IR 05000412/1986006

Insp Rept 50-412/86-06 on 860301-31.No Violation or Significant Safety Concerns Noted.Major Areas Inspected: Licensee Actions on Outstanding SER Issues,Preoperational Program Implementation & Three Test Procedures

ML20203F403
Person / Time
Site:	Beaver Valley
Issue date:	04/14/1986
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20203F400	List: IR 05000412/1986006 ML20203F396
References
50-412-86-06, 50-412-86-6, IEB-85-003, IEB-85-3, NUDOCS 8604250143
Download: ML20203F403 (10)

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U.S. NUCLEAR REGULATORY COMISSION

REGION I

Report'N /86-06 Docket N License N CPPR-105 Licensee: Duquesne Light Company Nuclear Construction Division P. O. Box 328 Shippingport, PA 15077 Facility Name: Beaver Valley Power Station, Unit 2 Dates: March 1 - 31, 1986 Inspectors: W. M. Troskoski, Senior Resident Inspector L. J. Pr*vidy, Resident Inspector Approved by: [. 49k) / 8d Lt. E. Trfipp, Chief, Reactor Projects Section 3A Date Inspection Summary: ' Inspection No. 50-412/86-06 on March 1 - 31 -198 Areas Inspected: Routine inspections by.the resident inspectors (124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br />) of licensee actions on previous inspection findings, preoperational program implemen-tation, review of the containment integrated leak rate test procedure, and remain-ing recirculation and quench spray system preoperational test procedure Results: No violations or significant safety concerns were identifie

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' DETAILS ( _ Persons Contacted R. Coupland, Director, Site Quality Control

. E. Ewing, Manager, Quality Assurance T. P. Noonan, Station Superintendent R. J.' Swiderski, Startup Manager D. Williams, Chairman, Joint Test Group The inspector also met with other licensee and contractor personnel during the course of'the inspectio . Project Status Summary By Order dated March 14, 1986, NRR extended the construction completion date-for Beaver Valley, Unit 2, (CPPR-105) to December 31, 1986. Based on the projected completion dates listed below, another extension request is expected within the next month or tw Construction activities are currently estimated to be 93.6% complete, with 292 of 450 subsystems turned over for flushing and proof-testing. For sof t-ware, 50 out of 123 preoperational (PO) and initial startup tests (IST) have

'been issue The remainder are in various phases of developmen Approximate dates for the major project milestones, as currently estimated by the licensee are as follows:

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Reactor Coolant System Cold Hydrostatic Test April 10, 1986

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Fuel Receipt September 1, 1986

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Integrated Hot Functional Test October 20, 1986

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Loss of Power Test February 2, 1987

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Integrated Leak Rate Test February 23, 1987

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Fuel Load May 1, 1987

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Initial Startup May 16, 1987

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Commercial Operation August 30, 1987 Major activities planned for next month include the RCS cold hydrostatic test, and accelerated ESF subsystem turnovers for proof and preoperational test . Inspection Program Status Summary Preoperational-Test Program Inspection completion status is approximately as

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L AREA  % INSPECTION COMPLETE L Overall Program 20 l.

i Procedure Reviews:

Mandatory 35 Primal 10 l

Test Witness:

Mandatory 0 Primal 5

Results Review

Mandatory 0 Primal 5 This inspection status is consistent with applicant test program progres Accelerated subsystem turnover is expected to occur after RCS hydro. At the end of this inspection period, there were approximately 91 open NRC inspection items including 8 bulletins, 7 violations, and 15 construction deficiency re-

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ports; the remainder are inspector follow or unresolved item . Licensee Actions on Previous Inspection Findings (0 pen) IFI (86-01-01): Review effect of SER Section 14 Open Items on test progra By letter dated March 20, 1986, DLC provided the License Project Manager, NRR, with a listing, depicting their understanding of the status of outstanding SER issues. Discussions with the LPM indicated that a supplement to the SER was expected to be issued soo This item remains open pending-review of this supplement with regard to the test program open item (Closed) IFI (86-04-04): Replace emergency diesel generator screen bolt The inspector toured the EDG rooms on March 27, 1986, and verified that new bolts were installed in retapped screen holes that were previously strippe This item is close (0 pen) Unresolved Item (83-02-04): This item is related to shimming require-ments for electrical support surface mounted plate installations. One of the two remaining concerns involves a specification change for electrical in-stallations (28VS-931) that would permit gaps of up to 1/8" between concrete walls (or embedment plates) and support baseplates 1/2" thick or large During Inspection No. 412/85-24, the licensee committed to performing an engineering evaluation for the generic qualification of gaps up to 1/8" on baseplates 1/2" thick and larger. An enveloping analyses was to be performed for typical baseplate configurations with various locations of gap areas'under the plat SWEC provided the study to the NRC for revie The study (Calculation N NS(8)-220)' involved a finite element analysis of a typical 12" x 12" x 1/2" plate with 4, 3/4 inch diameter Hilti bolt The support loads were

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applied via a 2" x 2" tube structural attachment at the center of the plat Four load cases were investigated in the study; axial tension (+8800 lbs),

axial compression (-8800 lbs), uniaxial bending (30,000 in-lbs) and biaxial bending (15,000 in-lbs each). The loads were chosen to give resulting anchor bolt loads approximately the maximum allowable. The analysis was performed for nine cases depicting various distribution of gaps under 1/2" thick plates and two cases for 3/4" thick plates. The area of gaps under the plate was varied from 0% to an extreme of 90% of the baseplate area. The results of the study concluded the following:

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For 1/2 inch thick surface mounted plates with at least three points of contact (not on the same side of the attachment) between the plate and the supporting surface, the increase in anchor bolt loads was approxi-mately 27%. The increase was determined by comparing bolt loads result-ing from the worst case evaluated to that based on a plate mounted with no gap lates with a gap configuration The study similar waswhich to that extended to 3/4" thick p' worst case" for the 1/2" thick was considered plate evaluated above. The increase in anchor bolt loads was found to be 41%.

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The aliowable anchor bolt loads used in the design specification are at least 52% higher than the allowable loads specified by the manufacturer (Hilti). The manufacturer's allowable loads are based on a minimum safety factor of The study performed by SWEC concluded that the built-in allowable anchor bolt load margin is higher than the largest anticipated increase that would result from the presence of 1/8" gaps in a worst case scenari The study did not evaluate how the plate stresses were affected by the various configurations. Such an evaluation should be addressed to ensure that any resulting increase in plate stresses (for the worst condition) would not ex-ceed the allowable code limits. This item remains open pending licensee evaluation of plate stresses, and further Quality Control inspection to ensure that the as-installed shims provide the necessary three point contac (Closed) 50.55(e) (78-00-02): Inadequate structural design in lower pump cubicle. On September 22, 1978, the licensee reported that the structural design for the lower recirculation pump cubicles was inadequate for all ex-pected loading conditions. A calculation review found that the Reactor Con-tainment Cofferdam had been incorrectly assumed to be a qualified QA Category I structure. This resJ1ted in the failure to include the contribution of lateral soil pressure acting against the pump cubicle outside wal The cubicles were subsequently redesigned to meet the PSAR commitment On March 21, 1986, the inspector observed installation of support beams in three of the four cubicles. At the conclusion of this inspection period, all work had been completed except for grouting between the beams and concret No unacceptable conditions were identified and this item is close p

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(0 pen) IFI (86-02-04): Review control of interface between temporary and l

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safety-related equipment during system flushes. During SIS pipe flushing, the inspector identified a concern whereby temporary hoses and test equipment did not. appear to be positively controlled to ensure that installed safety components would not be damaged during these activities.

. During this inspection period the pressurizer spray line flush was reviewed.

l The inspector noted that the pressurizer manway cover was removed and a tem-l porary plexi glass cover was installed. An opening was provided thru the l cover to permit the passage of a 4" fire hose which was attached to the 6"

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spray line inside the pressurizer via a temporary piping adapter. This piping adapter was an assembly consisting of a PVC coupling and SST nipple and elbow that was installed after the spray head was removed.

L The inspector reviewed the Disassembly / Reassembly Record initiated to follow l

the work on this equipment. This is required in accordance with FCP 302 "Re-moval or Disassembly / Reassembly of Permanent Plant Equipment." The record was properly executed by Construction and Site Quality Control (SQC). Also, the inspector witnessed a portion of the reassembly operations (i.e., removing the temporary piping adapter and installing the spray head) and noted the following:

, The disassembly and reassembly work was performed in accordance with the l detailed instructions provided in the Pressurizer Instruction Manual -

2504.150-001-00 . Protective plastic wrapping was installed on top of the heater banks for protection.

' SQC checked the inventory of items installed and removed from the pres-surizer to provide accountability.

, Construction checked with Westinghouse during the progress of the work.

l Upon restoring the spray head to normal, construction swiped the spray line to check for chloride contamination before reinstallatio Surfaces l were found to be clear of chloride . Prior to removing the temporary plexi glass cover, proper cleanliness l controls were established to upgrade the immediate work area to a Zone II Are The inspector found these operations satisfactory in that they demonstrated good interface betwean construction work and testing operations. This item l remains cpen pending further sampling of other areas af ter the RCS hydro.

(0 pen) Unresolved Item (84-14-03): Engineering Confirmation Program. The

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inspector met with various DLC and Stone and Webster Engineering Corporation f

(SWEC) personnel on March 11 and 13, 1986, to review the status of the Engi-neering Confirmation Program (ECP). Discussed were the overall program ac- ,

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r complishments at this point and the ongoing DLC efforts identified in the original ECP presented to NRC Region I on October 21, 1983. Additionally, the interface between DLC Nuclear Censtruction Division engineering personnel

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and the Startup Group was discussed. The inspector considered such a meeting desirable since a comprehensive review of the program has not been undertaken

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since mid 1984 although several individual areas have been reviewe At the March 11, 1986, meeting DLC summarized their efforts to date concerning the Design Basis Endorsement (DBE) portion of the ECP. DLC concluded that l the DBE portion of the program has given them an opportunity to accelerate

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their understanding of all plant design basis criteria. A typical example of this is the DLC review of 2BVM-42, " Cable Philosophy", where the content

! was validated and applicant engineers became cognizant of such things as built-in safety margins in the selection of cable sizes, the impact of system studies that will change with plant load growth, and an awareness of all de-sign inputs. Such information will play a key role in future cable modifica-tions and validation of existing cables as plant load grows. In parallel with the validation of key design basis documents, DLC also reviewed certain elec-

. trical calculations. A future benefit expected to be realized is the identi-l fication of selected candidates for future reviews as equipment or load

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changes are 11ade.

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During the March 11, 1986, meeting the inspector questioned the structure of the interface between the Engineering personnel of the DLC Nuclear Construc-tion Division (NCD) and the Startup Group. The inspector noted that this interface was mentioned as part of the " Engineering Scope and Participation,"

which was another facet of the ECP. The inspector questioned the extent of NCD engineering personnel involvement with SWEC design and DLC Startup Group

, personnel in developing a change in the main steam piping to the auxiliary

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feed pump turbine to incorporate drip pots for the purpose of condensate re-moval and proper piping warmup from cold conditions. This particular item was discussed in Inspection Report 412/85-26, and was addressed in detail at a followup meeting with DLC and SWEC personnel on March 13, 1986. DLC advised the inspector that the steam piping redesign was coordinated between SWEC de-sign personnel in Boston and was approved by NCD engineering personnel after calculations had been performed to support size and locations of the drip pot The' inspector requested copies of the calculations and the revised construc-tion drawings for review. These items have not yet been reviewe Concerning the test performance aspect of this design change, the inspector was advised that NCD engineering personnel have minimal contact in the pre-paration of test procedures. Their job is to produce final design and con-struct documents, such as equipment specifications and drawings. The respons-ible SWEC and DLC Startup Group personnel preparing the test procedures must take the current approved design and construct documents and devise proper o

test procedures based on these documents.

l The inspector found the program to be working as intended, and will continue to monitor and report on this item in future inspections.

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(0 pen) CDR (86-00-05): 50.55(e) Fire in Emergency Diesel Generator (EDG)

exhaust system. While completing a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> proof test run on the number 1 EDG, roofing tar caught fire and was extinguished by the BV-2 fire brigad The cause of the fire was a design deficiency in the EDG exhaust system: the exhaust pipe penetrates the roof and bends 180 degrees, subsequently discharg-ing directly down on the roof. Though there was fire brick directly under the pipe, it did not cover a sufficient area to minimize the heat buildu The inspector was informed by the Startup Manager that further EDG testing would be delayed until potential structural damage could be evaluated and necessary design changes implemente There was no damage to the ED (0 pen) Unresolved Item (85-20-01): Review engineering justification for non-staggered Cadweld splices during closure of the temporary containment con-struction entranc See Detail 5 of this report for further discussion . Containment Integrated Leak Rate Test (CILRT) Procedure Review The inspector performed a preliminary review of P0-2.47.07, Containment Type A Leak Rate Test (Containment Integrated Leak Rate Test CILRT) scheduled for Spring, 1987, to determine if there was adequate adherence to 10 CFR 50 Ap-pendix J, FSAR, the applicable requirements, guides and standards to which the licensee had committe The CILRT procedure includes the Containment Hydrostatic Test (CHT), which is the construction acceptance test for the containment building, that needs to be performed before the LILRT. The containment building will be pressur-ized to 1.15 times design pressure (52 psig) during the CHT. Four pressure plateaus will be observed so that radial and tangential deflection measure-

! ments and crack mapping can be performe Once reaching 52 psig, the con-tainment building will be depressurized in the same manner. The inspector questioned why crack mapping plans did not include a portion of the temporary construction opening on the north side of the containment building. This opening was closed during Fall 1985, and was discussed in NRC Inspection Re-port 412/85-20. Specifically, concerns were raised that exceptions to the exterior wall Cadweld splice staggering method used at the construction open-ing appeared to present a structural weakness that requires engineering justi-fication. This is Unresolved Item (85-20-01), and is still open pending lic-ensee response to the concerns. The necessity for crack mapping the temporary construction opening will be determined after review of that engineering justificatio Upon completion of CHT, the licensee will begin CILRT after allowing a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period before CILRT with the containment below 85% Pa (38 psig) per ANSI /ANS l 56.8-198 After the allotted time period, the containment buildirg will be pressurized to 45 psig (predicted pressure following a design basis accident).

l The leak rate will be monitored for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at which time the lic-

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ensee will perform a superimposed leak verification test.

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The inspector found the CILRT procedure to adequately provide for surveying containment temperature and humidity, installing and calibrating instrumenta-

[ tion, aligning containment systems, responding to excessive leakage during l

Lesting, sequencing CHT and CILRT, logging and plotting test data, controlling pressurization and depressurization, and analyzing test result The inspector had no further concern . Preoperational Test Review - Recirculation and Quench Spray System The inspector reviewed the following Containment Depressurization System pre-operational test procedures to verify technical adequacy and to ensure con-sistency with Regulatory requirements, guidance and licensee commitments:

P0-2.13.03, Quench Spray and Recirculation Spray System Nozzle Air Flow Tes P0-2.13.04, RWST Tes P0-2.13.05, Quench Spray Chemical Injection System Tes These tests are the last of five procedures developed for this system. The first two were previously reviewed in Detail 5 of Inspection Report 412/86-04.

! References FSAR Section 14.2.12.15.3 thru 5, Test Abstracts FSAR Section 6.2, Containment Systems SER (NUREG-1057) Section 6.2, Containment Systems l

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The only open licensing issue involving the Containment Heat Removal System concerns further applicant technical justification for the acceptability of

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an assumed 50% sump blockage during an accident. Net positive suction head calculations were previously submitted to NRR and found acceptabl Heat Transfer Calculations In reviewing the heat transfer calculations in P0-2.13.04, the inspector noted that acceptance limits were assigned to the heat transfer coef ficient (U) and not to the overall system performance capability (Q). The JTG Chairman stated that the value of Q was defined in applicable design documents and that the minimum value of U was determined from them. Since these calculations are separate from the preoperational tests, a sampling review will be performed to verify that minimum heat transfer capability specified in the P0s is cor-l rect. This area will be tracked by Inspector Follow Item (86-06-01).

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MOV Testing Each test requires successful operation of specific motor operated valves (MOVs) to complete its safety function. At Beaver Valley Unit 2 these valves l

have a limitorque operator to control motor operation. IE Bullet}n 85-03, MOV l

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Common Mode Failures During Plant Transients Due to Improper Switch Settings, addresses the control of these setpoints to ensure operation at expected dif-ferential pressures. Through discussions with DLC Maintenance engineers, the inspector was informed that the station was purchasing two MOVATS (Motor Operated Valve Analysis and Testing System) units for use in MOV signature analysis to compliment the inservice inspection program. Review of the scope

! of this program for safety related valves will be reviewed in conjunction with IEB 85-0 No deficiencies were identified by the inspecto . Procedure Development

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Test Program Objectives To verify that each test objective listed in FSAR Section 14.2.12, Initial Test Descriptions, Amendment 11, is scheduled to be included in the preopera-tional test program, the inspector cross-referenced each to either a System Operation Verification (50V), Preoperational Test (P0), or Initial Startup

! Test (IST) listed in the Master Test Index. Several apparent discrepancies i were identified and discussed with cognizant Startup Group (50G) personnel.

l These concerned several P0s and ISTs for which no test abstract had been in-

! cluded in the FSAR. The inspector determined that each was in a stage of i development by the responsible test engineer, prior to submittal to the Regu-l latory Affairs Department for inclusion in a future FSAR amendment. The only i exception is IST 2.01A.10, Startup Testing Program, which serves as a schedule log for other ISTs following core load. ForeachFSARtestobjectivecur-rently listed, an appropriate test procedure was in some stage of developmen The inspector had no further questions in this area.

l l Test Procedure Development SUM Chapter 3.5.1 defines the administrative control for preparation, review, approval and revision of test procedures. From a review of the P0s referenced in Details 5 and 6 of this report, the inspector verified overall adherence to the requirements. Af ter review and approval by the JTG and Station Super-intendent, the test procedures are not released for performance until author-ized. This ensures that the cognizant engineer reviews the procedure against current system design documents and that appropriate revisions are made prior to use. During review of the P0-2.13 series tests (Detail 6 of this report),

the inspector noted that the Operating Manual (OH) Power Supply and Control Switch List was referenced for performing system lineups. A review of the OM indicated that it was not up to date. Discussions with several test engi- i

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neers indicated that they were aware of this and that it was their responsi- l bility to verify the accuracy of OM information referenced in a test procedure, !

, as part of the design document review. The inspector was satisfied that these

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administrative controls were understood and implemented.

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l -10 8. Consultation with Workers.

! During the course of this inspection period, discussicas with licensee and

. contractor personnel indicated that a program had been instituted to ensure

" that NRC identified concerns w re quickly elevated to the appropriate DLC management leve This program would require all employees to document any item discussed with an NRC inspa: tor and forward it to his department super-i visor. Such a requirement appeats to be counter to the intent of 10 CFR l 19.15(b), Consultation With Workers During Inspections. This was discussed l

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with the Station Superintendent.

! At the exit meeting on April 1.-1986, the inspector was informed that a letter l- would be issued under the Startup Manager's signature, clarifying the station's policy. SVG personnel are to' notify their immediate management of any NRC safety concern but are not required to identify all items discussed with the NRC. The inspector found this approach to be acceptabl . Exit Interview.

L Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of the findings was further discussed with the licensee at the con-clusion of the report period.

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