IR 05000412/1986044

Insp Rept 50-412/86-44 on 861201-870116.No Violations Noted. Major Areas Inspected:Hot Functional Testing & Auxiliary Feedwater Sys Testing.Weaknesses Identified in Implementation of QA Program to Resolve Field Deficiencies

ML20211P656
Person / Time
Site:	Beaver Valley
Issue date:	02/06/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20211P551	List: IR 05000412/1986044 ML20211P546
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.3, TASK-2.G.1, TASK-TM 50-412-86-44, IEIN-83-69, NUDOCS 8703020409
Download: ML20211P656 (14)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-44 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: December 1, 1986 - January 16, 1987 Inspectors: J. E. Beall, Senior Resident Inspector A. A. Asars, Resident Inspector W. M. Troskoski, Senior Resident Inspector, BVPS Unit 1 L. J. Prividy, Resident Inspector B,7.Dav'dson,RadiationSpecialist S

Approved by: -

. ,13 0 2 7 97 Lv. E. Trippf Chief, Reactor Projects Section 3A / Otte Inspection Summary: Inspection No. 50-412/86-44 on December 1, 1986 - January 16, 198 Areas Inspected: Routine inspections by the resident inspectors (372 hours0.00431 days <br />0.103 hours <br />6.150794e-4 weeks <br />1.41546e-4 months <br />) of licensee actions on previous findings, site activities, Hot Functional Testing, auxiliary feedwater system testing, regenerative heat exchanger weld indications, preservice inspection activities, management meeting on main steam isolation valves, reactor protection system testing, fire wrap of HVAC ductwork, radiation monitoring system reviews, and TMI Action Plan Item resolutio Results: No violations were identified. Weaknesses were identified in implemen-tation of the QA Program in resolving licensee identified field deficiencies and assuring cable storage adequacy (See Details 9 and 14).

8703020409 870211 PDR ADOCK 05000412 O PDR

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DETAILS 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 . Project Status Summary Construction activities are currently estimated to be 98.1% complete, with 452 of 476 subsystems turned over for flushing and proof-testing. For soft-ware, about 99 out of 114 preoperational (P0) and initial startup tests (IST)

have been approved. 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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Integrated Leak Rate Test February 9,1987

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Loss of Power Test March 16, 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 Due to early completion of HFT, the licensee advanced the milestone date for Integrated Leak Rate Test and rescheduled the Loss of Power Testin . Inspection Program Status Summary Preoperational Test Program Inspection completion status is approximately as follows:

% INSPECTION COMPLETE AREA END OF THIS PERIOD END OF LAST PERIOD Overall Program 40% 35%

Procedure Reviews:

Mandatory 50% 35%

Primal 100% 50%

Test Witness:

Mandatory 35% 15%

Primal 100% 5%

Results Review:

Mandatory 15% 15%

Primal 5% 5%

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This inspection status is consistent with the applicant's test program pro-gres NRC inspection items as listed below:At the end of this inspection period, there NO. OF OPEN INSPECTION ITEMS TYPE OF ITEM END OF THIS PERIOD END OF LAST PERIOD Bulletins

7 Violations

5 Deviations

0 Construction Deficiency Reports 17

Unresolved and Inspector Follow 30

TOTALS 59 68 Licensee Actions on Previous Inspection Findings (Closed) Inspector Follow Item (77-SP-05): Perform inspection in accordance with Temporary Instruction 2500/1 (followup to IEB 79-15 (Closed) Inspector Follow Item (86-01-01):

open items resolution on the preoperational test progra Review effect of SER Sectio BVPS 2 SER Supple-ment 3 was issued in November 198 licensee adequately dispositioned eight of the nine open items.The SER Suppleme ing item is currently being tracked by Confirmatory Issue 52(a).This remain-mitted by the licensee.three more items were identified during the review of the inform Issues 52(b), (c) and (d).These items are also being tracked by Confirmatory Based on the resolution of the eight items and four issues, this item is closed.Issues to track resolution of the additional the existence of Confirmatory (Closed) Temporary Instruction (25-00-12): Inspection of the actions taken by the licensee and applicants of BWR facilities with Mark I and Mark II con-tainments in response to GE SIL 402 on Wetwell/Drywell Inertin This TI is applicable to BWRs only and is therefore close (Closed)for program Temporary fiscal year Instruction 198 (25-00-13): LWR water chemistry inspection This TI has been replaced by routinely con-ducted inspection program procedures and is therefore closed. for LWR chemistry control and chemical analysis (Closed) Tempsrary Instruction (25-00-14): Inspection of the location of the manual trip circuit in Westinghouse-designed plants with a solid state pro-tection syste The Westinghouse SSPS trip circuit was inspected and con-

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trolled drawings were reviewed during February 1986 as documented in NRC In-spection Report 50-412/86-03, detail 8. No deficiencies were identified and this item is close (Closed) Temporary Instruction (25-15-64): Near-term inspection followup to Generic Letter 83-28, Required Actions Based on Generic Implementation of Salem ATWS Event This TI provides for inspection of post-trip review, equipment classification, vendor interface, and maintenance programs for selected safety-related components within safety-related systems. In November 1986, an inspection was conducted in accordance with this TI and documented in NRC Inspection Report 50-412/86-39, detail 3. The inspectors concluded that licensee actions are consistent with the requirements of Generic Letter 83-28 with the exception of independent testing of the shunt trip attachment. The licensee is currently revising the shunt trip testing procedure and it will be reviewed in a future inspection. Also, the inspectors will observe actual surveillance testing to be performed in accordance with the revised procedur This item is close (Closed) Temporary Instruction (25-15-67): Survey of licensee's response to selected safety issues. This TI was a survey of operating plants conducted to determine what actions licensees were taking to address several safety issues. This TI expired in August 1985, and BVPS 2 was not operating at that time. This item is close (Closed) Construction Deficiency Report (86-00-16): Equipment qualification limits exceeded for main steam pressure transmitter. BVPS 2 SER requires that the licensee complete an analysis of safety-related equipment affected by a steam line rupture with a superheated steam condition. The analysis concluded that the main steam line pressure transmitter has EQ limits which will be exceeded under these circumstance BVPS 2 SER, Supplement 3, has assigned Confirmatory Issue 51 to track this analysis and subsequent licensee action This item is close (Closed) Inspector Follow Item (86-04-06): Review of test specifications for P0 2.13.02, Quench Spray Pumps and Controls Tes During a previous review of P0 2.13.02, the inspector identified several parameters and setpoints which had not yet been specified. The inspector reviewed the most current revision of the P0 and verified that several specifications have since been included and additional revisions are being processed. This P0 is tentatively sched-uled to be performed in the beginning of February, 198 (Closed) Inspector Follow Item (84-14-03)- Engineering Confirmation Progra This item had been opened to monitor the performance of the Engineering Con-firmation Program in endorsing the plant design bases. In the past two years, the inspector followed program progress and interactions between Duquesne Light and Stone and Webster engineers. The major efforts of the Engineering Confirmation Program have been completed. The program has and is satisfying its objectives and commitments. This conclusion was also made by the Engi-neering Assurance inspection as documented in NRC Inspection Report 50-412/

86-23. Based on these reviews, this item is close _ _ . _ _

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(Closed) I/E Bulletin 80-03: Loss of Charcoal from Tray Absorber Cell This bulletin involved Flanders Type II filters which had a significant loss of charcoal due to a poor filter retention screen / frame attachment design. BVPS 2 uses American Air Filter charcoal absorbers in the Control Room Emergency Air Filtration Systems and in the Auxiliary Building Exhaust and Leak Collec-tion Filtration Systems. The American Air charcoal filter retention screen is secured to the filter frame by spot welds that are spaced 1-1/2 inches apart, in contrast to the 6-inch spaced, riveted design of the Flanders Type II filter The inspector met with various project personnel, inspected the installed charcoal absorbers, and concluded that this bulletin should be close . Site Activities Throughout the inspection period, the inspectors toured the licensee facili-ties. General work activities were observed including construction, surveil-lance, testing and maintenance. The inspectors also monitored the licensee's housekeeping, security and preliminary radiation control activitie During the inspection period, several inspectors observed improvements in the Startup Group's control of preop test activities. This is attributed to the

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recent assignment of individual test engineers to act as lead test engineers for each preop test and, in some cases, lead for all preop tests for indivi-dual systems. With this organization, the status of systems and preop testing was more easily identified and tracked. The inspectors concluded that it was more effective to have one knowledgeable contact for each preop tes The inspectors attended the routine plan of the day meetings on a sampling basis. The agenda of these meetings consists of status reviews of ongoing preop tests, construction proof tests, and other work activities affecting plant statu During Hot Functional Testing, the scope of the meeting focused on major plateaus and hold points during the test. Since that time, attention has been focused on the next major test, specifically CILRT. For miscellaneous items which affect testing, an active punchlist is used. This method of con-trolling testing activities and monitoring test status has proven effectiv No deficiencies were identifie . Hot Functional Testing Hot Functional Testing (HTT) was in progress at the beginning of the inspec-tion period. The inspector reviewed Control Room operator and Preop test logs, witnessed selected tests and evolutions, and made frequent plant tours. Dur-ing HFT, the unit was heated up with reactor coolant pump (RCP) heat, the main turbine was rolled, the generator was energized and, on December 7, 1986, the generator was briefly synched to the grid. On December 21, 1986, the unit was restored to the cold condition and HFT was declared complet _ _ _ -

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During HFT, the inspectors witnessed several thsts involving the Auxiliary lFeedwater (AFW) syste The tests included auto starts and 48-hour endurance runs of the steam-driven and the two motor-driven AFW pumps. During these tests, the inspectors reviewed licensee data and made independent observations of piping movement, pump vibration, bearing temperature, bearing lube oil delta T,,and other parameters. For additional details, see Section The licensee prentice of conducting twice-daily meetings continued during HF The meetings shifted emphasis f, rom meeting HFT pre-requisites to problem solving and schedule adjusting. The licensee effectively resolved problems with pressurizer heaters, RCP vibration, steam ge;,eratur chemistry, diesel load sequencer interlocks, inadvertent valve operation, and the generator exciter, and made schedule changes which allowed all the required tests to be performed with HFT complete about a week earl '

No deficiencies were identifiede .

N Auxiliary Feedwater System Testing Major portions of AFW testing occurred during this inspection period. This included turbine driven auxiliary feedwater pump (TAFP) 48-hour endurance test, several TAFP cold auto starts, AFW flow measurenents' at various steam genera-tor levels, and motor driven auxiliary feedwater pump (MAFPF auto starts with feed line vibration monitoring. The inspectors obsersad many' portions of these test During lengthy operation of the AFW pumps, the inspectois monitored system parameters and also pump characteristics including lube oil temperature change across the pump bearings. No abnormal conditions were identifie s

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Of particular interest were the auto starts of the AFW pump The"laspecters observed both TAFP Terry turbine steam admission and AFW pur.!p discharge 1(^es for signs of excessive vibrations or displacement which could be inhicatiVe of conditions conducive to water hammer. The preop test procedure required monitoring of the pump dischurge lines but no the turbine steam admission line. After discussions with the inspector and viewing system configuration with the aid of the plastic plant model, the licenske measured steamlin.e de-flection with vibration monitoring instrumentation. For sevfral cold auto ~'

- starts of the TAFP, steam line vibrations'were negligible. The AFW sy'ste,m susceptibility to water hammer in the pump discharge lines was tested by in-ducing low low steam generator levels which initiated MAFP autp start Licensee personnel and NRC inspectors were situated at several' points along the line and did not identify any vibrations or movement suggesting water hammer condition <

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Currently, there are severa These include all three AFW'lpumps test deficiencies falling below open theinvolving design headthe AFW curves Systehw and safeguards area humidity too high while running the TAFP. The inspectors are following licensee resolution of these test deficiencie ,

No deficiencies were identifie :

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, Regenerative Heat Exchanger Weld Indications During October 1986, the licensee conducted a preservice inspection of the regenerative. heat exchanger (2*CHS-E23) girth welds. At that time, two welds

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" (Numbered 8 and 11) were identified a's containing ultrasonic reflectors which

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3. , required further evaluation. As is routine _in cases such ar. this, the_licen -

E see obtained the' vendor radiographs for the welds in an effort to determine

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the significance of the ultrasonic indication The vendor, Joseph Oat & Sons,

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Inc., radiographs showed linear indications which were classified as incomplete fusion areas. Based on this finding, the licensee reviewed all of the' vendor

, radiographs for 2*CHS-E23 to determine if any additional welds contained

' similar indication A final report detailing deficiencies and inconsisten-

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- cies noted during this review was issued on November 26, 198 Several in-stances of radiograph reader sheet date irregularities and welds containing -

areas with a lack of fusion were identified. The licensee made the determina-

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L tion to re-radiograph the welds which appeared to have questionable indica-4.

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tions.. These were welds numbered 2, 7, 8 and 11. On December 26, 1986, radiographs were taken for welds 2 and 8; however, welds 7 and 11 were in-

!. acces::ible for radiography due to interference with the heat exchanger tubo sheet. From review of the radiographs for welds 2 and 8, the licensee deter-  ;

r mined that weld 2 is satisfactory and weld 8 does contain an area with a lack

, of fusion and is therefore, unacceptable. This conclusion prompted the re-t porting of Construction Deficiency Report 86-17 in accordance with 10 CFR

50.55(e). The unacceptable portion of weld 8 will be ground out and reworked.

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To determine 'the condition of welds 7 and 11, the licensee brought in NDE consultants from Westinghouse. They were tasked with developing a technique of ultfasonic examination which could positively identify the unacceptable area in weld 8. This same technique would then be applied to welds 7 and 11.

]' The consyltants stated that they were able to positively identify the unac-l ceptable area-of weld 8 using standard industry NDE method Using the same

method, welds 7 and 11 were determined to be acceptabl Currently, the licensee has issued N&DR 29,986 to Stone and Webster Engineer-

! ing. This N&DR details the available information on 2*CHS-E23. Stone and

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Webster Engineering will determine how this N&DR is to be dispositioned. The .

l inipectors will follow the resolution of this issue in future inspection ,c Non-Destructive Examination Activities l' :

TNelicenseehasseveralongoingactivitieswhich,asawhole,determineand

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document the acceptability of piping, supports and welds. These activities include stress calculation reconciliation, ASME III review (N-stamp), ASME I^ _ XJ inspection (baseline) and system walkdown. During a review of QA corres-

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pandence, the inspector determined that certain safety-related system welds

, had been identifled in April,1986 as not meeting ASME III requirements, but 7 that no action had been taken as of December 1986. The weld deficiencies had

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been identified by personnel performing pre-service inspection (PSI) for ASME

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XI baseline data and had been documented in memos at that time.

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The apparent cause of the licensee's failure to address the weld deficiencies was a difference of professional opinion within.the QA organization. -The jurisdiction of PSI . inspectors making inspections under ASME XI to make' find- -

!, ings under ASME III was questioned instead of the technical validity'of the findings themselves. The inspector expressed his concerns to the Senior Vice President, the senior licensee manager on site, rand requestea a meeting with-

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the. involved parties in a week's tim fj -

The meeting was held on December 22, 1986, and was attended by the Senior-Vice President and members of his staff including the'QA Manager. The licensee's initial position was that ASME III examinations were complete and that only the less restrictive ASME XI criteria were applicable. The inspector asked which welds were N-stamped at the time of the ASME XI inspection and which were N-stamped as of the time-of the meeting. The licensee did not know; the

!. inspector questioned the licensee's assertion that ASME III did not apply if

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the N-stamp status was not known. Further research revealed that none of the welds had been N-stamped. .The inspector requested the licensee to address the potential weld deficiencies technically, not administrative 1y, that the

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, disposition of each deficiency be. documented, and that the documentation be

, made available for NRC review.

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The licensee informed the inspector at the time of the exit meeting of this inspection on January 20, 1987, that the welds'would be re-examined. The

examination would be conducted by qualified inspectors from both'the PSI team and the ASME IIILgroup at the same time. Weld deficiencies identified would

. be repaired as necessary. The. inspector expressed the concern that the lic-ensee's-QA Management did not evaluate and disposition this item in a timely -

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and technical manner prior to NRC, involvement. This item is Unresolved

' (86-44-01) _ pending further review of this item, including re-examination, disposition and repair. Section 13 also discusses weaknesses identified in

implementing the licensee's QA progra . Management Meeting on Main Steam Isolation Valves The Main Steam Isolation Valves (MSIVs) at BVPS 2 are ball-type valves (24

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inch) as opposed to Unit 1, which uses a combination of check valves as MSIV The Unit 2 valves are similar to the MSIVs at Niagara Mohawk's Nine Mile Point

Unit 2, which is a boiling water reactor (BWR). Problems have been experi-

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enced including ball scoring, roller bearing failure and potentially inade- ,

quate trip force. Additional details on the history and current status of the Nine Mile Point, Unit 2 MSIVs are contained in Inspection Report N /86-5 Representatives from the licensee, including members of senior management,

, met with Region I management and technical staff at the NRC offices in King of Prussia on December 15, 1986, to discuss a modification and testing program

, intended to resolve the identified MSIV problems. The licensee's presentation

included a part-by part analysis of MSIV components, in process changes in j materials, hardware and settings, and proposed test program objectives and f

acceptance criteria.

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'The MSIVs'were being reassembled at the close of the inspection. perio The licensee currently plans to complete valve reinstallation, cycle each valve several times, and reinspect-the valve balls before the end of February, 198 .~ Reactor Protection System Testing

, Nuclear Instrumentation System-Discussions with testing personnel indicated that one of the secondary objectives of the Hot Functional Test was to allow for troubleshooting -

of any electrical ~ noise signals picked up by the Source Range Monitor System. Because the new crimp on style cable fittings were not available L

in time to support the tests, temporary connectors (compression fittings)

were used on the triaxial cables while under the temporary jurisdictional control of the Startup Group. Discussions with the Startup Group indi-cated that Test Package No. 2 index tracking numbers had been assigned  :

to these temporary connections to insure that the correct, qualified connectors would be permanently installed after testing. Electrical Test Procedure 822 would then be used to verify the cable integrity. Both the intermediate range monitor and power range monitors are scheduled'

to use the new connectors; temporary ones were not installed as their detectors are ion chambers, which are not as susceptible to nois All Nuclear Instrument Containment Penetrations (2RCP 3B, 6B, 21E, 19E),

failed the vendor recommended testing and all the source range monitors were reading a constant output of IE4 to 1ES cpm. Troubleshooting found that the noise was eliminated when the outside containment penetration

pigtail was changed. However, 3 to 4 days later, the same problem re-curred. Further testing of the Westinghouse Model Type III 600 Volt

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penetrations indicated that, all cable resistance tests : failed to meet the acceptance criteria (conductor to inner-shield, conductor to outer-l shield, inner-shield to outer-shield) plus several penetrations had apparent insulation breakdown Appropriate N&DRs were issued; all of the Westinghouse tri-ax containment penetrations (4 for the NIs plus 4 others) will be replaced. The inspector questioned whether an evaluation had been performed in a timely manner to datermine reportability. At the exit meeting, engineering representatives acknowledged the concern and stated that the determination would be made within several day The inspector will continue to follow this issue.

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l The inspector observed portions of P0 2.02.01, Nuclear Instrumenta-

! tion System Test. During a backshift inspection, the licensee

~; encountered difficulties with the span of intermediate range monitor NI-35, prompting the use of 2T-NME-2-2.07, Initial Operation and l Alignment of Nuclear Instrumentation System Intermediate Range NI 35,

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Rev. O, to troubleshoot. A test deficiency was written because NI-35 local indication lost accuracy as the test signal was scaled upward.

l Testing then proceeded to intermediate range monitor NI-36 without further inciden '

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t ' 10 Testing for the four power range monitors, also experienced alignment problems. Consequently, the test engineer opted to have DLC I&C tech-nicians rerun the alignment procedures for each of the power range moni-tor Initial source range monitor testing had been completed from the cable connectors through the instrumentation racks. The inspector was informed that difficulties had been encountered with noise in the cable connectors and that troubleshooting was ongoing. Additionally, the section of the procedure requiring flashing of the source range monitor detector with a neutron source would not be done until closer to fuel load. Review of the post HFT schedule indicated that the Phase I excore nuclear de-tection and associated cable testing per IOP-02-01B was scheduled for March,1987, which will provide a proof-test for any corrective action neede .

No deficiencies were identifie b. RPS and ESFAS Test Procedure Review P0-2.01A.02, Reactor Protection System and Engineered Safety Features Actuation System Time Response Test, was reviewed to ensure technical adequacy and consistency with Regulatory requirements, guidance and licensee commitments. Specifically, the inspector reviewed FSAR Section 14.2.12.2.2, Preoperational Testing, FSAR Section 7.3, ESF Actuation System, and applicable functional diagrams and tables to verify that each trip function is tested by a simulated channel signal. The test calcu-lates the total response time by adding the individual sensor response time (provided by the vendor), the analog and logic circuitry delay time, the reactor trip breaker delay time and the gripper release time (pro-vided by the vendor). The RTB delay time is calculated as the time be-tween 30% voltage on the undervoltage coil and interruption of voltage to the trip breaker. This method is conservative because the shunt trip response time is substantially shorter. The test procedure also provides verification of each permissive, prohibit and bypass function along with alarm and logic function c. Solid State Protection System Test The inspector witnessed selected portions of P0 2.01A.06, Solid State Protection System Cabinet Test Panel Logic Test, during the week of December 29, 1986, to verify adherence to the procedure and applicable administrative controls. The procedure checked the self-testing equip-ment associated with the SSPS and then utilized the test panel to verify the logic and master relay actuation of the SSPS. The test was success-fully completed with the exception of one outstanding test deficiency issued to resolve a discrepancy between the procedures testing require-

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ments and the vendor technical manua Discussions with the test engi-neer indicated that after resolution, the test procedure would be sub-mitted to the JTG for final test results evaluatio No deficiencies were identifie . Fire Wrap of HVAC Duct Information Notice 83-69, Improperly Installed Fire Dampers at Nuclear Power Plants, was issued to alert licensees of potential generic problems involving the improper installation of fire dampers in ventilation ducts which penetrate fire barriers in safety-related areas. Previous review of licensee activities in this area is discussed in NRC Inspection Report 50-412/85-09. At that time, the inspector verified that the licensee was taking appropriate action by revising HVAC designs to require fire wrap at ductwork sections bounded by fire rated barriers and fire dampers. During this inspection period, the inspector noted that a significant amount of fire wrap work was in progres The inspector reviewed the design, installation, inspection, testing, and maintenance activities associated with the fire wrap wor Several HVAC drawings were reviewed to determine if ductwork sections that require fire wrapping were properly designated. The inspector verified that plant configuration corresponded to the design drawings. Several Requests for Information and Engineering and Design Coordination Reports were generated during the fire wrap installation. The inspector reviewed the dispositions by Stone & Webster SEG and noted that they were effective in resolving the discrepancie The inspector noted that H&K personnel (the fire wrap contractor) are respons-ible for marking the exterior of the wrap to identify locations of HVAC access doors installed for fire damper maintenance and surveillance. This marking is temporary and will enable re-identification late The inspector ques-tioned why DLC - SQC IP-8.3.9 did not contain an inspection point for this re-identification since there are similar inspection requirements for wrap of electrical items such as cable trays. The licensee stated that IP-8. is being revised to address this re-identificatio The inspector will review this revision and re-identification activities in a future inspectio During review of FCP-837, Release for Installation of Fire Wrap Material, the inspector noted that it did nut address control of fire w.ap rework which may be necessary after HVAC system preop testin SQC personnel stated the FCP-837 was currently being revised to address the rework issue for proper construc-tion control. This revision will also be reviewed in future inspection The inspector questioned S&W engineering personnel to determine at what point in the design and construction process the seismic calculations of HVAC and electrical supports were verified for adequacy considering the added load of the fire wrap materials. Detailed seismic calculations were performed and documented to ensure structural adequacy of supports prior to commencement of fire wrap installation for each portion of ductwor .

N 12 Periodic surveillance testing and maintenance of fire dampers will present-an additional work task that has not been experienced at Unit 1. Specifically, fire wrap " plugs" will have to be removed to uncover damper access doors and correctly re-installed afterward The inspector brought this to the atten-tion of the DLC Maintenance Director. He assured the inspector-that mainten-ance personnel will be trained how to remove and replace the fire wrap without impairing.its effectiveness._ In future inspections, the inspector will ob -

serve routine surveillance and maintenance practices involving fire wra No violations were. identifie . Radiation Monitoring System Review A review of the Digital Radiation Monitoring System (DRMS) Specification and

, Operating Manual Chapter was conducted. These documents detail the system monitors for the expected radiation fields in the various process and effluent l streams. The-inspector determined that the detection capabilities and ranges are appropriate for the respectiva stream The inspector also reviewed the Steam Generator Blowdown Sample Radiation Monitor Test procedure, transfer calibration and associated test data infor-mation from the test performed in November 198 The inspector noted that the licensee encountered some difficulties with establishing liquid flow and conducting purge tests. -This item was tracked and subsequently cleare '

The inspector reviewed the licensee's information pertaining to isokinetic nozzle sizing for the DRMS for compliance with ANSI-N-13.1, 1969, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities" as specified-by BVPS 2 FSAR Section 11.5.2.3.5. The calculations for nozzle sizing were performed for the licensee by Stone & Webster Engineering. The inspector identified a math error and inconsistencies in the pressure and temperature duct flow corrections and subsequent nozzle size calculations. The inspector discussed these concerns with-the licensee. Subsequently, the licensee stated that the document was not' appropriate for the General Atomics ~DRMS System which is currently' installed. The inspector was advised that documentation would be supplied by GA for isokinetic nozzle sizing in the installed DRMS System. This area will be reviewed in future inspections, i No violations were identified.

1 QA Program Weaknesses Associated with Cable Storage During a routine site tour, an NRC specialist inspector identified deficien-i cies in the storage of safety-related electrical cabl The detailed findings i are presented in Inspection Report No. 50-412/86-47. In summary, the inspec-tor found many instances where cable reels were stored in outdoor areas with-

! out the cable ends sealed as required by the licensee's engineering specifi-

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cation. Followup inspection revealed the following weaknesses in the area i of QA: the QC inspection checklist implementing the engineering specification I

omitted r.he cable end seal requirement, the QC inspection program was inade-

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quate in that it would require several years to complete one inspection cycle, other instances of identical cable storage deficiencies at other locations on site were identified by QC inspectors without root cause followup, the licensee's trending program did not capture the repeated identified deficien-cies, and the QA audits did not confirm the technical adequacy of the QC checklist Another deficiency involves the apparent weakness in QA oversight in that the above program deficiencies were not self-identified and corrected; this item is similar to the QA weakness discussed in section 9. Some of the licensee's initial corrective actions were reviewed c'uring a recent NRC team inspection (50-412/87-03).

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

Licensee commitments in response to TMI Action Plan requirements have been reviewed by the staff and are documented in the BVPS Unit 2 Safety Evaluation Report (SER) and its supplements. Several of the TMI items are still under review and items may require further attention if significant changes become necessar During this inspection period, the inspectors verified the licen-see's compliance with the following items: Item II.D.3 Direct Indication of Relief and Safety Valve Position Reactor coolant system power operated relief valves (PORVs) and safety relief valves shall be provided with a positive position indication in the control roo This indication must be derived from a reliable valve position detection device or a reliable indication of flow in the dis-charge pipin BVPS 2 has three pressurizer code safety valve Each valve has a Minco Platinum RTD located in the valve discharge piping before the line joins a common relief valve header to the pressurizer relief tank (PRT). The temperature indicators will initiate alarm signals for both the control room annunciators and plant computer if discharge piping temperature exceeds 20 degrees F above ambient temperatur Each individual tempera-ture indicator has its own computer printout, allowing for immediate identification of the lifted valv However, the annunciator is shared between the safety valves and PORV Three pressurizer solenoid-operated PORVs are installed in parallel off a common line from the pressurizer and discharge to a header which is shared with the safety valves to the PRT. A single Minco Platinum RTD in the PORV's discharge header provides temperature indication for PORV position verificatio As with the safety valves, a 20 degree F rise in discharge piping temperature will initiate an alarm signal for the control room annunciator and plant computer. Because there is only one RTD, the computer printout will not distinguish which PORV has lifte .

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Individual PORV position is identified on the main control board by valve position indicating lights which receive position indication from reed switches on the valve stems. There is also a position indicating light for PORV 2RCS*PCV-456 on the Alternate Shutdown Pane The inspector reviewed the applicable procedures and system descriptions contained in OM Chapter 6 and BVPS 2 FSAR Chapters 5 and Also, the inspector witnessed portions of PORV preop testing in accordance with P0 2.06.06, Testing of PRT, PORVs and Temperature Alarms. During per-formance of this preop test, the licensee experienced minor setbacks with inadequate heat tracing between the block valves and PORVs, and two PORVs which would not stroke during the tests at low primary pressures. Re-solution of these test deficiencies is being followed by the inspector . BVPS 2 SER 7.5.2.3 describes the NRC review and approval of the PORV and safety valves position indication design as identified by BVPS 2 FSAR Amendment Based on these reviews, this item is close Item II.G.1 Emergency Power for Pressurizer Equipment The motive and control power connections to the PORVs block valves and pressurizer level indicator instrument channels shall be capable of being powered from either normal offsite power sources or the emergency power sources. Also, the power connections to the emergency buses shall be through devices that have been qualified in accordance with safety grade requirement Each PORV's block valve motive power is supplied from 480 V AC Class 1E circuits, the PORVs receive their power from 125 V DC Class 1E circuits as do the pressurizer level indication channel This pressurizer equipment is qualified to Class 1E standards, therefore, on loss of off-site power, th b equipment's supply will automatically switch over to onsite emergency powe The inspector reviewed procedures and system descriptions relevant to pressurizer equipment power supplies contained in OM Chapter 6 and BVPS 2 FSAR Chapter 8, and the inspectors observed portion of preop test ~P0 2.06.07, Pressurizer Pressure and Level Test. BVPS 2 SER 8.3.3.4 de-scribes an additional review of this area and concludes that the BVPS 2 design is adequate. Based on these reviews, this item is close . Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report perio _ _ _ ______ _ _ _ _ _ _ _ _ _ _ _ _ a