IR 05000412/1987035

Insp Rept 50-412/87-35 on 870413-0527.No Unresolved Items or Violations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Site Activities,Preoperational Test Program Implementation & TMI Action Plan Requirements

ML20216C528
Person / Time
Site:	Beaver Valley
Issue date:	06/17/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20216C457	List: IR 05000412/1987035 ML20216C451
References
TASK-2.B.1, TASK-2.B.2, TASK-2.B.4, TASK-2.E.4.2, TASK-TM 50-412-87-35, IEB-80-22, IEB-85-003, IEB-85-3, NUDOCS 8706300302
Download: ML20216C528 (16)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-35 Docket N i 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: April 13 - May 27, 1987 l l Inspectors: J. E. Beall, Senior Resident Inspector i L. J. Prividy, Resident Inspector-S. M. Pindale, Resident Inspector, Unit 1 A Fink , Lead Reactor Engineer, DRS

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Approved by: 8-E. E. Trfpp, Chief, Reactor Projects Section 3A h7

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Inspection Summary: Inspection No. 50-412/87-35 on April 13 - May 27, 198 Areas Inspected: Routine inspections by the resident inspectors (360 hours) of licensee actions on previous findings, site activities, preoperational test program , implementation, TMI Action Plan Requirements, control room wall removal and diesel 1 generator instrument panel vibratio ! Results: buring the inspection period, the licensee completed various key construc- I tion and preoperational test activities in preparation for receiving a low power j license. The Site Security Plan was implemented (detail 9), a common control room l Was established (detail 8), several meetings were held to assess the licensee's-readiness for an operating license (detail 13), and the loss of offsite power test was conducte No unresolved items or violations were identifie \ l , 8706300302 870619 ADOCK0500g0

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DETAILS  ; i Persons Contacted During the report period, interviews and discussions were conducted with mem-bers of the licensee's management and staff as'necessary to support inspection activitie ; l Project Status Summary

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At the close of the inspection period, construction activities'were estimated to be in excess of 99% complete and the' licensee was. issued a low power license on May 28, 1987,.the day after this inspection period ended. The seven day delay from the licensee's earlier estimate was due to the need to repair certain seals to meet environmental qualification requirement Approximate dates for power ascension testing, as estimated by the licensee !

.at the end of this inspection period were as follows:
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Begin fuel load (Mode 6) May 29, 1987  !

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Enter Mode 5- June 6, 1987

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Enter Mode 4 June 19, 1987

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Enter Mode 3 . June 20, 1987

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Initial criticality (Mode 2) July 1, 1987

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Full Power License (>5%) July 5, 1987

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100% Power August 9, 1987 I

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Plant Trip from 100% Power . August 16, 1987 '

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Achieve Commercial Operation August 25, 1987 At the close of the inspection period, the licensee was experiencing some difficulties with.the four' safety-related recirculation spray heat exchanger Internal debris and tube wear were identified and the severity of damage and the extent of necessary corrective action were not yet determined. The heat exchangers are of vertical, counter flow, shell and tube design with a tube length of 37 feet and a shell ID of about 33 inches. Replacements are not readily available such that severe damage to these components could impact project schedul . Inspection Program Status Summary Preoperational Test Program Inspection completion status was approximately as follows:

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   % INSPECTION COMPLETE  l AREA  END OF THIS PERIOD END OF LAST PERIOD j
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Overall Program 95 80

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Procedure Reviews: Mandatory 100 100 Primal 100 100 ! i

' Test Witness:

Mandatory 100 70 Primal 100 100 Results Review

Mandatory 75 50 1 Primal 75 50 This inspection' status is consistent with the applicant's test program pro-- gress. At the end of this inspection period, there were approximately 41 r open NRC inspection items as listed below: i N0. OF OPEN INSPECTION ITEMS TYPE OF ITEM END OF THIS PERIOD 'END OF LAST PERIOD Bulletins 1 3 Violations 4 5 1 Deviations 0 0

Construction Deficiency j Reports 10 15 j

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Unresolved 26 27 TOTAL 41 50 4 .' Licensee Actions on Previous Inspection Findings i J

(Closed) Construction Deficiency Report (87-00-05): Crimping of tubing con-nected to yent ports on Category I control valve This item resulted from the improper interpretation of a Field Construction Procedure (FCP). Short lengths of soft copper tubing were connected to the vent port of solenoid operated valves (S0V) which control air to various Category I control valves. This copper tubing was installed as a temporary measure during construction to maintain internal cleanliness. riowever, this installation was not authorized nor was it identified as being temporary and a concern existed that the copper tubing could restrict the air flow through I

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'the' S0V.and thereby cause malfunction of the associated Category I control valvo. The licensee took corrective actions to remove this~ temporary. copper tubing where it was not needed and to revise their FCPs to properly control its temporary use. Based on a review of the FCPs and a review of several !
. Category I control valves in the plant, the inspector. determined that these corrective actions'were acceptable. This item is close !

-(Closed) Construction Deficiency Report (86-00-03): -Improper material ~ supplied for-snubber brackets. :This' item was initially reported as a potentially sig--

nificant deficiency. However, a subsequent licensee evaluation concluded that thisLitem was not' reportable and'should be withdrawn since the use'of the q incorrect bracket material would not have adversely affected the ability of- 1 the. brackets to handle the design loads. Therefore, this item is now close j (Closed) V.iolation.(86-47-02): Cable reels in outside storage, Location P,. had thei.r cable ends cut and exposed to the environment with-no. protective end caps ~or the tape cover over the cable ends was frayed and in such condi-tion that the cable" ends were not protected from the environmen .The.l icensee has performed a walkdown inspection of the cable' reels in_ storage and staging areas to. check for cable end seal-integrity. . Training classes have been given to the electrical craftsmen'and supervisors.on site and Field- i Change-Procedures (FCP 430 & 431)'were revised and issued on January 15, 1987, which addressed the. cable sealing subject. Site Quality Control performed an independent verification of the cable end caps on cable. reels in the stor-age yards and identified their findings on N&D 38863A. The results of the N&D have been-revie'wed and dispositioned by Engineering on February 15, 198 j The. inspector reviewed the licensee's actions and sampled reels in storage in yard areas and the plant. No problems were identified; this item is' close (Closed) Construction Deficiency Report (86-00-06): 480 Volt ITE/Teleme-canique Molded Case Circuit Breaker The_ breaker reset mechanism which is held by a slotted clip (Latch Cam).had i moved laterally and broke free, no longer providing tension to the reset mechanism. This allows the breaker operating arm to move freely and not reset . the breeker. Slotted clip cam breakers were manufactured during the period ! of October 6, 1982 through April 1, 1983. After April 11, 1983, the vendor l replaced the slotted clip design to a rectangular piece of metal with a hole ) in the-center, thus eliminating movement.and maintaining part tension, j l ,

,The licensee developed a plan to inspect all 1650 480 Volt ITE/Telemecanique molded case circuit breakers and repla::ed all latch cams of the slotted clip design with the rectangular design. The licensee's Quality Control organiza- ,

tion inspected the breaker replacement and documented their inspection result l The inspector selected a sample of ten replacement breakers and verified that they we~re of the proper design and complied with the QC inspection record This item is close !

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9 (Closed) Unresolved Item (87-17-01): The licensee has provided one hour fire wraps .to cable trays in the cable ::preading room, however, the licensee did not provide fire wrap protection to the cable tray supports. The licensee had Stone & Webster Engineering Corporation (SWEC) perform an evaluation of the cable tray support systems in fire areas CT-1, CV-1 and SB-3 (SWEC Calcu-lation No. 12241-B-226). Cable trays in these areas are protected by a one I hour fire wrap with automatic C02 suppression. Two areas, PA-3 and PA-4, have . both the cable trays and supports fire wrapped because neither area is pro- l vided with automatic suppression in the areas of the cable tray The SWEC calculations were performed to determine the time at which the auto-matic fire suppression-system would suppress a fire compared to the time at which the first support in a cable tray system would lose its load-carrying capacity (yield) from the dead weight of the cable tray. The SWEC calcula-tions used the ASTM E-119 time temperature curve to develop the temperatures for heat input into the steel support to emulate the actual fire test used to determine the fire rating of a material. The results of the SWEC calcula- < tions showed that the suppression system will suppress the fire before the ! support reaches its yield temperature.

, This item is close (Closed) Construction Deficiency Report (87-00-06): Automatic Cloring of Brown Boveri/ITE 4KV Category IE Circuit Breakers. A 4KV breaker 4KV*E10 was racked onto the bus, 125 V DC control power was applied and the springs charged the breaker immediately closing and opening it several times before the control power could be turned off. There are thirty-two Brown Boveri/ITE Category IE 4KV Breakers, Type 5HK-350 on site that have' the potential problem experienced with the 4KV*E10 breake Brown Boveri determined the breaker closing and opening was attributed to the shock resulting from the end of the closing cycle. When the closing spring completes its charging cycle, the close latch roller hits against the close latch. The resulting shock from the latch and associated linkage is such that the latch roller does not latch. This starts the charging cycle over agai Brown Boveri added an anti-shock spring to the close latch in the circuit breaker operation mechanism. The additional spring forces the clearances in the close latch and associated linkages to be taken up in one direction. This reduces the shock from the latch and linkages so that when the close latch , roller, which is part of the cam assembly, is driven by the closing springs { at the end of the charging cycle against the close switch, the latch is held { close l The Beaver Valley Unit 2 4KV breakers have been modified by Brown Boveri and ; documented on licensee's N&D 12461. During the vendor inspection of breaker 5HK-350, SN #50661F-8-08967, visual inspection of the operating mechanism identified the following conditions: l J

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. Grease in the mechanism was dry and tacky,
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The close latch appeared to be rounded off,

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.The close roller was wor Each breaker was inspected for similar problems and modified by the vendor as necessar The information was supplied to the licensee's maintenance- !

organizations for updating breaker PM informatio This. item is close i (0 pen) Bulletin (85-BU-03): Motor operated valve common mode failures during j plant transients.due to improper switch settings, i The licensee has completed the following items of Bulletin 85-03: Review and closeout the design basis for the operation of each valv ! This documentation should include the maximum differential pressure expected during both opening and closing the valve for both_ normal'and abnormal events to the extent that these valve operations and events are included in the existing approved design basi Using the results from Item a above, establish the correct switch set-tings. This shall. include a program to review and revise, as necessary, the methods for selecting and setting all switches (i.e., torque, torque bypass, position limit, overload) for each valve operation (opening and closing). Individual valve settings shall be changed, as appropriate, to those- l established in Item b abov j 1 Prepare or revise procedure to ensure that correct switch settings are 1 determined and maintained throughout the life of.the plant. Ensure that applicable industry recommendations are considered in the preparation of these procedure l The licensee has addressed items a through d of Bulletin 85-03 by letter dated .l August 15, 1986. Item e of Bulletin 85-03 requires the licensee to submit a written report to the NRC that: (1) reports the results of item a and (2) contains the program to accomplish items b through d including a schedule for completion of these items. The licensee stated that a final report ad-dressing the Item e requirements of Bulletin 85-03 will be submitted to the NRC on June 1, 1987. This item remains open pending NRC review of the licen-see's final report dat (Closed) Unresolved Item (87-11-01): Verification of the acceptability of the pipe restraint method used at fire system piping locations of direction change and under safety related tuilding ~ m

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A review of the Site Quality Control (SQC) inspection records for the yard fire protection systems don. vented the following types of. data required to be verified by SQC during the installation of the underground fire loop: I

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Approved drawings used;,

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Pipe laying, x' j

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Jointing,  %%

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Anchorage, and '

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Hydrostatic tes ' ' On a sample basis, the inspector reviewed the following SQC inspection reports for the. installation of the yard fire protection systg ,

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SQCIR#ME-MV-6018,Februarh23,1983

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SQC IR #ME-MV-6025, June 22, 1985

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SQC IR #ME-MV-6017, February 3, 1983 ,'

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The above SQC inspection reports listed the drawing number used, the hydro

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test number, bolted rods, tie rods and clamps, concrete thrust blocks and retainer type mechanical joints used. Based on the inspector's review of the if ( SQC data records, this item is close ,;  ; 4 .s (Closed) Unresolved Item (86-31-01): The Solid State. Protection System (SSPS) proof-testing determined that both SSPS Train ACand ikUV' output cards failed, thus preventing a trip signal to trip the RPS bi'ea.Wrsg s d

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m? . Investigation of the test fai,itire (Test Deficiency Ira $ rt~No. 3435) determine i that the 15 volt and 48 volt power supply leadg weh! miswired in the OR and SYNC cables. Fifty-four S$PC circuit cards wsre' det'ergined to be damaged due to the voltage reversal of the cables. ,  ; \

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i Noncompliance and Disposition Report (N&PR) No. 12407 idahtified the cables , and SSPC circuit cards that h20 to bt japaired or replacedp Test Deficiency j Report (TOR) No. 3435, December 4,.1086, listed the subsysti.Te, test procedure

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g number and system requiring retest when the SSPC circuit,ca 6s and cables verf reinstalled in the system. Retesting ,as requibd by Starttip , Manual (SUM) , Chapter 7.2, " Design Charige and Rewor's Control," Chapter 7.2.5, "Jurisdic3 I tional Tagging," related Si[M chapters' ~an(t construction Field Change Prbeedures are documented in N&DR No. 1240L and TDR No. 3435. Retest verificaticin was e documented on December 4, 1986.t This. item is close .(Closed) Unresolved Item (87-11-02): he effectiveness of fire fighting equipment within the purple switchgear ("B" train) is restricted due to the , following: x N

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The hose station is a dry pipe kystem that requires manual actuation from % s the control room to get water to flow through this station. The control s, room must first ascertain that,there is a fire in the room before turning 1 on the water, and q s , ,'

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.    -The hose station is' located inside the room thus subjecting the hose to
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In the licensee desig'n for the purple switchgear room and for each of the other. fire areas where hose stations require manual action from the control

  . room for dry pipe systems, an early-warning smoke detection system exists to
  . provide indication.of-a fire in the area.. Inspection of the purple switchgea room by the' inspector concluded that smoke detection equipment exists and has been tested.to demonstrate its operabilit l     . .
  (The' licensee has designated stairwell locations autside the following affected Y,. rooms that will have additional lengths of hoses; mounted on standard fire
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East (purple) and West (orange) emergency switchgear rooms, Elevation l 735'.

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Control' Building Cable Turael, elevation 712' i

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Cable Tunnel / Cable Vault, elevation 735'  !

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[ ' d -Sprvice Building Normal Switchgear Room, elevation 760' _

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  ,- Cable Tunnel / Cable Vault, elevstion 755'
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3 Cable Vault and Rod Control Building - Contiguous Areas, elevation 718' 3l\ --

   -Service Building, elevation 780'
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Main. Steam Valve Building

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Primary Auxiliary Building

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Condensate Polishing Building

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Waste Handling Building

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Turbine Building This item.is close (Closed) Violation (87-02-02): Missing bolts on HVAC fan supports. This item y- pertains to bolts that were not installed for HVAC fans (2HVD*FN271A and B) O to properly atta::h them to their structural steel support The root cause ! of the failure to detect this problem was that the QC Inspection Plan (IP) l 7.4.1 did-ant include specific instructions to verify the bolting of HVAC i i a

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I equipment te its supports. In addition to making an immediate change to IP 7.4.1, the licensee's corrective actions included a comprehensive reinspection of; dWQA Category I HVAC equipment installations. The inspector reviewed the

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re Mu and corrective actions based on this reinspection; no additional de-ficiencies were identified. Also, the inspector specifically reviewed the-

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ccapleted work on HVAC fans 2HVD*271A and B and identified no deficiencies, This itcm is close k I b '

    (Closed) IE Bulletin 80-22: Automation Industries, Model 200-520-008 Sealed-
  . , . Source Connector This Bulletin involved a specific type of connector used 1
    or some radiography sources which was found to become separated with serious  ,

exposure potentiai. The licensee conducted a review in 1980 of the on-site cconectors and determined that none of the referenced referenced conductors  !

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W re on site. A followup. review conducted during this inspection period con-firmed that the licensee did not possess or use any of the referenced connec-

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a '(Closad) Unresolvec' ltem (87-12-01): Containment Integrated Leak Rate Test '

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W Uli) report review. The CILRT was performed in February,1987, and this  !

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item :.;volved the NRC review of the test results as documented in the licen- i s /) see's fctmal test report. The ClLPT report was reviewed for conformance to j O' '

 /   10 CFR 50 ~ Appendix J requireinents and consistency with IE Information Notice
,  'h  85-7 No deiigiencies were identified; this item is close !
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1 Site Actnities . . l'-  ; 1 ,

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Throughout pa inspe,dion period, the i.vpectors toured the licensee facili-  ! ties. ' Gen.wal work ,activRies were observed including construction, surveil-  ! lance,ftesd ng and reintennnce.. *The in!pectors also monitored the licensee's i

-  a   housekeaoing, security and' preliminary radiation control activities. In par-  !

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tituir, tha ins 6tetors mon',tored the Mcensee's implementation of the Site f

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Securliy Man.(see detail M and the ren.lsal of the control room wall (see

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j During tni's'9e-ipd, the licensee successiblil(conducted the site blackout portions of. tie Loss Of Offsite Powar Test (L0m}. The' LOOP is a major pre- 1 operational test which confirms thac the emergency diesel generators will  ! } s start on Jeannand that required loads will receive.couer in a controlled f

3 j.up.enct' fseIdent9adspecialistinspeatorswitnessedthe'keyporthnsof l 0 ; the LOOP wtnch wairoione train at a tim The inspectors reviewed t.he test ! Arocedwes, procedure adherence, test persovel knowndge, operator a,vareness  ;

 *-   , M 'e'quipmen t perfmrmce. Fac harc' ware deficiencies e"e identified; further  {
  ,  inipection details se conte %ed in Inspection Report SU411/87-3 !
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   ' Also, durire this irdp(ctior, the licensee completed the MSI'/ replacewat  l begun oi. Mercb l6,. RS The rew Atwood & Horrill "Y" patte m globe' valves ;
, c    were instdkd anitie hy&q comp %Mabead of the licensee's scheduled May
,(    21, 1987, j'uel load S te. 'lieinspadtajswitressedtheremovaloftheorigi-  J
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6 , M znall Gulf & Western ball-type valves, the structural modifications made to

  : accommodate the new, wider actuators, the. installation and routing of control 7   cable, the insta11ation of the new valves, and the hydro of the-associated Category I piping. No ' deficiencies were' identified.

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  ;0n May 18, 1987, th'e licensee identified two items as potentially reportable
  .under 10 CFR 50.55(e) which involved the environmental qualification of cer-
  'tain seals.'and connections. These items'were generally resolved by May 27,.
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  ?1987, and issuance of a low power license was. expected within a day or s .
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  (See detail 10. for additional information).

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  .No violations were identifie . TMI Action Plan Requirements (NUREG 0737)

Licensee commitments'-in response to TMI Action Plan requirerrents:have been -

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reviewed by the staff and are documented in the BVPS Unit 2 Safety Evaluation j Report (SER):and its-supplements. Several~of the TMI items are still under- 1 review and items may require further attention if significant changes become j T necessary. : During this inspection' period, the inspectors verified the licen- 1 see's compliance with the following items:

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  -II.B.4 Training for Mitigating Core Damage a

The licensee was required to develop a training program te teach the use of I installed equipment and systems to' control or mitigate accidents in which j

  . severe core. damage-is postulated. Shift Technical Advisors (STAS) and operat- .!
  ' ing' personnel from the plant manager through the operations chain to the j-licensed; operators are required to be trained for mitigation of core damage, I while supervisors and technicians in.the radiation control, chemistry and instrumentation and control:(I&C)' departments are required to receive training in mitigation of' core damage commensurate with their. responsibilities. The NRC. Safety Evaluation Report:(SER) for'BV-2 found that the licensee's-program
  : complies with;the requirements.of this ite :

The. inspector reviewed the mitigating core damage program as implemented, i Text.and course' material-used for the training were found to be adequat 'The program identified'the job classifications and requirements for partici-pation' The station mitigating core damage training program provided to

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,   radiation control, chemistry and I&C departments, consists of six lectures 1 covering the six required subjects and is presented in two year cycles. The :

licensed personnel and STA mitigating. core damage program is included as part i

  :of the overall initial and requalification training programs. The licensee-has met.the requirements of TMI Item II.B.4. This item is close II.B.2 Design Review of Plant Shielding   :

LThe licensee was required to perform a radiation and shielding design review v of the spaces around systems that could, as a result of an accident, contain , highly radioactive material. The review was to identify the location of vital t

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!~ ( areas and equipment, such as the control room, radioactive waste control ) j stations, emergency power supplies, motor control centers and instrument areas j J in which personnel occupancy may be unduly limited or safety equipment may ' be unduly degraded by the radiation fields during post accident operations ) of these systems. The licensee's plant shielding design review was submitted j to NRC Licensin The NRC Safety Evaluation Report (SER) for BV-2 reviewed the licensee's submittal, which addressed areas requiring access following ]; an accident and identified location, occupancy requirements and maximum ex- j pected dose levels. The SER concluded that BV-2 meets General Design Cri-  ; terion 19 for vital areas requiring extended, continuous and infrequent occu-pancy. The inspector toured specific plant areas where access would be re-quired following an accident. These areas included the control room, hydrogen analyzers, post accident sampling system area, hydrogen recombiner area, safeguards building manual valve area and RHR suction valve transfer are The inspector confirmed that equipment in these areas requiring local opera-tion were installed in shielded area The FSAR identifies doses which would be expected for occupancy in specific l vital areas following an accident, based upon the specific routes and occu- 4 pancy time The inspector verified that the licensee is currently developing l a program which includes generic radiation work permits (RWP), with the as- j sociated survey and dosimetry requirements, predicted dose rates, maps and  ! travel routes for the required areas. The program is being tracked by the licensee's internal tracking system and is expected to be completed in July, . 198 The program will be implemented via the Emergency Operating Procedures Chapter of the Radiation Control Manual (RCM). The inspector had no further questions. This item is close II.E.4.2 Containment Isolation Dependability The Containment Isolation System (CIS) allows the normal or emergency passage of fluids through the containment boundary preserving the ability of the boundary to prevent or limit the escape of fission products that may result I from postulated accidents. Containment isolation at BV-2 is accomplished in l two phases. The containment isolation phase A (CIA) signal isolates all non-  ; essential system lines penetrating containment. The containment isolation i Phase B (CIB) signal isolates the component cooling water supply / return lines  ; for their reactor coolant pumps, control rod drive mechanism shroud coolers and service water lines to the containment recirculation air coolers. The  !' SER for BV-2 reviewed the licensee's containment isolation provisions against NUREG-0737 Item II.E.4.2 requirements and concluded that the licensee's system was acceptable. The inspector reviewed Preoperational Tests P0-P. 01A.04 (Safeguards Actuation Test) and P0-2.36A.02 (Electrical AC Independence Test).

Between the two tests, both the CIA and CIB (both trains) were functionally tested. Although minor test deficiencies were identified, no major problems were encountered with the containment isolation system. Additionally, the i individual containment isolation valves were functionally tested during their j respective system preoperational tests as a part of the normal system turnover process. The test deficiencies were noted, scheduled for resolution and _ __ _ _ _ _ - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - . - - - - - - - - . - . - - _ _ ------ ..- --------- _ --- _ _ _ s

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tracked by'the preoperational' test group. Resolution'of the minor-test;de-n- ficiencies will be routinely checked by resident / specialist inspections. This em item is close 'II'.B.1 Reactor Coolant System Vents ' The licensee was required to install. reactor coolant system (RCS) vents on the reactor vessel head and pressurizer to vent noncombustible gases from the RCS which may inhibit core cooling during natural circulation. The-vents' were required not to represent an unacceptable increase in the probability.of a

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loss'of,~ coolant accident (LOCA) or a challenge to containment integrity. The BV-2 Vent System consists of three power operated' relief ~ valves (PORVs) an associated block valves:for the pressurizer and a head vent system with two parallel flow paths, with1 redundant isolation valves for the reactor. The SER for BV-2 concluded that-the design for the vent paths for the reactor-vessel head'and pressurizer is. acceptable. The inspector reviewed the BV-2 Technical Specifications (TSs). The BV-2 TSs are modeled after those for- I BV-1, except that the RCS vents TS for BV-2 includes only the RCS head vent 1 system,' while the pressurizer vent portion of the RCS vent system is incor- ' porated:into the PORV TS (3/4.4.11). It was also noted that the TS 3/4.4.11

 'does not require that the licensee verify flow through the pressurizer vent- I path for the RCS, as it:is required for the RCS head vent system. This dif- :

' ference Lis due to the difference between the BV-1 and BV-2 pressurizer head vent system. The pressurizer head vent' system for Unit 1:uses a dedicate { system for venting the pressurizer, while the BV-2 system uses PORVs as its i vent flow path. This' difference was reviewed'and accepted.by NRC Licensin Implementing procedures for the RCS vents were also reviewed by,the inspector, i

The procedures met the requirements'of Plant Technical Specifications and TMI L

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Action Item'II. ( During review of Procedure 2.6.4.E, Filling and Venting the RCS,'it was noted that a' mechanism to independently verify that RCS . vent valves are closed was 1 not evident. This concern was brought to the licensee's attention, who stated 1 that plant procedures are currently being reviewed to verify that the actions / alignments requiring independent verification have been satisfied. The in-spector. reviewed the Preoperational Test for the reactor vessel head vent system (P0-2.06.13). Minor test deficiencies were noted on the test results report.and are tracked and resolved by a licensee internal open item tracking ) n lis The preoperational test for the PORVs (P0-2.06.06) was also reviewe I The same open items tracking list exists for the pressurizer head vent syste Neither test contained significant deficiencies. ' Implementation of the TS requirements, necessary procedure provisions and resolution of test deficien- ! cies will be reviewed through routine resident inspection. This item is E . close .B.1.2 Independent Safety Engineering Group i The establishment of an Independent Safety Engineering Group (ISEG) was to , be performed for all applicants for an operating license. The licensee de-veloped an ISEG in accordance with the TMI Action Plan requirements, as docu-I i

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mented in' the BV-2 FSA The group, consisting of five dedicated, full-time-members, will function to examine unit operating characteristics, NRC issu-ances, industry advisories, Licensee Event Reports and other sources of unit '

 ' design / operating experience 'information. The requirements for the ISEG are outlined in' plant-Technical-Specification 6.2.3 (ISEG).

.The~ inspector' reviewed the ISEG charter and implementing procedure Member-qualifications.were verified to meet the requirements set forth in. the charte The.ISEG'is currently fully implemented and functional, and serves both BV-1 and BV-2. Instructions and guidance are.provided to ISEG members for assign - ment.and completion of formal'ISEG plant observations and for assignment, tracking and review of formal ISEG evaluations. This item is close : Diesel Generator Instrument Vibration The ability of an emergency diesel generator.to fulfill its safety function '~ depends on, among other. things, the proper functioning of its controls and' monitoring instrumentation. Major diesel engine damage has occurred at other sites from vibration-induced wear on engine controls and monitoring instru-

 : mentation. Prolonged periods of' vibrational stresses are.possible with in-ternal combustion engines and can impact the' calibration, accuracy, and con--

trol signal output of sensitive instrumentation. The licensee committed, in a letter dated March 5, 1985, to monitor the vibration levels of diesel instrumentation during diesel generator preoperational testing. 'Instrumenta-tion found to be experiencing vibration levels in excess of vendor recommen-

 ' dations 'would be' relocated as necessary to achieve acceptable _ levels of _

vibratio The . licensee took data on vibration levels during the full'l'oad endurance runs of the diesel-generators in late 1986. Further, full load testing is planned during: low puwer testing and instrument vibration _ levels will again be moni-tored. ; Unacceptable vibration levels will be addressed by instrument reloca-tion and retesting as necessary to resolve this long term potential; safety concern. NRC resident and specialist inspectors physically inspected the'

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engine. instrumentation during the 1986 testing and during the diesel auto E starts associated with the site loss of offsite power testing on May 16-17, 1987. Observed vibration levels were found to be low indicating that no im-mediate concern exists with respect to instrumentation operability. The re-sults and evaluation of the licensee's vibration data will be' reviewed in a later inspection for potential long term effect No violations were identifie . Control Room Activities During the inspection period, the licensee accomplished various construction and testing activities concerning the Unit 1/ Unit 2 control room envelope-and ultimately achieved a common control room configuration. Prior to the com-plete removal of the temporary wall between the Unit 1 and Unit 2 control rooms, a pressure test was conducted on April 26, 1987, on the combined Unit

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1'and Unit'2 control room envelope in accordance with Startup' Proof Test Pro-cedure No.=2T-HVC 44A-2.14. Based on discussions with test personnel and a j review of the test results, the inspector determined that portions (Unit l' process rack, Unit I relay room, Unit 1 equipment room) of the common control j j room envelope did not' attain a positive. pressure with respect to certain ad-

.jacent areas. This was attributed to leaks.in the Unit 1 ventilation ductwork' -i l

and air balancing problems in' the Unit 1 control room ventilation systems.' The licensee performed ductwork repairs, sealing and air balancing for the: j

Unit'l~ systems. Retesting was satisfactorily perfor.ned on May 18,.1987, j e during the performance of Section VII.E.:of Preoperational Test Procedure l

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'2.44A.01 "UnitL1/2 Control Room Emergency Pressurization Tests."' Specific-- -{

ally, the licensee confirmed that' the control room was at a higher pressure i

'than each adjacent area by.' opening appropriate Unit 1 and Unit 2 control roomi 'l envelope doors to each adjacent area and. confirming' air flow out of the con- I trol . room,     d

1 The inspector reviewed the test deficiency record for P0-2.44A.01 and deter-' mined-that the remaining deficiencies were being properly. tracked for.~ correc- 'l tion. Of the~32 test deficiencies listed, 13 were stillEopen but resolutions

i were.being actively pursued. The inspector had no further concern !

.No violations were identifie , Implementation of Site Security Plan    i q

On April 13, 1987,.the licensee formally implemented th'e Site Security' Plan ] such that Unit 2 was within the protected area. Some minor problems were l experienced but within a few days, operations had stabilized. The licensee' j maintained the interior fence separating the two units but with one place for inter-unit passage with a continuously manned guardpost for additional contro of. access to Unit 1. These measures were effective in providing a' gradual phasing in of security while preventing construction workers from impacting activities at Unit 1 which was then at full power.

" Vital access control was established as keycard readers were made op'erationa j Some problems with individual readers and the lockdown of certain security 1 doors were. resolved in an orderly manner. At.the end of the inspection period, ' with the issuance of a license imminent, the security. system' was fully opera- : tiona Further details of the licensee's progress in implementing the i security plan are presented in Inspection Report 50-412/87-39, , No violations were identifie ' , 1 Environmental Qualification of Seals On May 18, 1987, the licensee reported under 10 CFR 50.55(e) that 50 seals were not installed in accordance with field installation specifications and might not meet applicable environmental qualifications (EQ) requirement , Potentially affected were 27 solenoid operated valves (SOVs) and 19 resistance i thermal detectors (RTDs) as reported in' Construction Deficiency Report (CDR) ;

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87-00-16'and 4 Reed' position' indicators as reported in CDR 87-00-1 These !,', items were identified during the licensee's plant-wide EQ review. The NRC l staff deferred issuance of an Operating License until the EQ deficiencies were W repaire In the case of the-27 S0Vs, the cable jackets were required to be' removed' prior to seal applicatio The engineering intent was to'have all. jacket mater.ial removed,' but for these items although the overall cable jacket was : removed, the individual conductor braided jackets were left' intact. The lic-ensee conducted.some preliminary tests which confirmed that the seal with the'

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braided con'ductor jacket could not meet EQ requirements. The affected S0V

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seals were replaced with the exception of .four seals on the reactor' vessel head which could not beLdone until the head was installed. After further review of the 19 RTD seals; the licensee determined that those seals were not

 ' affected by the deficiencies ' identified on the SOV seal In CDR 87-00-17, the licensee determined that a required QC witness point had been missed for 4.of the.6 seals on Reed position indicators for the pressuri-zer power operated relief valves. The witness point involved the application of graphoil tape to prevent moisture intrusion along'the exposed threads of the electrical raceway seal assemblies. The licensee' elected to add a second seal of a'different' type.to assure full protection against moisture intrusio This redundant seal:is composed of a' qualified ~ sealant on the threads covered by a stainless steel hose clamp clamped'on to provide a shield for the sealan ;

to'any beta radiation. On May 25, 1987, the licensee reported the completion of-all corrective actions on the CDRs with the exception of the four seals '. on the reactor vessel head. Licensee actions, -including the 10 CFR 50.5S(e) < final reports,-will be reviewed in a future inspectio No violations were' identifie . . Site Meeting On May 11, 1987,'NRC Chairman Lando Zech, and Region I Administrator W!111am Russell, conducted a site visit to Beaver Valley. The Chairman reviewed the status of Unit 2 with respect to readiness to receive an operating license,- observed simulator drills, toured the site and talked with all available chemists, licensed operators and other personnel. The Chairman also visited the adjacent Shippingport' site to observe the-status of decommissionin On May 18, 1987, the NRC Near Term Operating License (NT0L) Panel made a site visit and received a licensee briefing on readiness for licensing. The NT0L panel consisted of William Russell, Region I Administrator; Bruce Boger, Assistant Director for Region 1 Reactors, NRR; John Stolz, Director of Project , Directorate I-4, NRR; William-Kane, Director, Reactor Projects, RI; Edward ' Wenzinger, Chief, Projects Branch 3, RI; Peter Tam, Beaver Valley Project-Manager, NRR; and the site resident inspectors. The NT0L panel conducted a site tour to witness work in progress as well as to review the status of project completio l l I

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  ' Me'etings were held by the'resideilt inspectors with senior: facility ' management L'  ,

L-periodically.during the course of this. inspection to discuss the inspection

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with the licensee at3the conclusion.of.the report' perio ' ,

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