance

for the Operational Phase of. Nuclear Power Plants"

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Hope Creek Generating Station (HCGS) Technical Specifications,

Proof and Review Copy HCGS Final Safety Analysis Report (FSAR, Chapter 14, " Initial a

Test Program")

HCGS Safety Evaluation Report, Chapter 14, " Initial Test

Program"

Station Administrative Procedure, SA-AP.ZZ-036, Revision 1,

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" Phase III Startup Test Program"

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Specification NEB 0 23A1137, Revision 0, " Hope Creek Startup'

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Test Specification" HCGS Power Ascension Test Matrix, Revision 3

5.2 Startup Test Procedure Review 5.2.1 Scope The startup test procedures of Attachment A were reviewed for the applicable attributes listed in Inspection Report 50-354/86-03 Section 4.3.

In addition the procedure related to the Main Steam isolation valves was reviewed for MSIV closure time limits, ap-propriate initial conditions and that all valves were tested.

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The procedures related to the high pressure coolant injection system (HPCI) and reactor core isolation cooling (RCIC) were reviewed to assure acceptance. criteria for. flow time. response, speed response and detection of steam leakage and that appro-priate ' initial conditions were specified.

5.2.2

' Discussion The licensee was in the process of revising Revision 0 of the issued procedures based on lessons learned during use of these procedures ~ while training test engineers on the simulator and to assure consistency among the procedures with the administrative program requirements. Wherever possible the inspector reviewed the revision 1-procedure for technical adequacy. Procedures TE-SU.ZZ-172, ZZ-173, ZZ-175, ZZ--179, AE-236, BG-701, BG-702, BG-703 and GT-721 were reviewed for agreement with commitments on acceptance criteria. Unless discussed below, the procedures were consistent with the above attributes. The items below will be reviewed again in subsequent inspections.

HPCI and RCIC The HPCI and RCIC procedures were reviewed for technical adequacy.

The licensee agreed to include the following items in the appropriate procedure.

The licensee plans to complete the RCIC testing during test condition heatup. As a result precautions'will be added to the procedure to assure that the 5'.' reactor power levels would not be exceeded during the RCIC testing.

In addition the minimum and maximum power levels to safely perform the RCIC vessel injection testing will be specified.

MSIV Testing The inspector noted differences in the test acceptance criteria between the proof and review copy of Technical Specifications and the GE test specification with regard to MSIV stroke times. A change in the GE test specification is being prepared to correct this item.

In addition the test was planned to be conducted at 600 psig rather than at rated pressure as. indicated in Regulatory Guide 1.68.

This item was also in the process of being changed to agree with the Regulatory Guide.

5.2.3 Findings No violations were identifie.

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5.3 Overall Power Ascension Test Program 5.3.1 Scope The inspector utilized the Power Ascension Test Matrix, Revision 3, and compared the licensee program with Regulatory Guide 1.68.

The inspector also reviewed the Level 1 Power Ascension Schedule, Revision 0, to assess the logic utilized in the testing program.

The inspector focused on the heatup phase of testing.

The inspector also utilized selected procedures from Attachment A to ascerta'in whether these procedures addressed Regulatory Guide 1.68 testing commitments. The inspector will assess the licensee program against Regulatory Guide 1.68 frequently in subsequent inspections as the program progresses.

5.3.2 Discussion The licensee power ascension testing program was found to be consistent with Regulatory Guide 1.68 except for the following:

Regulatory Guide 1.68 section 4g indicates that proper re-

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sponse of process and effluent radiation monitors is demon-strated based on analysis of samples from the process and or effluent system during low power testing. The licensee did not have such a test planned but considered preop data could satisfy this item. The licensee subsequently indi-cated one would be prepared.

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Item 41 indicated MSIV stroke tesitng is performed at rated temperature and pressure. The licensee procedure indicated this would be performed at 600 psig. As noted in the procedure review section, this was identified and is in the process of being corrected by the licensee.

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Item 5ff indicates that important ventilation systems should be demonstrated to maintain their service areas within design limits at 50% and 100% power. The licensee procedures only addressed primary containment and steam tunnel temperatures. The inspector noted that FSAR Section 9.4.2.1.c contains limits for important ventilation systems that did not appear to be addressed.

The PATP Technical Director noted that licensee plans for this testing were being developed and would be performed.

The above items will be reviewed in subsequent inspections.

During review of the PATP schedule, the inspector questioned the logic of performing control rod scram time testing at rated pres-sure in parallel with RCIC and HPCI testing. The. inspector was concerned that the HPCI and RCIC dynamic testing could impose a reactor scram and the control rod scram response should be veri-fled prior to the demanded scram.

The licensee representative indicated that sufficient control rod drive scram data would be obtained prior to the transient type testing to demonstrate ac-ceptable response of the control rods if a scram was automatical-ly demanded. The inspector noted the licensee resnonse and in-dicated that the specific testing based on a more refined test plan would be assessed in a subsequent inspection.

5.3.3 Findings No violations were identified within the scope of this review.

6.0 Fuel Channel Shims In a previous inspection (50-354/86-03) the inspector inquired concerning the observed shim plates near the top of the fuel channel and stated at that exit that during a subsequent inspection the inspector would pursue the vibration program impact.

The inspector reviewed fuel channel shims again in this inspection.

The inspector was provided a copy of drawing 829E488 revision 4 sheet 2 which illustrates the use of shim plates approximately 2 inches by 4 inches by 93 mils thick affixed to the side of the fuel channel opposite the channel spacer buttons.

The inspector identified several areas of concern; structural, thermal hydraulic,

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vibratory, and loose part potential.

In addition the inspector inquired if the NRC was notified of these shim plates since inspector preparation for the inspection could not find a reference to the shim plates in the FSAR or documents referenced therein.

Just before the interim exit meeting on January 31, 1986 the inspector was informed that the shim (adaptor) was a 1974 design change implemented by the vendor in accordance with his QA program and there was no safety impact.

Further the NRC had not been notified due to the level of design detail.

A site technical contact was also identified. At the interim exit the in-spector indicated that further discussion would be required.

In a telecon on February 5,1986, the inspector indicated that the 10 CFR 50.59 review and relevant technical information material would be reviewed; the licensee-requested that this be performed during the inspector's next site visit.

This was acceptable to the inspector.

7.0 QA/QC Interface with Preoperational Test Program The inspector reviewed recent QA surveillance reports-(QASR) regarding different activities of the licensee's startup group.

The following QASR's were reviewed:

QASR-7435, Witnessing of portions of Detailed Test Procedure (DTP)

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58-0008, conducted on January 22, 1986.

The QA inspector witnessed five mandatory witness points. All results were satisfactory.

QASR-5815, Witnessing of a portion of PTP-SN-1, conducted on November

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23, 1985.

The QA inspector witnessed the ADS accumulator inlet check valve leakrate tests for valves 1-SN-V045 and 1-SN-V114.

The actual leakrates were within test specifications.

QASR-7210, Witnessing of torquing, termination and slow / fast test for

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eight Rosemont Transmitters per DTP-SB-0113.

Surveillance conducted on January 16, 1985.

Results were satisfactory.

QASR-5943, Verification of the disposition of SDR-BJ-0429, conducted

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on November 23, 1985. The QA inspector witnessed the torque switch installation and greasing of the valve. All work was satisfactorily completed.

7.1 Findings No unacceptable conditions were noted.

8.0 Plant Tours Discussion The inspector made several tours of various areas of the facility to observe work in progress, housekeeping, cleanliness controls and status

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of construction and preoperational test activities. The inspector noted a distinct improvement in plant housekeeping in several areas of the facility.

8.1 Findings No violations were observed.

9.0 Independent Verification The inspector independently verified the response times for DTP-SB-0008 as discussed in paragraph 3.2 of this report.

9.1 Findings No unacceptable conditions were identified.

10.0 Unresolved Items Unresolved items are matters which require additional information in order to determine if they are acceptable, violations or deviations.

The unre-solved items identified during this inspection are discussed in Paragraph 3.3 of this report.

11.0 Exit Interview A management meeting was held at the conclusion of the inspection on Febru-ary 7, 1986 to discuss the scope and findings as detailed in this report (see Paragraph 1 for attendees).

In addition, an interim exit meeting was held with the licensee on January 31, 1986. No written information was provided to the licensee at any time during the inspection.

The licensee did not indicate that proprietary information was contained within the scope of this inspectio s

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ATTACHMENT A Startup Test Procedure Review TE-SU.SE-121 APRM Calibration During Heat-Up Revision 0

.TE-SU.BD-141 RCIC CST Initiation, Revision 1, Draft TE-SU.BD-142 RCIC Vessel Injection, Revision 1, Draft

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TE-SU.BD-144 RCIC System Cold Quick Start to the Reactor Pressure Vessel, Revision 1, Draft TE-SU.BD-145 RCIC Surveillance Test Demonstration Revision 1, Draft TE-SU.BJ-151 HPCI CST Injection, Revision 1, Draft TE-SU.BJ-154 HPCI Surveillance Test Demonstration Revision 1, Draft TE-SU.ZZ-162 Water Level Measurement Test, Revision 1, Draft TE-SU.ZZ-172 NSSS Piping Expansion - Walkdown, Revision 0 TE-SU.ZZ-173 NSSS Piping Expansion - Sensor Readings During Initial Plant Heatup, Revision 0

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TE-SU.ZZ-175 B0P Piping Expansion - Data During Initial Plant Heatup, Revision 0 TE-SU.ZZ-179 Gaseous Radwaste Piping Expansion During Operation, Revision 0 TE-SU.AE-236 Feedwater Startup Controller Tests, Revision 0 TE-SU.AB-251 MSIV Functional Test, Revision 0 TE-SU.BG-701 RWCU - Normal Mode Performance Test, Revision 0 TE-SU.BG-702 RWCU - Blowdown Mode Test, Revision 0 TE-SU.BG-703 RWCU - Hot Standby Mode Test, Revision 0 TE-SU.GT-721 Drywell and Steam Tunnel Cooling System - Normal Operation Performance, Revision 0