IR 05000443/1986050

Insp Rept 50-443/86-50 on 861020-30.No Violations Noted. Major Areas Inspected:Initial Fuel Loading Witnessing

ML20214V011
Person / Time
Site:	Seabrook
Issue date:	11/28/1986
From:	Florek D, Hunter J, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20214U970	List: IR 05000443/1986050 ML20214U966
References
50-443-86-50, NUDOCS 8612090489
Download: ML20214V011 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-50 Docket N License No. NPF-56 Licensee: Public Service of New Hampshire P. O. Box 330 Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit-1 Inspection At: Seabrook, New Hampshire Inspection Conducted: October 20 - 30, 1986 Inspectors: hdv Cb P. C. Wen, Reactor Engineer LIfab!N date D

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J. G. Hunter, Reactor Ehgineer

& \\f1l,l9b date Approved by: l. <

D. J. Florek, Chief, Test Programs kI

/ datle Section, OB, DRS Inspection Summary: Inspection on October 20 - 30, 1986 (Inspection Report No. 50-443/86-50).

Areas Inspected: Initial Fuel Loadin; Witnessin Results: No violations were identifie NOTE: For acronyms not identified, refer to NUREG-0544, " Handbook of Acronyms and Initialisms."

0612090489 861203-PDR ADOCK 05000443 0 PDR

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DETAILS New Hampshire Yankee (NHY)

• D. Abely, Maintenance Department Supervisor J. Burson, Shift Test Director B. Couture, Reactor Engineer

• R. Cyr, Maintenance Manager R. Ferrell, Licensing Coordinator

• G. Grillo, Assistant Operations Manager B. Gwinn, Shift Test Director

• G. Kann, Test Group Manager

• G. Kline, Technical Services Manager

*J. Marchi, Startup QC Manager A. Merrill, Reactor Engineer

• D. McLain, Startup Manager

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• D. Moody, Station Manager

• L. Rau, Reactor Startup Supervisor

• Temple, QC Inspector

• C. Vincent, QC Supervisor

• L. Walsh, Operations Manager

• J. Warnock, Nuclear Quality Manager Yankee Atomic Electric Company (YAEC)

W. Middleton, QA Staff Engineer

• J. Singleton, Special Projects QA Manager

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United States Nuclear Regulatory Commission (U.S.NRC)

• R. Barkley, Resident Inspector-Indian Point III

• A. Cerne, Senior Resident Inspector

• D. Ruscitto, Resident Inspector The inspectors also contacted other administrative and technical licensee personnel during the course of the inspectio * Denotes those present at the October 30, 1986 Exit Meetin .0 Fuel Loading 2.1. Chronological Events On October 17, 1986, the NRC issued a license authorizi,g fuel loading and non-nuclear testing at the Seabrook-1 Nuclear Power Plant. On October 22, 1986 at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, the licensee commenced fuel loading activities per startup test procedure 1-ST-4, " Initial Core Loading," Revision Core loading was completed on 0712 hours0.00824 days <br />0.198 hours <br />0.00118 weeks <br />2.70916e-4 months <br />

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October 29, 198 ..

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The licensee performed core loading on the swing and midnight shift The NRC inspectors provided fuel loading coverage during the entire core loading period. The chronological events were as follows:

October 22, 1986 1600 Commenced Initial Core. Loading (1-ST-4). Prior to the fuel loading the following were verified: (i) core loading prerequisites (1-ST-3) was met, (ii) a response check for all source range channels (N31, N32 and 3 tempo-rary detectors was performed satis-factorily, and (iii) no stratification existed within reactor vessel and that the boron concentration in the reactor vessel was greater than the required 2000 pp Core loading operations began with the insertion of primary source fuel assembly (FA) C04 into core Position L-1 The second primary source FA C30 was loaded into core position G-0 October 23, 1986 0705 Suspended fuel loading at the completion of Step 18 of 1-ST- Recommenced fuel loading at Step 19 of 1-ST- October 24, 1986 0738 Suspended fuel loading at the completion of Step 55b of 1-ST- Recommenced fuel loading at Step 55C of 1-ST- In the process of relocating primary source FA C30 into the new core posi-tion G-14, a high flux at shutdown alarm from N32 was received. Contain-ment evacuation was carried out accord-ing to test procedure 1-ST-4. It appeared that the transferring path was too near to the source range detector N32. The increase in N32 response exceeded its high flux shutdown alarm l

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setpoint which was set at relative low value (10 CPS). Based on the 1/M plot from all source range channels, sub-criticality existe .

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2015 FA C58 was loaded at core position F-15 where its location was near the source range detector N32. About 10 minutes after completing this step, a high flux at shutdown alarm from N32 was received. Prior to this event, N32 had been reading 7-8 cps. The alarm was caused by the statistical variation of N32 reading exceeding the alarm set-point of 10 cps. Based on the 1/M plot from all source range channels, sub-criticality existed. The inspector witnessed that all personnel were orderly evacuated from the containmen October 25, 1986 0724 Suspended fuel loading at the completion of Step 81 of 1-ST-4 1518 Recommenced fuel loading at Step 82 of 1-57-4 1540 Fuel loading was temporarily suspended due to an upender problem in the Fuel Buildin Resumed fuel loadin Fuel loading was temporarily suspended because valve CAP-V3 was found ope This resulted in a question on the adequacy of containment integrity. The inspector accompanied by the licensee Unit Shift Supervisor independently verified that the affected containment isolation valves were shut and tagged out. Subsequently, fuel loading was resumed at 0044 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> on October 26, 198 October 26, 1986 0115 At Step 88 of 1-ST-4, in the process of transferring FA B05 from the Fuel Build-ing to the upender, a Fuel Building bridge crane operator inadvertently depressed a wrong button and FA B05 bottom nozzle went slightly below the top of Spent Fuel Storage Rack. No contact of FA and Spent Fuel Storage Rack was observed, however, the abrupt

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motion caused the bridge crane to lock up due to exceeding its overload cutoff limit. This problem was evaluated by the licensee engineer and operation staffs prior to resuming normal opera-tio Experienced difficulty in loading FA B05 into core position C-10. Field change #6 was written to allow loading FA B05 into core position R-11, and later use a " boxing" scheme to load back to C-1 Suspended fuel loading at the completion of Step 94 of 1-ST- Recommenced fuel loading at Step 95A of 1-ST- Returned FA B05 back to the Fuel Build-ing for further inspection by the fuel vendor (Westinghouse) representative and Reactor Engineering Superviso Both parties concurred that no fuel damage had occurred on FA B05 as a result of the earlier incident (0115).

2055 FA B05 was loaded into core position C-10 without proble October 27, 1986 0358 Experienced problem on unlatching FA C17 into core position B-03. The computer used in the refueling machine crane position indication needed to be reset after periods of usag After performing computer initialization,

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FA C17 was successfully unlatched into

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core position B-0 Suspended fuel loading at completion of Step 118 of 1-ST-4 1710 Recommenced fuel loading at Step 119 of 1-ST- October 28, 1986 0659 Suspended fuel loading at completion of Step 155 of 1-ST-4

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1512 Recommenced fuel loading at Step 156 of 1-ST-4 October 29, 1986 0712 Core loading operations completed with loading FA C44 into core position L-15.

. 2.2 Fuel Loading Activities Witnessing Scope Throughout the entire fuel loading period, the inspectors verified that the following requirements were met:

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Continuous communication was being maintained between the control rcom and refueling station;

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Continuous monitoring of the source range channels and use of inverse count rate ratio calculations

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Boron concentration met TS required values and was sampled per operating procedure 05-86-1-6;

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Special valve check to prevent inadvertent boron dilution was performed correctly per procedure 0S-86-1-7;

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All assigned personnel were qualified;

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Fuel status tag boards were updated following each fuel move;

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Plant conditions were being maintained as required by TS;

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Shift change and test briefing were prc ly conducted;

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Access control to Containment and Fuei duilding had been established and was in plac Discussion Valve CS-V744 incident The licensee in accordance with facility operating license NPF-56, is required to maintain a boron concentration of at least 2000 ppm in the reactor coolant and makeup water syste.ns during fuel loading and pre-criticality testing. Special license conditions include mechanically locking and closing a number of valves to prevent water at boron concentration levels less than 2000 ppm from flowing into the reactor coolant system (RCS), and independent valve position

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verification to preclude inadvertent dilution for those occasions when valves are unlocked and manipulated to perform plant chemistry

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operations. The license also requires the licensee to verify, on a daily basis, that the required valves are chain locked closed, when not open for plant chemistry related activitie Station operating procedure OS-86-1-7, "Unborated Water Source Locked Valve List," is used to verify that valves providing potential paths of unborated water sources to the RCS are closed, chained, and locked as required by the special license conditions. During the surveil-lance performed on October 23, 1986, an auxiliary operator accom-panied by the NRC inspector identified that valve CS-V744 appeared to be locked open rather than locked closed even though the attached danger tag recorded the valve as closed. The licensee verified that the valve was open and proceded to close the valve and lock it. The inspector witnessed the closing, locking and tagging of the valv CS-V744 is a three-inch valve located in the Chemical and Volume Control System (CVCS) downstream of the demineralizer beds and pro-vides a flowpath to the spent resin sluice tan Upon discovery of the open valve, the inspector verified that redundant valves upstream of CS-V744 were in the closed positio A subsequent review of the flow path to the RCS revealed no real potential for an inadvertent baron dilution, because the letdown line was isolated, no pumps were rurning in the CVCS system, and the demineralizers were not " cut-in". The inspector also witnessed the subsequent RCS boron sampling and analysis which provided acceptable result The licensee initiated a site incident report and notified the NRC of the failure to meet a license condition. The licensee also initiated additional training of all auxiliary operators and some reactor operators to show the proper method of verifying that the ,

required valves were chain-locked closed. The inspector accompanied the shift superintendent and the operators during the performance of the surveillance on October 24, 1936 and witnessed the shift superintendent instructing the operator The resident inspectors are reviewing the origin of the problem, and the licensee's corrective actions. Any enforcement actions will be discussed in the resident's inspection report 50-443/86-4 Fuel Building Crane Operation On October 24, 1986 through direct observation in the Fuel Building, the inspector noticed that only one person handled the operation of lifting a FA from the fuel storage rac Because the FAs are new fuel, no water was in the fuel pool. As a result, no damping from water to minimize FA swing was provided as would be available during normal refueling operations. The fuel pool operation was further complicated by the single crane speed design. The fast stop usually

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gave some degree of FA swing toward the adjacent wall. In every instance, the operator had to use one hand to stop FA swing, while he used the other hand to control bridge / crane movemen The inspector expressed a concern of the operator depressing the wrong button under adverse conditions. This concern was dis-cussed with senior plant management, who immediately instructed the operation be improved by having two persons on the bridge during lifting of FA from the fuel storage rack. This practice was instituted continuously throughout the rest of the fuel loading perio Boron Concentration After verifying that no stratification existed within reactor vessel and that the boron concentration in the reactor vessel was greater than the required 2000 ppm, the RHR loop water was sampled and analyzed for baron concentration at regular intervals during the core loading period. The inspector verified through control room log review and direct observation of the chemistry technician performing boron sampling analysis that boron concentration was maintained between 2091 - 2100 pp On October 23, 1986 the NRC inspector and a licensee technical QC inspector witnessed the chemistry technician perform the sampling of the RCS using procedure CX0910.01. The procedure required the technician to determine flow rate using FI-2856 and then utilize an enclosed flow rate versus time curve to determine the system flush time required to produce an adequate sample. Due to system isola-tion, FI-2856 was not in service and the technician calculated the required flush time using his knowledge of the system rather than utilizing the curve enclosed in the procedur The method used by the technician to determine the flush time was questioned by both the NRC and QC inspectors since it was not in accordance with the procedure or properly documented. The procedure, as written, did not allow for alternate methods of determining the required flush time. Both inspectors discussed the problem with the Chemistry Department Head and were informed that the procedure would be revised to provide guidance for those occasions when the flow indicators were not operable. The QC inspector wrote QC Surveillance Report 86-00856 to track and evaluate the corrective actions regard-ing the deficiencies of the sampling procedur The NRC inspector accompanied the chemistry technician on October 24, 1986 and witnessed the sampling of the RCS and verified that it was performed in accordance with CX0910.01. The licensee provided the inspector with the cnange made to the procedure that

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allows the technicians to use alternate methods of obtaining sample The inspector reviewed the procedure changes and had no further question FA B05 On October 26, 1986, at Step 88 of 1-ST-4, FA B05 was lifted from Fuel Building fuel storage rack W17, and was trolleyed along the north direction to row 'F' location to align it to the upende Once near 'F' location, the operator realized that he overtravelled beyond the 'F' marker and tried to return to 'F' location. However, I he accidentally depressed the " lower" hoist button and FA button nozzle was lowered slightly belew the top of fuel storage rac He then immediately pressed the stop button to correct the erro No FA/ fuel storage rack contact was observed. Subsequent fuel inspection performed indicated that no fuel or storage rack damage occurred as a result of this inciden l t

Core Verification After completion of fuel loading, the licensee performed core veri-fication. The as-loaded core using a TV camera and video recorder was compared with the intended core loading plan as described in the Westinghouse Nuclear Design Report (WCAP-10982). Through independent review performed by the reactor engineer and QC personnel, the core verification was determined satisfactoril No discrepancies were identifie Findings The licensee's fuel loading was accomplished in accordance with approved procedure (1-ST-4). Prior to fuel load the core loading prerequisites (1-ST-3) were complete Licensee performance throughout the entire fuel loading period was deliberate, and carefully controlled. RCS boron concentration and RCS temperature were consistently maintained within the required limits. Subcriticality monitoring (1/M plot) through all 5 source

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range detectors was continuously evaluated. At no time were loading operations interrupted as a result of unexpected changes in these evaluation After completing fuel loading, the as-loaded core was verified to be in agreement with the intended core loading pattern. No outstanding discrepancies or problems exis .- - . . _ - _ - _ _ - _ _ _ _ . _ . - - - - - _ ,

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3.0 Independent Calculations / Verifications The inspectors performed independent calculations / verifications during fuel loading period as followed:

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Independently verified that the subcriticality monitoring (1/M plot)

was calculated correctly on a sampling basi Independently verified that the loading steps were consistently in agreement with the test procedur .0 QA/QC Interface The inspector verified that startup test procedure 1-ST-4 had provisions for QA witnessing of certain test steps. Witness points were stamped in the master copy of the procedure and initialed by QA personnel. The inspector also noted that QC inspectors accompanied auxiliary operators performing the locked valve surveillance and chemistry technicians per-forming boron sampling activitie Throughout the entire fuel loading period, the inspector noted that QA/QC provided shift test coverage. The inspector also reviewed OC surveillance reports performed on a sampling basi No unacceptable conditions were identifie .0 Management Meeting Licensee management was informed of the scope and purpose of the inspection at an entrance meeting conducted on October 22, 1986. The findings of the inspection were discussed with licensee representatives during the course of the inspection. An exit meeting was conducted on October 30, 1986 at the conclusion of tne inspection (see paragraph 1 for attendees).

At no time during this inspection was written material provided to the licensee. Based on the NRC Region I review of this report and discussions held with the licensee representatives at the exit, it was determined that this report does not contain information subject to 10 CFR 2.790 restric-tions.