IR 05000412/1986040

Insp Rept 50-412/86-40 on 861117-21.No Violations Noted. Major Areas Inspected:Preoperational Test Program,Including Tests Conducted During Hot Functional Test,Qa Interface Activities & Unresolved Items Identified in Previous Insps

ML20212C271
Person / Time
Site:	Beaver Valley
Issue date:	12/16/1986
From:	Briggs L, Pullani S, Vankessel H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20212C217	List: IR 05000412/1986040 ML20212C211
References
50-412-86-40, NUDOCS 8612290436
Download: ML20212C271 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

t Report N /86-40 Docket N License No. CPPR-105 Category B Licensee: Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 Facility Name: Beaver Valley Power Station, Unit No. 2 Inspection At: Shippingport, PA Inspection Conducted: November 17-21, 1986 Inspectors: dub lgth I.2 fl6fW

' H. F. yhnKess , Reactor Engineer 'datt

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s Reactor Engineer l$bll$$

1. Briggs,' 'date Approved by: / daEj /.2.[/6/W

' Pull /ni, Act Chief 'da t'e Test P ection, 08, DRS Inspection Summary: Inspection on November 17-21, 1986 (Inspection Report j No. 50-412/86-40)

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Areas Inspected: Routine unannounced inspection of the preoperational test

program, including the witnessing of preoperational tests as conducted during

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the Hot Functional Test, the review of activities in the QA interface and the I

review of unresolved items identified by the inspector in previous inspections.

l The 3 day latch release test of Main Steam Isolation Valve A operator was j witnessed at the temporary site test facilit Results: No violations were identifie Note: For acronyms not defined refer to NUREG 0544, " Handbook of Acronyms and Initialisms".

8612290436 861218 PDR O ADOCK 05000412 PDR l

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DETAILS Persons Contacted Duquesne Light Company (DLC)

R. Bernier, Engineer, Phase-2 (S&W)

R. C. Callaway, Test supervisor, Phase-2 T. Capkovic, Test Engineer (S&W)

M. Covey, Test Engineer

• A. W. Crevasse, Deputy QA Manager P. Czarnowski, Engineer (S&W)

• N. J. Daugherty, Director Systems Testing D. Esielionis, Assistant Superintendent, Site Engineer (S&W)

W. Fragapan, Engineer, Westinghouse, GTSD E. Freeman, Test Engineer, Phase 1 and 2

• J. P. Godleski, Senior Test Engineer N. Harlow, Test Director, NSSS J. P. Higgings, Test Engineer, Pase 2

• D. C. Hunkele, Director Operations QA

• J. D. Johns, Supervisor QA Surveillance F. Michael, Test Coordinator, Phase-2

• T. P. Noonan, Superintendent Operations and Maintenance E. Peace, Nuclear Operations Engineering Supervisor L. M. Rabenau, Compliance Engineer P. Sowatskey, Engineer, Westinghouse, GTSD J. Steinke, Test Engineer, Phase-1

• R. J. Swiderski, Startup Manager R. Thale, Nuclear Operations Engineering

• L. P. Williams, Director Startup, Phase-1

• R. G. Williams, Software Development Supervisor U.S. Nuclear Regulatory Commission

• A. A. Asars, Resident Inspector J. Beall, Senior Resident Inspector Licensee's Action on Previous Inspection Findings (0 pen) Unresolved Item 412/86-38-01, " Emergency Diesel Generator Concerns" Scope A number of concerns were identified during the review of Phase-1 test reports for the Emergency Diesel generators (A and B), as follows:

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. The lack of retest and SWR information in the final test repor The potential for bypassing of the Nugent fuel oil filters through support plate warpage or casting sand in the filter head outlet casting as experienced at another nuclear plan Lube oil contamination found on lube oil strainers may not be system preservative but soot from normal diesel engine operatio Fuel oil consumption data needed to be checked against required tank volume, as provided in the FSAR, for emergency operatio Incorporation of the engine balancing procedure and criteria in procedure and criteria in procedure 2T-NNS-368-2.24 for " Initial Loading of EDG, 2EGS*EG2-2."

In this inspection, the progress made on this item by the licensee was determine Discussion The licensee had assigned items b and c to the Operations Maintenance Department, Mechanical Sectio The other items will remain with Phase-1 testing. Items b and c were discussed with said department. For item b, it could not be determined if parts were stocked for the Nugent oil filters nor could it be deternined what the filter support plate or the filter housing heads looked like because a detail drawing of the filter parts was not availabl The maintenance supervisor will provide this information in a future inspection. For item c, there was no chemical analysis available for the deposits found on the lube oil strainers of the EDG The maintenance supervisor will inquire about the possibility of having an analysis done for any new deposits should they develop in the course of further testin With regard to item a above, work was in progress to collect the SWRs and available retest data for incorporation in the final test package Item d, fuel consumption data, will be accepted as is. A more accurate determination of the fuel consumption and the criteria (based on FSAR requirements) data will be done during the Phase-2 test Item e has not been addressed ye Conclusion Progress is being made on the 5 items listed above, but answers for all items are not yet availabl . . _ _ _ _ ._ . _ . .__ ._ ___ _

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' Preoperational Test Witnessina

. Scope-Selected steps of the Phase-1 and Phase-2 test procedures listed in

", " Attachment A were witnessed. Test witnessing included observations of the attributes listed in Section 3.2 of Inspection Report 50-412/86-2 ,

Discussion o

For the thermal expansion test performed under procedures PO-2.06.12, the inspector selected five different piping points to followup. All of these

, points had measurements outside the calculated margin for the 250 F temperature plateau of the HFT. Two of these points were inspected in

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situ.. (A total of 150 channels are wired into the computer of the Westinghouse Structural Analysis Mobile Unit (SAMU) trailer.) The two

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points inspected were channels 070, 071, and 072 (3 directions) for 2 FWS-PSSP-006 and 058, 059, and 060 for a main steam line. Readings for

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058 and 060 were outside the allowable margin and so was the reading for

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07 The point on the main steam piping is in the center of a 90*

l' horizontal elbow. This elbow is restrained vertically on top and botto It can grow freely in the horizontal plane. The growth in the horizontal

. plane at the 250* F plateau of the HFT is not sufficient to overcome the

! friction of the pipe on the vertical suppor This point was still j outside the allowable margin at the 350 F platea It is expected that j the friction force will be overcome before reaching the 450* F leve .

?- Many of the surveillance points are outside the calculated allowable

margin because of the lack of an actual temperature measurement. Expected j pipe measurement is calculated on the basis of the temperature plateau of

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the HFT. The actual local temperature may be much lower. Many of the

! points outside of the allowable margin fall into this category. Each of these points, however, is checked out by the responsible test engineer at

! each HFT temperature platea The inspector will continue to monitor the 5 selected piping points at the i different HFT temperature plateaus.

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The vibration checks made on the Reactor Coolant Pumps (P0-2.06.02) via

, the instruments on the main control board indicated that most frame i vibration points were outside of the allowable vibration limit of 3 mils t (4 mils and up).

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The shaft vibration on all 3 pumps was close to or at the 15 mils

. allowable value. Westinghouse was consulted on this matte They are sill balancing (adding washers to the coupling) the 3 pumps.

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The inspector witnessed the first roll (uncoupled) on the steam turbine l driver of the Auxiliary Feedwater Pump. The noise level and vibration I were very lo The test, however, had to be discontinued because it i interfered with the Steam Generator Blowdown Test.

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During this inspection, the inspectors observed various phases of the alignment, setup and check out of the atmospheric steam dump valves. The system has three atmospheric steam dump valves (2SVS*PCV101A, B and C),

one valve for each main steam generator. One additional valve (2SVS*

HCV104) is provided for residual heat removal (RHR) and removes heat from all three steam generators through a check valve arrangement which prevents a three steam generator blowdown if one steam generator has a steam line break upstream of its main steam isolation valve. All valves are Copes-Vulcan 10 inch reverse seating gicbe valves with Borg-Warner hydraulic actuators. Valve trim is 300 series stainless stee Initial checkout of the RHR valve had been satisfactorily completed on November 9, 1986 using procedure ITP-126, Borg-Warner E-H Valve Tes Portions of the checkout of the 'A' atmospheric dump valve was being performed on November 12, 1986 using the same procedure when a valve hydraulic cylinder stroke length of 3.50 inches was observed. Acceptance criterion was 3.25 1 0.12 inches. Test Deficiency Report 3316 was issued for problem resolutio It was determined by inspection of the 'A' valve that the valve stem which is threaded into the valve disk, internal to the valve, had been broken by the hydraulic force applied in the close direc-tion. As a result of the ' A' valve failure, all valve hydraulic actuator system pressures were set to function at a lower value to prevent further stem failures. New hydraulic pressure setpoints were provided by Stone

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and Webster Engineering and will be used during HFT. Subsequent to the HFT, new 400 series stainless steel valve trim will be installed and hydraulic pressures will be set to the original value and the valves retested. Various segments of the resetting of the hydraulic system pressures were witnessed by the inspector In addition, portions of the preoperational testing of the above valves was also observed. During the initial retesting of the ' A' and RHR valve (after hydraulic pressure was lowered) the test engineer noted that neither valve would fail to its closed position on a loss of 480 volt AC

power to the valv This resulted in initiation of two startup work request (7358 and 7359). It was later determined that a lead had inten-

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tionally been lifted during Phase-1 testing to override the fail closed function. The lifted leads were appropriately entered into the lifted lead and jumper lo This oversight occurred because the test engineer for Phase-2 testing (preoperational) was not assigned to this system during Phase-1 testing and was not aware of the lifted lead and apparently failed to check prerequisites (lifted lead and jumper log) prior to the performance of that portion of the test. This item was discussed with the licensee management and is a further example of violation 412/86-28-02 which was issued to the licensee on November 17, 1986. The licensee's plans for corrective action and when those actions will be complete will be issued within 25 days of the November 17, 1986 dat . __ _ _ _ . - _ _ , _ ___ .__ . . _ _

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Findings Apart from the test control problem observed during the Phase-1 testing of the Atmospheric Steam Dump Valves and the RHR Valve, no unacceptable conditions were noted during the test witnessing of the tests listed in Attachment . Independent Inspection Effort Scope In a continued effort to monitor the activities on the MSIV modifications planned by Crosby, the vendor, and the licensee to correct several generic problems and the impact they have on the startup program, the inspector obtained the latest information on developments in this regar The inspector also witnessed the 3 day latch release test on the valve A operator in the temporary test facility at the sit Reference Report Number G&W-FSD-2538, " Topical Report Nuclear Main Steam Isolation Valve Systems", dated January,1979, issued by Gulf and Western Manufac-turing Company, Fluid bystems Division, Energy Products Group. To more fully understand the development work and analysis done for these valves prior to their use at BV-2, a copy of the reference was obtaine Discussion The valve operators of valves A and B have been mounted on their pads in the temporary test facility and latch release tests are in progress. The valve operator for valve C was shipped to Crosby for the performance of latch release tests at their facility. Having one operator at the Crosby facility will facilitate the testing of the new air operator for the latch release mechanism which is currently being designe The test methodology for the latch release tests at the site calls for a 3 day, 7 day, and 30 day release test of both valve operators (A and B).

The pivot doors for the latch release have been reinforced to prevent appreciable deflection under the push force of the (Solenoids) plunger For the latch release test, the push force is supplied manually via a load cell (BLH Electronics, type T3P1) with a capacity of 500 lbs. The signal of the load cell is supplied to a strip chart recorde An accurate record of the push force transient can be obtained in this manne . , _ _ - _ - -. - - - -

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In the redesign, the solenoid towers will be removed. The newly designed air operator will take its place after completion of the push test for latch releas The inspector witnessed the 3 day release test for valve A (with the new reinforced door). The push force was recorded as 155 lbs. The instantan-eous retest showed a push force of 99 lbs. The data appears to becoming more consisten There are no plans to do any modifications to the balls of the ball valves of Beaver Valley The vendor feels there are no further problems anticipated for the balls. The cracking problem of the latch bearings also appears to be resolved via the adoption of the new roller housing material. This would leave the problem of the latch release for which the tests are being performe The impact of the MSIV modification program on the startup test schedule is not fully known yet but its completion is now estimated to be after May 1, 1987, the current scheduled date for core loadin Conclusion The licensee has not abandoned the application and plans to complete the valve modifications as planned by Crosb . QA/QC Interface Scope Responses by DLC startup management to Surveillance Deficiency Reports (SDRs), reviewed previously, were examined for correctness and adherence to the requirements of the Startup Manual. Additional SDRs for recently performed POT surveillances, as listed in Attachment B, were reviewe References BVPS:RJS:12241:3695, DLC liemorandum from R. J. Swiderski to C. Ewing, " Response to Quality Assurance Surveillance No. S0V-32-86",

dated September 19, 1986.

' BVPS:RJS:12241:3608, OLC Memorendum from R. J. Swiderski to C. E.

l Ewing, " Response to Surveillance No. OPS-03-86", dated June 19, 198 Discussion Responses to S0V-32-86 and OPS-03-86 were reviewed (see reference 1 and 2). In the case of OPS-03-86, additional input from the safety meeting, referred to in reference 2, is needed to close out the item. The response to 50V-32-86 appears adequate.

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The SDRs listed in Attachment B were reviewed for their importance to safety. In POT-24-A-86, jumpers were installed for the performance of the test which was suspended prior to its execution. The control of the jumper, however, was not transferred to the Operations Jumpers and Lifted Lead Log. In SOV-36-36, a 6.25 inch Test Auxiliary Switch Actuator (ASA)

was installed in lieu of a 3 inch ASA. No entry was made in the Jumper and Lifted Lead Log as required in the prerequisites of the procedur The other surveillances deal with the incorrect entry of changes in the test records. The inspector will continue to monitor the responses of startup management to the SDRs listed in this inspection report and earlier report Findings No unacceptable responses were identifie . Plaht Tours and Meetings The inspector made several tours of various areas of the facility to observe work in progress, housekeeping, cleanliness controls and status of construction and in particular testing activities. Testing activities observed by the inspectors are discussed in Section 3 of this repor The inspectors also randomly attended the licensee's morning Hot Functional Test (HFT) Plan of the Day meeting during which the current status of HFT activities and any problems were discusse No unacceptable conditions were identifie . Exit Interview At the conclusion of the site inspection, on November 21, 1986, an exit interview was conducted with the licensee's senior site representatives (denoted in Section 1). The findings were identified and previous inspection items were discusse At no time during this inspection was written material provided to the licensee by the inspecto 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 inform-ation subject to 10 CFR 2.790 restrictions.

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ATTACHMENT A I 412/86-40, Eeaver Valley-2 Test Witnessing

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Proc. N Title Rev. N Approval Date P0-2.06.08 Integrated Hot Functional Test 0 10-10-86 PO-2.21A.01 Atmospheric Steam Dump Valves 0 04-17-85 and RHR Release Valve Test P0-2.03.01 In Core Thermocouple and RTD 0 02-27-86 Cross Calibration P0-2.06.07 Pressurizer Pressure and Level 0 05-28-86 Control Test P0-2.06.02 Initial RCP Performance Test 0 07-09-85 S0V-2440.05 Containment Air Recirc. Fans 0 08-28-86 P0-2.06.12 System Vibration and Thermal 0 11-14-86 Expansion Testing Ouring HFT w

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ATTACHMENT B 50-412/86-40 Review of QA Surveillance Deficiency Reports (SDRs)

(SDR No.) Test Date Sury. N Descrip'.lon SDR Activity Proc. N Reviewed POT-18J-86 Errors in antering changes P0-2.15A-01 11-07-86 (SDR-1) to proced.fre P0T-23A-86 TCN #8 not incorporated in PO-2.06.06 11-10-86 (SDR-1) precedure P0T-24A-86 Jumper not transformed to PO-2.24B.01 10-23-86 (SDR-1) Jumper and Lifted Lead Log POT-28B-86 Changes not entered in test P0-2.10-01 11-12-86 procedure 50V-36-86 Different ASA size used but S0V-2.29.01 10-08-86 not entered in logs

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