IR 05000412/1987002

Resident Insp Rept 50-412/87-02 on 870117-0228.Violation Noted:Inadequate QC Insp Plan for Insps of HVAC Equipment Seismic Supports & Subsequent Bolting Deficiencies in Such Supports.Unresolved Item Also Opened Re Two PORVs

ML20207S261
Person / Time
Site:	Beaver Valley
Issue date:	03/09/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207S247	List: IR 05000412/1987002 ML20207S246 ML20207S249
References
TASK-1.A.1.1, TASK-1.A.1.2, TASK-1.A.1.3, TASK-1.C.2, TASK-1.C.3, TASK-1.C.4, TASK-2.K.2.13, TASK-2.K.3.05, TASK-2.K.3.17, TASK-2.K.3.31, TASK-TM 50-412-87-02, 50-412-87-2, IEB-79-01, IEB-79-1, IEB-85-001, IEB-85-1, NUDOCS 8703190208
Download: ML20207S261 (16)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-02 Docket N License N CPPR-1 Licensee: Duquesne Light Company Nuclear Construction Division P. O. Box 328 Shippingport, PA 15077 Facility Name: Beaver Valley Power Station, Unit 2 Dates: January 17 to February 28, 1987 Inspectors: J. E. Beall, Senior Resident Inspector L. J. Prividy, Resident Inspector A. A. Asars, Resident Inspector W. M. Troskoski, Senior Resident Inspector, BVPS Unit 1 C. H. Wo dard, Reactor Engineer, DRS Approved'by: b. M F 3/9/87 E. E. Trfpp, Chief, Reactor Projects Section 3B ' Ddte Inspection Summary: Inspection No. 50-412/87-02 on January 17 - February 28, 1987 Areas Inspected: Routine inspections by the resident inspectors (461 hours0.00534 days <br />0.128 hours <br />7.622354e-4 weeks <br />1.754105e-4 months <br />) of licensee actions on previous findings, site activities, preoperational test program implementation, EDG sequential start testing, control room wall removal status, TMI Action Plan Requirements, charging pump test deficiencies, Reactor Trip System, reactor vessel pressure transient protection, CO Actuation 2 of Fire Dampers, HVAC equipment seismic supports and containment integrit Results: One violation was identified involving an inadequate QC inspection plan

, for inspections of HVAC equipment seismic supports and subsequent bolting defi-ciencies in such supports (see detail 11) and one unresolved item was opened to track demonstration of correct low pressure operation of two PORVs which failed during preop testing (see detail 9.c). Important preoperational tests conducted during this time included Containment Structural Acceptance Test (SAT) and Inte-grated Leak Rate Test (ILRT) (see detail 12) as well as MSIV testing (detail 5.d),

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i 8703190208 870312 PDR ADOCK 05000412 G PDR ,

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DETAILS 1. 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 inspectior, activitie . Project Status Summary Construction activities are currently estimated to be in excess of 98% com-plete, with 466 of 476 subsystems turned over for flushing and proof-testin For software, about 104 out of 114 preoperational (P0) and initial startup tests (IST) have been approved. The remainder are in various phases of de-velopmen Approximate dates for the major project milestones, as currently estimated by the licensee are as follows:

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Loss of Power Test (Site Blackout Portion) April 2, 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 3. 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 45 40 Procedure Reviews:

Mandatory 70 50 Primal 100 100

Test Witness:

Mandatory 50 35 Primal 100 100 Results Review:

Mandatory 15 15 Primal 5 5 This inspection status is consistent with the applicant's test program pro-gress. At the end of this inspection period, there were approximately 56 open NRC inspection items as listed below:

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NO. OF OPEN INSPECTION ITEMS TYPE OF ITEM END OF THIS PERIOD END OF LAST PERIOD Bulletins 5 6 Violations 6 5 Deviations 0 1 Construction Deficiency Reports 16 17 Unresolved and Inspector Follow 29 30 TOTAL 56 59 4. Licensee Actions on Previous Inspection Findings (Closed) Unresolved Item (86-01-03): Preoperational program appears to lack double verification requirements for correct restoration of mechanical com-ponents. This item was last discussed in Detail 12 of Inspection Report 412/86-3 The administrative' requirements of SUM 7.1, Systems / Subsystem Turnover Program and related subsections adequately addressed the initial inspector concerns. Additionally, station experience through the hot func-tional testing phase has been satisfactory in this area. This item is close (Closed) Unresolved Item (86-07-01): Re-establish RCS cleanliness conditions and strengthen administrative controls. This item was opened to follow prim-ary system cleanup activities after metal shavings were introduced into the system when a steam generator nozzle dam had dropped into the C reactor cool-ant pump suction line. After completion of the RCS cold hydrostatic test, this item received additional attention in Inspection Report 412/86-11 (de-tail 3.3), and Inspection Report 412/86-12 (detail 8). During this inspection period, continued implementation of those controls were observed in the field during the full flow safety injection test and the pressurizer PORV mainten-ance. Licensee action remains satisfactory and this item is close (Closed) IFI (86-01-02): Strengthen Temporary Operating Procedure (T0P) use by editing OM Chapter procedures for incorporation, or pre-mark referenced procedures and distribute as a package. Inspector review of working TOPS in the Control Room files indicated that this concern has been adequately ad-dressed by the station. The result has been the development of TOPS that are adequately structured for human factors use. This item is close (Closed) Bulletin (79-8U-01): Equipment qualification of class 1E equipmen This bulletin did not specifically request a response from BVPS 2 because of its construction status. However, after NRC reviewed the responses from lic-ensees required to do so, the need arose for a team inspection of licensee

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EQ programs. In November, 1986, NRC conducted a preliminary check of the EQ program at Unit 2. No major deficiencies were identified. The actual EQ team inspection will take place after fuel load. This item is close (Closed) Unresolved Item (86-31-03): Potential for High Energy Line Break in the Safeguards Building. This item involved the potential break of the steam line to the steam driven auxiliary feed pump (TAFP) due to the introduction of a water slug into the line during a steam generator overfill scenari A licensee study evaluated the effects of a steam generator overfill, includ-ing the introduction of water into the steam line to the TAFP. Calculation showed that the piping and hanger loads were acceptable, but did not com-pletely resolve the effects on the steam inlet nozzle to the TAFP. Subsequent calculations showed that the nozzle loads were also acceptable, thus support-ing the conclusion that a steam generator overfill would not cause a high energy line break in the Safeguards Building. This item is close (Closed) Unresolved Item (85-16-02): Engineering evaluation of the impact of hot piping on cable in close proximity. Stone & Webster Engineering has com-pleted the preliminary evaluation of potential accelerated thermal aging of cable from nearby hot piping. This evaluation confirmed that this is a potential construction deficiency and reportable per 10 CFR 50.55(e). On January 12, 1987, the licensee reported this issue as CDR (87-00-01). The unresolved item is being closed and resolution of this issue will be tracked by the CDR numbe (Closed) Construction Deficiency Report (86-00-10): Fire Dampers. During HVAC System testing, two horizontally-mounted fire dampers did not close fully after the fusible link had been removed. Missing springs on the bottom blade (s) were identified as the cause for the incomplete closure. Further investigation revealed that the potential for this problem existed on 22 safety related, horizontally-mounted fire dampers. In accordance with the requirements of 10 CFR 50.55(e), the licensee filed a Construction Deficiency Report on August 8, 198 The main corrective action was the addition of field-installed springs to the damper bottom blades in accordance with a design change provided by the vendor (American Warming and Ventilating, Inc). This design change and subsequent satisfactory retest resolved the nonconformance and disposition report that was issued for the defective fire dampers. The vendor provided an installa-tion procedure and a seismic qualification report for the field-installed spring The inspector determined that the licensee's engineering, construction, in-spection and retest efforts pertinent to this design change were satisfactor Based on this review, this item is closed.

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(Closed) Deviation (85-25-02): Crane Monorails - Safety /Non-Seismic Interac-tion. This item addresses several concerns with the crane monorails that are installed above various safety-related equipment. Specifically, questionable weld quality existed in load carrying members of the monorail for the C charging pump. Also, the general project program describing interaction be-tween nonseismic and safety-related equipment was not well define The inspector reviewed the implementation of the licensee's corrective actions as defined in their initial and supplemental responses dated January 16 and February 13, 1986, respectively. The inspector met with varicus project per-sonnel and reviewed the repairs and resolutions of the deficiencies that were identified for the crane monorails. These repairs adequately resolved the prior monorail deficiencies. Also, the inspector noted that the project pro-cedure, 2BVM-165 " Hazards Analysis Programs" was revised to document the safety /nonseismic interaction assessment completed for each monorail located over safety related equipmen This item is close (Closed) Construction Deficiency Report (86-00-08): 21C Reactor Coolant Pump damage and cleanliness control. This item is being closed based upon the in-spection conducted for the closure of Unresolved Item (86-07-01) mentioned abov (Closed) IE Bulletin (85-BU-01): Steam Binding of Auxiliary Feedwater Pump This bulletin requested licensees to take actions to prevent steam binding of auxiliary feedwater (AFW) pumps including detection of elevated piping temperatures, development of procedures for restoring the AFW to operable status upon detection of steam binding, and completion of hardware modifica-tions where warranted to reduce the likelihood of steam bindin The inspector's review determined that BVPS-2 operating logs (0M Chapter 54, L11-11) require a " hands on" check of AFW discharge piping each shift when AFW system operability is required. The licensee has developed procedures (0M Chapter 24.4) to restore AFW operability upon detection of elevated feed-water discharge temperatures. The inspector reviewed the procedures for technical adequacy, responsiveness to the concerns of IE Bulletin 85-01, and guidance for actions to take upon detection of multi-train steam bindin No deficiencies were identifie Certain hardware features were identified which also reduce the likelihood of steam binding the AFW pump Each AFW pump has an automatic recirculation control valve which functions as a spring-loaded check valve when the pump is not operatin This valve is in addition to two swing check valves in each discharge heade Each discharge header contains a relatively long (licensee calculated average is 147 feet) run of uninsulated pipe which would facilitate heat dissipation and help mitigate the effects of potential backleakage. This bulletin is close (Closed) Unresolved Item (86-31-02): Test Procedure P0-2.01A.01 did not pro-vide for testing the " General Warning Alarm Reactor Trip" under the conditions when both trains of the reactor trip system are made inoperable by the auto-

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matic opening of all reactor trip and bypass trip breakers. This trip is in-cluded in the " General Warning Alarm Reactor Trip" alarm conditions in Section 7.2.2.2.3 of the BVPS 2 FSAR and is also discussed in Section 7.2.2 of NUREG-1057, SER for BVPS 2, Supplement 1. The inspector confirmed by a procedure review that the licensee has implemented Test Change Number 2 to Test Proce-dure P0-2.01A.02 effective November 6, 1986, to include this required testin This item is close (Closed) Violation (86-31-04): Rigid Sway Strut Functional Interference. This item originated in 1984 (Inspection Report 50-412/84-18) with the identifica-tion of potential interferences in certain strut paddles, brackets and swa strut Tight clearance between strut paddle and bracket had resulted in some rigid sway struts not being able to achieve the designed 10 degree lateral movemen Later NRC Inspection Reports (50-412/85-09 and 86-07) provided updates on this ite Upon completion of licensee corrective action, the inspector identified additional deficiencies in the field which led to the subject Violatio The root cause of the Violation was inadequate QC acceptance criteria provided to the QC inspectors for use in making the evaluation for acceptability. The licensee's corrective actions included revising the QC acceptance criteria and reinspecting all Category I rigid sway struts. Of the 426 supports re-inspected, 34 deficient supports were identified. An additional 5 supports are not readily accessible but will be reinspected prior to fuel loa The identified hardware deficiencies and the remaining 5 supports have been entered into the licensee's deficiency tracking system for closur This Violation is closed; the completion of the licensee's resolution of de-ficiencies will be reviewed in a future inspection. The licensee's corrective actions effectively addressed the individual deficiencies identified by the NRC in hardware and criteria. However, similar concerns relating to accept-ance criteria provided to QC inspectors are discussed in detail 1 . 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 activities. This included the following: The Unit 2 switchboard operator received a bomb threat on February 6, 1987, at 10:12 a.m. The male caller stated that there were three bombs on site set to detonate at 3:45 p.m. He did not specify where the bombs were located. The licensee initiated a search of the entire Unit 2 site utilizing both Unit 1 and Unit 2 security personnel. By 2:00 p.m. the search was complete without discovery of any explosive devices. An announcement was made at 2:30 p.m. advising plant personnel that a threat had been received but no bombs were found. At this time, personnel were given permission to leave the site early, if they wishe . .

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This is the fourth bomb threat received at BVPS 2. The three previous threats were received in September of 1986, coincident with fuel receipt and are discussed in NRC Inspection Report 412/86-2 The licensee projects a scheduled date of April 1,1987, for the removal of the steel wall that currently separates the Unit 1 and 2 Control Room This item has been discussed in Unit 1 Inspection Report No. 50-334/87-0 During this inspection period, the inspector attended the weekly meetings which are being conducted to coordinate all site activities associated with the wall removal. It appears that the licensee is properly co-ordinating the work and addressing NRC questions that have arisen in response to Unit 1 Technical Specification change request A recent NRC inspection (50-412/87-05) in the area of security determined that the required hardware and procedures were not yet in place to sup-port the expansion of the Unit 1 protected area to include Unit 2. The implementation of full site security is a necessary prerequisite for the issuance of a license to load fuel at Unit 2. A meeting was held onsite on February 18, 1987, with licensee management to discuss the status of security program development. The meeting was attended by the Unit 1 and Unit 2 senior resident inspectors, a physical security specialist inspector, the NMSS security reviewer, the cognizant Region I Projects Branch Chief and the Radiation Safety and Safeguards Division Directo The inspectors are following the licensee's actions on this matte During this inspection period, the licensee reinstalled and cycled the MSIV Leak rates were found to be acceptable with no adverse trends identified after cyclin The A and 8 MSIVs closure times were less than the 5 seconds required by technical specification. However, the C MSIV hydraulics required some additional rework and testing. This testing was temporarily suspended during February for containment testing. The inspectors are periodically monitoring the status of the valve Sequential auto starts on the No. 2-1 Emergency Diesel Generator were conducted during the inspection period. The No. 2-2 EDG will be tested during the next inspection period. The inspectors witnessed several of the auto starts and monitored critical parameters during EDG operatio The air start system for No. 2-1 EDG had required additional flushing during the beginning of the auto starts. After flushing, all starts were successfully completed. The inspectors will follow future EDG testin No violations or unacceptable conditions 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

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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: (Closed) Item I.A.1.1, Shift Technical Advisor (STA)

Each licensee must provide an on-shift STA to advise the shift superviso The STA's must possess a bachelor's degree in engineering or a related discipline and have received training in response to and analysis of plant transients and accidents. Implicit in fulfilling these duties is the necessity for plant systems trainin The proof and review draft of BVPS 2 Technical Specifications (TS) states that during Modes 1 through 4, a STA must be on shift as part of the minimum shift composition (Table 6.2.3). This table will be footnoted for both units' TS to clarify that the STA may be shared between units if qualified to do s Currently, four of the eight STAS that are stationed at Unit 1 are under-going Unit 2 cross-training. The others will attend classes in Marc This program includes three weeks of classes designed to cover the Unit 2 systems with emphasis on the differences between the units and a loca-tion check off sheet similar to that utilized in operator trainin An additional training session for the STA role in the Appendix R Safe Shutdown Procedure will be scheduled after the procedure is issued. The inspector also questioned what training would be provided for new STAS in the event that additional personnel are hired. In this case, each new STA will be required to complete the Unit 1 STA Training Program and the Unit 2 Cross-Training Program before assuming watch on shif The inspector attended portions of the STA cross-training sessions and discussed the lesson content and adequacy with several STAS. No defi-ciencies were identified; this item is close (Closed) Item I.A.1.2, Shift Supervisor Administrative Duties The objective of this item is to decrease the shift supervisors admini-strative responsibilities which could distrc~t him from plant operatio Operating Manual Chapter 48, Conduct of Operations, has been issued as a shared chapter for both Unit 1 and Unit The shift supervisors'

responsibilities are outlined in Section 1, Procedure A, on duties and responsibilities of the Operations Group. The inspector reviewed the duties and confirmed that each is directly related to plant operations and should not cause any unnecessary distraction This item is close (Closed) Item I.A.1.3, Shift Manning This item defines the shift manning required for normal operations and overtime restrictions. Table 6.2-1 of the proof and review TS defines the minimum shift crew composi. tion. This table provides for two each

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SR0's, R0's, and auxiliary operators and one STA with the provision that an STA can fill the position of STA for both units if qualified on both units. Section 6 of technical specifications and OM Chapter 48, Section 1, Procedure 8, details the limits on personnel overtime. The inspector noted that an individual is permitted to work up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight and an eight hour break is required between work periods. This deviates slightly from the figures provided in NUREG-0737 and is due to previous labor agreements. Unit 1 carries the same limits on personnel overtime and in practice has met the intent of overtime restriction Based on this review and current overtime practices at Unit 1, this item is close d. (Closed) Item I.C.2, Shift Relief and Turnover Procedures The objective of shift relief and turnover is to transfer the knowledge of critical plant status and availability from the off going shift to the on-coming shift. OM Chapter 48, Section 1, Procedure C, details the instructions for a comprehensive exchange of pertinent information be-tween shifts. The inspector reviewed the instructions and shift turnover checklists and found them adequate. Based on this review and the fact that personnel currently eligible to receive a license to operate Unit 2 have operated at Unit 1 and performed numerous adequate shift turnovers, this item is close e. (Closed) Item I.C.3, Shift Supervisor Responsibilities This item requires that licensees establish a definite line of command and clear delineation of authority in the Control Room and Operations Group. The inspector verified that OM Chapter 48 adequately describes the Operations Group organization and chain of command both during normal operations and in emergency situations. Also included are detailed de-scriptions of the responsibilities and duties of each member o' the Con-trol Room staff. Based on this review, this item is close f. (Closed) Item I.C.4, Control Room Access Licensees are required to charge an individual with the responsibility and authority to control access to the Control Room in the event of an emergency. OM Chapter 48 tasks the shift supervisors with eliminating non-essential personnel from the Control Room both during normal opera-tion and in emergency situations. The shift supervisor may delegate the administrative assistant to control access, but the shift supervisor is ultimately responsible. At Unit 1, the shift supervisor has been ob-served to exercise this authority when necessary. Based on this review, this item is close g. (Closed) Item II.K.2.13, Thermal Mechanical Report - Pressurized Thermal Shock (PTS)

This item requires a detailed analysis of the thermal - mechanical con-ditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. Of particular concern is the condition

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of severe overcooling of the vessel concurrent with or followed by re-pressurization, commonly called PTS. Unresolved Safety Issue 49 is cur-rently researching this issue. BVPS 2 SER concluded, in Appendix C, that there is reasonable assurance that the unit can be operated safely with-out undue risk to the public before final resolution of USI 49. Based on this review, this item is close Item II.K.3, Final Recommendations of B&O Task Force This item is divided into several subsections; the-following were re-viewed during this inspection period:

(1) (0 pen) II.K.3.5, Automatic Trip of Reactor Coolant Pumps during LOCA This item required Reactor Coolant Pumps (RCPs) to automatically trip in the case of a small break LOCA. The BVPS 2 SER states that the Westinghouse generic reports which propose a delayed RCP trip are currently under review. Any necessary actions resulting from this review will be addressed at that tim . (Closed) II.K.3.17, Report on Outages of ECCS Applicants for operating licenses were required to develop methods for compilation of availability statistics for ECCS component The data should include information on equipment outage date and duration, the cause of the outage, and the corrective actions take BVPS 2 plans to utilize a system identical to that of Unit Per-tinent information is generated by the Maintenance Work Request (MWR) and entered into the MWR tracking program in accordance with Station Administrative Procedure 3D. The SAP adequately details the information required on the MWR. Based on the review of main-tenance activities with respect to the MWR and the issuance of a SAP 3D applicable to both units; this item is close . (Closed) II.K.3.31, Plant Specific Calculations to Show Compliance with 10 CFR 50.46 This item required licensees to submit for review a new small-break LOCA analysis which is plant specific to show compliance with 10 CFR 50.4 The BVPS 2 SER documents the review of this submittal and the conclusion that the BVPS 2 FSAR analysis is adequate in determining compliance with the ECCS acceptance criteria. Based on this review, this item is close . Full Flow SI - Charging Pump Retest Retest of the C high head safety injection / charging pump was witnessed by the inspector on February 19, 1987. TOP 87-18, CHS Pump Startup for SIS Full Flow, provided various flow rates through the normal high head hot leg injection path while data was collected using both in place flow transmitters and strap-on flow transducer Both methods agreed to within about 1%.

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The reason the CHS pumps were retested is that they failed to meet the design minimum operating curves during performance of P0-2.11A.03. The A and B pumps performed acceptably for their ECCS function; but the C pump had a lower than minimum pump head at runout. TOP 87-18 was written to minimize potential leak paths during a retest by providing double isolation.- Test results agreed with the previous data, indicating the need to rework and retest CHS-P-21 At the conclusion of this inspection period, the station had elected to install a spare rotating assembly and perform new flow test The original pump assembly is to be shipped to Pacific, the pump vendor, for further testing at their laborator Station actions will be monitored in future inspection No violations were identifie . Reactor Trip System Test Witnessing On February 7,1987, the inspector witnessed the retest of P0-2.01A.01, Reactor Trip Breakers, Section A, which was rerun to correct prior test deficiencies. Both bypass trip breakers tested satisfactorily. A minor mechanical problem developed when the startup operator attempted to rack the A reactor trip breaker onto the bus, as it appeared to jam. An SWR was initiated and DLC electrical personnel subsequently determined that the cause was the relatively loose tolerances between the rollers and the guide bar. After physically realigning the breaker, it was success-fully racked onto the bus and subsequently tested without further even During observation of testing, the inspector noted that DC power to the breaker control was removed by pulling fuses, as there apparently is no DC control switch. Additionally, a walkdown of the Unit 2 control board indicated the absence of remote reactor trip breaker position indicating lights (which have already been installed at Unit 1). Both items were brought to the Station Superintendent's attention. The inspector was informed that the. station would review these concerns and develop a formal position. Licensee actions will be reviewed in future inspection During operation of the reactor trip breakers, (Westinghouse DS 416),

the inspector noted an unusual amount of vibration during recharging of the trip springs. A concern was identified regarding the effects of vibration as an aging factor on electronic components, particularly the UV and shunt coils. The test engineers initiated a request for informa-tion to Engineering. Westinghouse representatives witnessed several breaker cycles and concluded that the components were functioning pro-perly according to design as the charging springs are used infrequently during the lifetime of the plant, the inspector had no further questions at this tim . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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12 Test Results The inspector reviewed the test results of P0-2.01A.01, Reactor Trip Switchgear and Control Rod Drive MG Power Supply Test, Issue 1 and Test data compared favorably with acceptance criteria, vendor documents and FSAR commitments. A total of 23 test deficiencies were identified during testing per Issue 1 and 6 deficiencies during testing per Issue 2. The inspector reviewed the resolution of each as specified in the applicable test deficiency form. All were satisfactorily dispositioned and documented. No safety concerns were identifie No violations were identifie . Reactor Vessel Pressure Transient Protection

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The inspector walked down portions of the low temperature overpressure pro-tection system (0PPS) and reviewed test data, OM procedures and prints to verify that the system had been designed, constructed and tested to the sta-tion's commitments contained in Section 5.2.2 of the FSA Design The OPPS was designed to Category I Seismic criteria as indicated by the NSS Category I seismic line breaks on OM Figure 6-2A and FSAR Section 5.11 references to the seismic design codes employed on the pressurizer relief pipin Redundant protection against a single failure is provided by two 100% subsystems consisting of: a block valve, PORV, instrumenta-tion channels and electrical supplies for each train. Each instrumenta-tion channel and electrical supply is independent of the other. The system is designed to assure that there is no initiating event that could also fail the equipment needed for accident mitigation. Individual trains are " armed" by a switch on the main control board, after which no further operator action is required to prevent an overpressure transi-en This is the standard Westinghouse desig The setpoints used by the two trains are offset to avoid the unnecessary simultaneous lifting of both PORVs. These setpoints are electronically l calculated by comparing an auctioneered RCS T-cold signal with wide range

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pressure. This function, selected to protect against exceeding the 10 CFR 50, Appendix G, limits, is specific to the BV-2 reactor vessel and weld material composition. The low end of this setpoint is 449 psig at 70 Administrative Controls and Procedures l

Through a review of OM 2.6.4E, Filling and Venting the RCS, Issue 1, Re-vision 1, the inspector verified that the station has developed specific l requirements to minimize the time spent in an RCS water solid condition.

! Additionally, an operator is continuously assigned to the section of the control board with instructions to shutdown any operating charging pump, I

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1 should RCS letdown be isolated for any reason while the plant is solid and the pressurizer is not vented to the atmosphere. This protects against the loss of the Number 2 vital bus which would isolate letdown and remove one of the two PORV trains from service, thereby reducing system reliabilit Control Room annunciators have been installed and tested to warn opera-tors of the approach to an overpressure condition (setpoint is 350 psig)

to allow time for manual actio Procedures for placing and removing the OPPS from service have been in-corporated into the preliminary OM Startup Procedures. Various technical specification surveillance requirements referred to in SER, Section 5.2.2.2, OPPS During Low-Temperature Operation, are in the final stages of revie Surveillance Testing P0-2.06.06, Testing of Pressurizer Relief Tank, PORVs, and Alarms, was reviewed by the inspector to verify that logic for the PORVs and block valves were satisfactorily checked and cold OPPS actuation demonstrate The PORVs used at Unit 2 are pilot operated Crosby (Garrett was original vendor) valves, electrically controlled by S0Vs. During HFT, these PORVs failed to stroke at low pressures of 80 - 200 psig. As these valves were successfully stroked at a normal operating pressures (2235 psig), the station indicated that the lower pressures were thought to be below the level necessary for pilot operatio The inspector, along with a licensee and vendor representative, observed the disassembly of RCS-PCV-455 C and no significant visual damage was apparent, though some pitting was noted on the stellite seat. The sta-tion removed 0.003 inches, which is below the ASME Section XI specifica-tion of a repair, and reinstalled the valv RCS cleanliness conditions were maintained and QC provided the necessary coverag Subsequent discussions with the Unit 2 Station Superintendent indicated that DLC planned to retest the PORVs at closer to their expected operat-ing pressure (400 psig) after fuel load by the use of reactor coolant pump hea Demonstration of correct low pressure operation of RCS-PCV-455D and 456 will be followed as Unresolved Item (87-02-01). Test Results Evaluation:

During performance of P0-2.06.06, two other deficiencies were identifie Reactor Coolant safety valve 2RC-RV551B leaked during the HFT, as indi-cated by elevated tail pipe temperatures (test deficiency number 18).

Discussions with DLC Maintenance Engineers indicated that the valve was sent offsite to Crosby for failure cause determination and valve repai As no spare is available, the station is currently awaiting the valve's return for reinstallation.

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A second problem concerned electrical wires to the 50V on RCS-PCV-45 Contractor personnel informed the inspector that the wires appeared to have been pulled out of the 50V while energized. An N&DR was issued and it was determined that the terminal block would need replacement. No generic problem due to a malfunction in the 50V was apparen No violations were identifie . CO, Actuation of Fire Dampers During an inspection of fire dampers, the inspector identified that the copper CO 2 actuation line for fire damper 2HVR*DMPF-205A was severely crimped at a bend of the copper tubing between the inlet to the damper sleeve and the CO 2 actuator. The inspector brought this deficiency to the attention of DLC -

SVG personnel who were responsible for the proof-testing of all fire damper Additional inspections were made at other CO 2 actuated fire dampers to deter-mine if this problem was widespread. No further examples of crimped copper tubing were identified. The inspector reviewed test report number M 1966 dated September 13, 1986, which documented the satisfactory performance of 2HVR*DMPF-205A in accordance with Generic Test Procedure, MTP-510, Rev. 3

" Initial Operating Procedure for Fire Dampers." This test demonstrated that the fire damper is operational with the copper tubing crimpe It appears that the above condition is an isolated problem and can be cor-rected by normal site procedures. Accordingly, DLC - SQC has issued N&D 34,499 on 2HVR*DMPF205A for an engineering resolution to the copper tubing noncon-formance. The inspector had no further comments or concerns related to this matte No violations were identifie . HVAC Equipment Seismic Supports During a routine inspection, the inspector identified that both Diesel Genera-tor Building Secondary Supply Air Fans did not have the correct number of bolts attaching them to their seismic structural steel supports. Of the three 1/2 inch A325 bolts required by the controlling design and construction draw-ing for each of four supports, only two had been installed. The supports had been inspected and accepted by QC inspectors following approved QA procedure The two fans (2HVD*FN271A and B) are both safety related components and each is supported by two seismic Category I structural steel supports. Each sup-port is required to have three bolts by the controlling construction drawing (SM-7) via reference to the applicable structural drawing (RS-29C). After the inspector identified this item, the licensee reported the bolts' omission under 10 CFR 50.55(e).

The QC inspection was performed under Site Quality Control Inspection Plan IP-7.4.1, " Inspection of QA Category I and other Seismic HVAC System I In-sta11ations." The individual inspections (ME-MI-5060-1 and 2 for 2HVD*FN271A

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and ME-MI-5059-1 and 2 for 2HVD*FN271B) were performed on August 21, 1985,

)~ 'and in each document, the inspection attribute verifying the acceptability of structural steel supports (ME-MI-027) was checked " Satisfactory."

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Discussions with licensee personnel indicated that QC personnel verifying the

, attribute had checked the HVAC equipment only for location and installation i with respect to the duct run. Inspection of the installation and bolting of i the equipment to the supports was not performed. Each fan's supports were

! checked by a different QC inspector indicating that the inspection procedur did not give adequate guidance to the inspectors. This is an apparent Viola-tion (87-02-02).

The identification of additional deficiencies in other HVAC Category I equip-

! ment supports during subsequent licensee reinspections is further indication that the root cause of this apparent violation is inadequate QC guidance.

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This is similar to Violation 86-31-04 involving inadequate procedures for in-specting rigid sway strut functional interference and also to Violation 86-47-02 involving inadequate procedures for inspecting stored cable. In all

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three cases, inadequate QC inspection guidance resulted in hardware deficien-cies being accepted by the licensee's QA/QC program and remaining undiscovered

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until identified by NRC inspectors in the field.

[ Licensee corrective actions in response to the individual violations have been limited to reinspection of similar potentially defective components and cor-rection of the NRC identified deficient procedures. The common root cause of the violations is an apparent weakness in QA oversight in that inadequate QC inspection procedures are not self-identified and corrected (see also, 2 Inspection Report 50-412/86-44, section 14). Licensee corrective actions i should address the review of QC inspection guidance for technical adequacy.

1 Containment Testing During the week of February 8,1987, the licensee conducted the Containment Structural Acceptance Test (SAT) and Integrated Leak Rate Test (ILRT). These were performed to verify the structural integrity of containment and demon- >

strate that it can withstand pressures postulated to be present during a de-sign basis accident without exceeding an allowable leak rate. The inspectors observed licensee preparation and conduct of the tests. This included a final walkdown of containment before close out, frequent tours of crack wrapping areas, and monitoring pressure and leak rate during ILRT. Region based in-l spectors were also on site for the duration of the tests (see NRC Inspection Report 50-412/87-12).

After completion of containment pressure testing the licensee conducted con-tainment vacuum system testing per 50V-2.12A.01. While containment was under a vacuum, a Hydrogen Recombiner was operated to verify operability under post-accident conditions. Recombiner 2HCS*C21B did not produce any flow into con-tainment at 9 psia. However, flow sufficient to lift the inlet check valve was achieved at about 14 psia. Only the "B" blower was run. Currently, Stone and Webster Engineering is evaluating the test data and will determine if a

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design change is necessar The inspectors noted that this situation appears to be similar to a recombiner problem at Unit 1 which is discussed in NRC Inspection Report 50-334/86-18, detail 10. The inspectors will follow licen-see action in this area in future inspection No violations were identifie . Electrical Separation of Cable The inspector reviewed the status of the licensee's program to achieve elec-trical separation of cables as specified in FSAR Section 8.3.1.4. The licen-see's previous practice was to issue a quarterly report internally which docu-mented commodity quantities, responsibilities and schedules. The final such report was issued on November 7, 1986, and was reviewed by an NRC specialist inspector during part of his onsite inspection as documented in Inspection Report 50-412/86-4 Program status updates shifted to a less formal but more frequent basis to facilitate tracking as the project neared completion. At the end of this inspection period, substantial units of key commodities re-

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mained to complete the licensee's scope of work needed to accomplish cable separation. Of the origind scope of work, approximately 51% of the cable wraps (1491 of 2902) and 53% of the cable tray covers (18,999 of 35,989 linear feet) remain to be completed. Other activities such as completing termina-tions, installing cable tray clips and supports, and addressing unsupported cable are also in progress and represent over 20,000 man-hours of effort according to licensee projection The licensee has increased the site work force to about 600 including over 1,100 electricians. The quantities of work completed reflect the higher staffing with the completion of 298 cable wraps and 4806 linear feet of cable tray covers during the last week of the inspection period. The licensee cur-rently plans to complete a large majority of the remaining work prior to the implementation of site security, and the consequent reduction in craft per-sonnel, now scheduled for early Apri No violations were identifie '

14. Exit Int'erview 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 period.