IR 05000322/1987013

Safety Insp Rept 50-322/87-13 on 870601-05.No Violations Noted.Major Areas Inspected:Startup Test Program,Including Test Results Evaluation,Test Witnessing,Qa/Qc Interfaces, Independent Insp & Tours of Facility

ML20235G548
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	06/23/1987
From:	Briggs L, Florek D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20235G531	List: IR 05000322/1987013 ML20235G529
References
50-322-87-13, NUDOCS 8707140292
Download: ML20235G548 (7)

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-13 ) Docket N License N NPF-36 Licensee: Long Island Lighting Company 175 East Old Country Road Hicksville, New York 11801 Facility Name: Shoreham Nuclear Power Station Inspection At: Shoreham, New York Inspection Conducted: June 1-5, 1987 Inspectors: L".

[ Briggii L M-f6 Reattor Engineer

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t, ~/$ DJ Florek, Chief, Test Program Section i date / 08, DRS ' Inspection Summary: Inspection on June 1 4 1987, (Inspection No. 50 * 2/87-13) Areas Inspected: Routine unannounced inspection u." startup test program including the overall program, test results evaluation, cest witnessing, QA/QC interfaces, independent inspection and tours of the facilit Results: -No Violations were identifie Note: For acronyms not defined refer to NUREG-0544 " Handbook of Acronyms and Initialisms" B707140292B{$[$$22 PDR ADOCK ppg G

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DETAILS 1.0 Persons Contacted

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J. Alexander, Operating Engineer A. Arena, Nuclear Systems Engineer R. Bonitch, Watch Engineer J. Cocuzzo, Quality Control Inspector

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R. Crowe, Operation Staff Manager M. Giannattasio, Nuclear Systems Engineer

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G. Gisonda, Supervisor, Nuclear Licensing

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R. Glazier, Quality Control Supervisor

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R. Grunseigh, Operational Compliance Engineer

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A. Himle, Power Ascension Test Program (PATP) Test Coordinator

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C. Losnedahl, Reactor Engineer (Acting)

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D. Smith, Compliance Engineer

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W. Steiger Jr. , Plant Manager 2.0 Startup Test Program 2.1 References

  - Shoreham Nuclear Power Station (SNPS) Final Safety Analysis Report :
  - SNPS Safety Evaluation Report
  - Regulatory Guide 1.68 " Initial Test Program For Water Cooled Reactor )

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  - SP-12.075.01 " Administration of Startup Testing"  )
  - ANSI N18.7-1976 " Administrative Controls and Quality Assurance for I the Operational Phase of Nuclear Power Plants"
  - General Electric Startup Test Specification No. 22A5727, Shoreham i MPL item No. A41-3610    J 2.2 Overall Startup Test Program The inspector discussed the PATP status with the Test Coordinator and activities necessary to begin Test Condition 1 (TC1) testin TC1 testing is up to 25 percent of rated oower. It was indicated  1 that following successful testing at the current test condition  l heatup (up to 5 percent power) and plateau test results approval the j plant would essentially be ready for the next test platea .3 Startup Test Results Review Scope The test results of Attachment A were reviewed against the following I attributes. Procedures were reviewed to assess that each was approved in accordance with the administrative procedure, test changes were noted if appropriate, test objectives were met, test

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l exceptions were noted if appropriate and were processed in accordance ) with the administrative procedures, data sheets were completed as required, test steps and data sheets were properly signed and dated, test data received engineering evaluation, test results were compared with established acceptance criteria, review and acceptance of test results is documented, test results receive QA or independent review, and test results were approved by appropriate managemen In addition the inspector reviewed the Plateau Summary Report for test condition heatup (August 1986). Test Review Committee (TRC) review was completed on March 25, 1987. This review included assessment of the testing performed as of that time, test results, test excebtions and tests to be deferred until a later dat Review of Operations Committee (ROC) review had not been completed for this most recent Plateau Summary Repor Discussion The test results were found to satisfy the test objective Identified test exceptions were being processed in accordance with the administrative procedure. No additional test exceptions were identified based on inspector review. The plateau summary for the prior testing ione in test conditions heatup (July 3-October 3,1985) was reviewed du ing Inspection 50-322/86-18 and was determined to satisfy the FSAR snd license commitment The licensee's ROC wall perform a final review of the test results of the August 3, throug;. August 31, 1986 test plateau and the TRC and the ROC will review the current Test Condition Heatup summary report prior to beginning Test Condition Findings No violations were identifie .4 Startup Test Witnessing Scope The inspector witnessed a portion of the tests discussed below. The tests were witnessed and compared to the attributes identified in Inspection Report 50-322/85-29, Section Discussion On June 3, 1987 at about 9 p.m. the licensee made final preparations to run Section 8.2 of STP-15, Revision 8, HPCI System. Original plans were to run the HPCI System CST to CST to perform HPCI controller checks and tuning if necessary. In addition, SP 46.009.03, " Motor Operated Valve Dynamic Test" was to be performed on E41*MOV038. This test would time the stroking of MOV038 (HPCI discharge return to CST) _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _

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. with full system differential pressur Subsequent to the above testing the licensee planned to secure the HPCI pump and cool down the suppression pool in preparation for the HPCI endurance run portion of STP-1 The above testing also gathered some preliminary information and data concerning expected conditions for the upcomming endurance run such as suppression pool temperature increase versus time and stable lube oil temperatures (expected values).

Controller tuning checks proceeded without any problems; however, when the licensee attempted to stroke E41*MOV038 it would not stroke against the system differential pressure (about 1200 psid). The licensee attempted to close the valve for about 140 seconds, the expected stroke time was about 80 seconds. Valve motor Operator phase current was greater than 50 amperes (range of clamp on ammeter), the expected value was 24 an. pares nominal. Subsequent to the above failure the licensee terminated testing until the motor operator  ; problems could be further evaluate Prior to this test E41*MOV037, an upstream valve in series with MOV038, had experienced a similar problem (NRC Inspection Report 50-322/85-31, paragraph 3.1), The HPCI pump had been run a total of 60 minutes with a rise in the average temperature of the suppression pool of 30 degrees, lube oil temperatures stabilized at about 135 F. Motor operated valvas are further discussed in Paragraph 4.0 of this report. HPCI endurance testing was completed on June 5 (5:10 to 6:01 a.m.) and MOV 37 stroking was attempted without success (see Paragraph 4.0). , Findings No violations were identifie .0 QA/QCInterface The inspector observed QC personnel conducting surveillance of the HPCI (STP-15) startup test being conducted. The inspector also discussed other QC activities with the QC inspecto No unacceptable conditions were observe .0 Independent Inspection Activity During the review of test results the inspector independently verified the results of completed startup tests and acceptability of licensee test exception resolutio In addition to the above the inspector followed up on valve operator problems discussed in Paragraph 2.4 of this report and licensee act'on to , resolve them. Information obtained via discussion with the licensee and preliminary document review indicated that the upstream HPCI test valve E41*MOV037 had experienced problems stroking against greater then 1000 psid in August 1985. At that time a mair.Lenance work request (HWR 85-4911)

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was issued to repair the valv E41*MOV037 is a Velan 8 inch globe valve with a limitorque 125VDC SMB-1-25 motor operator. According to the valve manufacturer the motor operator was changed from an SMB-3-40 limitorque operator in 1976. The vendor's calculations (Velan No. 072-11-76, Rev 1) indicate that the smaller operator would be able to operate the valve at up to 1475 psid and also protect the valve from over torque damag During this inspection it was determined via licensee discussions with the vendor, architect engineer and NSSS supplier that the MCC starter for MOV 037 was a size 2 reduced voltage starter retained from the original design when the larger motor operator was to be used. It was decided to modify the starter to make it an across the line starter and increase the operator's torque. This modification was performed on June 4, 1987 under Engineering and Design Coordination Report (E&DCR) No. L-1471. A safety evaluation was also performed and attached to the E&DCR. This modification was tested at 3:53 a.m. on June 5, 1987. Initial stroke testing was conducted with MOV 037 and 038 closing at the same time. In this configuration both valves stroked fully closed. Then ooth valves were opened and the licensee attempted to stroke MCV 037 by itself. At that time MOV 037 drew excessive overcurrent and trippe During the June 3, 1987 test observed by the inspector the downstream HPCI test valve E41*MOV 038 also failed to stroke against system differential pressure, see Paragraph 2.4. This valve is a Velan 10 inch gate valve with a limitorque 125 VDC SME-3-40 motor operator. This valve is not used to throttle HPCI test flow to the CST. It is however, designed to be able to close under full system differential pressure. This item will be further evaluated as licensee corrective actions progres On June 5,1987, the licensee decided to postpone shutdown of the plant to allow engineering to instrument the MOV 037 and 038 MCC's and perform further stroke testing to gather data. The licensee's final corrective i' action will be based on the result of the data gathered from those test On June 1, 1987 the inspector also witnessed licensee actions during a ! heatup. The reactor pressure was being increased to 350 psig in 1 anticipation for an increase to normal operating pressure (NOP) and temperature later that evening. Two items were noted by the inspector and ; found to be in accordance with (IAW) Technical Specifications (TS).

- The RWM was bypassed. A second individual was present verifying rod i movement in accordance with TS 3.1.4. The RSCS had 2 rods bypassed. Allowed by TS 3.1.4.2 b.1 and b Less than 3 rods per RSCS group and second party verification of { bypassed rod j j No violations were identifie ,

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5.0 Facility Tours In addition to numerous tours of the Control Room the inspector also toured the facility to assess plant and equipment conditions including cleanli-ness, apparent physical condition, fire hazards and work in' progress. The inspector noted that the reactor building appeared to be in very good condition with no discrepancies observed. The inspector had several questions concerning observations in the turbine building. These items were discussed with the licensee and found to be in compliance with licensee practices and procedure i No violations were identifie .0 Exit Interview At the conclusion of the site inspection, on June 5,1987, an exit interview was conducted with the licensee's senior site representatives (denoted in Section 1). The findings were identified and inspection items were discusse 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 net contain information subject to 10 CFR 2.790 restrictions.

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ATTACHMENT A Inspection 50-322/87-13 1986 TC Heatup and Retest SVT Results Review l TRC Mt i Number Name Section Performed Approved STP-1 Chemistry and Radiochemistry of Reactor Coolant STP-5 Control Rod Drive 8.3(950psig) 86-16 STP-9 Water Level Measurements l STP-13 Process Computer 8.1 ."d STP-14 RCIC System Test 8.3.1,2,3,4 and 5 86-16 STP-15 HPCI System Test 8.1.1,2 and 3 86-18 8.2.1 and 8. STP-37 Reactor Bldg. Closed Loop 8.1, 8.2 and > Cooling and Drywell Cooling STP-70 Reactor Water Cleanup 8.1 and Appendix A 87-02

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