IR 05000443/1985035

Insp Repts 50-443/85-35 & 50-444/85-03 on 851209-860131.No Violations Noted.Further Evaluation of Util Drawing Controls & TMI Action Plan Status Required.Major Areas Inspected: Procedures,Records & Testing of Feedwater Sys

ML20137Z955
Person / Time
Site:	Seabrook
Issue date:	03/04/1986
From:	Beall J, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20137Z938	List: IR 05000443/1985035 ML20137Z934
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.5, TASK-1.C.6, TASK-TM 50-443-85-35, 50-444-85-03, 50-444-85-3, NUDOCS 8603130247
Download: ML20137Z955 (20)

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t U.S. NUCLEAR REGULATORY COMMISSION Region I 50-443/85-35 Report N /85-03 50-443 Docket N CPPR-135 License N CPPR-136 Priority -- Category A/B Licensee: Publice Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Units 1 & 2

,, Inspection at

Seabrook. New Hampshire Inspection conducted: December 9,1985 - January 31,1986 Inspectors: A.C.Cerne, Sr. Resident Inspector (SRI)

D.G.Ruscitto, Resident Inspector

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R.S.Barkley, Resident Inspector Reviewed by: 3 h/o J .Beall, Project Engineer, ORP

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date signed

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- Approved by: 2 k D .Gallo, Chief, Projects Section 2A

dite s'igned

Division of Reactor Projects, (DRP)

! Inspection Summary:

( Inspection on December 9,1985-January 31,1986(Report Nos.50-443/85-35 &

l 50-444/85-03)

i Areas Inspected: Routine inspection by the resident inspectors of work activities, procedures, records and the conduct of testing relative to the emergency and startup feedwater systems redesign / rework; fuel handling and

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storage equipment / facilities and the readiness for fuel receipt; fire protection; piping & isometric drawing controls; and follow-up of an ev'ent involving the loss of shielding for the Boron. Concentration Monitor. The (- inspectors also . reviewed licensee action on previously identified items, including 10 CFR 50.55(e) reports, and performed plant inspection-tours. Unit 2

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inspection activities consisted of review of the preservation / protection programs in progress and plans for the completion of work required to support Unit 1 operation. The inspection involved 256 inspection-hours of Unit I and

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four inspection-hours of Unit 2 activities.

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8603130247 860307

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PDR ADOCM 05000443 G PDP l

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Results: No violations were identified. Further NRC evaluation and follow-up of licensee plans for P&ID and other drawing controls for functional verifica-tion and operational usage considerations are required (paragraph 8). Also, NRC follow-up of TMI Action Plan item status reveals the need for further NRC confirmatory inspection .

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DETAILS Persons Contacted J. DeVincentis, Project Engineering Manager (NHY)

R. E.- Guillette, Ass't Construction QA Manager (YAEC)

G. A. Kann, Phase 2-6 Test Group Manager (NHY)

G. Kingston, Compliance Manager G. J. Kline, Reactor Engineering Supervisor (NHY)

D. A. Maidrand, Assistant Project Manager (YAEC)

J. L. Marchi, Startup QC Manager G. F. Mcdonald, Construction QA Manager (YAEC)

J. W. Singleton, Field QA Manager (YAEC)

G. S. Thomas, Vice President of Nuclear Production Interviews and discussions with other members of the licensee and contrac-tor management and staff were also conducted relative to the inspection items documented in this repor . Plant Inspection-Tours (Unit 1)

The inspectors observed work activities in progress, completed work and plant status in several areas of the plant during general inspections of the plan The inspectors examined work for any obvious defects or non-compliance with regulatory requirements or license conditions. Particular note was taken of the presence of quality control inspectors and quality control evidence such as inspection records, material identification, non-conforming material identification, housekeeping and equipment preserva-tion. The' inspectors interviewed craft personnel, supervision, and quality inspection personnel as such personnel were available in the wnrk area Specifically, the inspector examined several electrical conduit supports in the Emergency Feedwater (EFW) building, the containment electrical pen-etration areas, and the east main steam and feedwater chase, checking the as-built configurations against the standard electrical design details noted on the quality control records. One nonstandard condition was not-ed, questioned and found to be acceptable based upon prior engineering evaluation (reference: . Engineering Change Authorization ECt50/105765A).

Certain installed electrical cables in the control building "B" train switchgear room were checked against the computerized conduit and cable schedule program (CASP). The inspector verified the existence of proper cable routes, raceway fill and redundant train separation and confirmed any noted nonstandard conditions to have been authorized by valid design changes (i.e., ECAs).

The inspector observed a design modification in progress on a motor con-trol center (MCC-E614). This design change was intended to increase the reliability of safe shutdown equipment (reference: 10 CFR 50, Appendix R).

The inspector witnessed in progress work activity on the control circuitry for the supply fan for the "B" train service water switchgear room

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(SWA-FN-408). He noted the installation of an additional control circuit fuse in parallel with the existing fuse, which is active only when the fan is being controll'ed from the "B" train Remote Safe Shutdown Panel (CP-108B).

The inspector reviewed the documents (i.e., Work Request, WR-SWA-125 and ECA 03/109519A) providing authorization and implementing details for the work. He also reviewed the in process QC inspection report and discussed with both the craft and the quality control inspector the substitution of No. 12 AWG wire for No. 14 wire, based upon the unavailability of the thinner wir Initiation of a Request For Information (RFI) to confirm the acceptability of such a substitution was verifie The inspector verified that the electrical power to motor operated valves, FW-V-156 and 163, on the line connecting the startup feed pump to the emergency feedwater header was supplied from emergency electrical bus E5 as stated by the licensee in a letter to NRR (SBN-392, dated November 30, 1982). He checked the conduit and cable schedule program (CASP), veri-fying that the valve operators were properly classified and powered from motor control centers (MCC's) supplied from bus E The inspector also witnessed application of protective boots on bus bar connections in the nonsegregsted phase bus duct running from diesel gener-ator "A" to bus E5 in the essential switchgear room. The boots were being applied in preparation for the installation of a BISCO fireproofing seal in the duct. The inspector reviewed work request DGG-0328 which author-ized the work, as well as the vendor information needed to determine the torque requirements on the bus bar connections. He verified that the craftsman performing the work was cognizant of the torque requirement Proper QC coverage of the activity was observe Near the aforementioned bus duct, two 1 1/2" by 3" holes were noted in a BISCO seal for a wall penetration for three cable tray The inspector questioned the responsible YAEC QA surveillance engineer regar-ding BISCO program requirements for the identification, control and repair of openings in penetration seals. The licensee's representative stated that through the use of controlled and properly autnorized release forms, assurance is provided that any rework on completed seals is approved and that any existing openings (e.g., the holes noted by the inspector) are repaire The inspector witnessed removal of the reactor vessel upper internals package and transfer to' its storage stand within the refueling cavit The inspector observed a Westinghouse representative measuring the gap surrounding the upper internal keyways prior to the lif ting of the inter-nals packag The upper forging weld was visually examined by the Westing-house representativ The inspector observed measurement of the spacing between the reactor vessel flange and the lower internals hold-down spring once the internals package was removed. The em erved work activities were being procedurally controlled by preoperational test (PT-43) for post-hot fur.ctional inspection of the reacto .

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The inspector questioned the coordinate system used in referencing the reactor vessel internals measurements and how they related to the geo-graphic orientation of the plan He was informed that Westinghouse's standard practice is to use a coordinate system that orients the O degree reference of the vessel at the 90 degree azimuth of the coordinate system of the plan This practice was adopted at Seabrook. The inspector then

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questioned whether the station Reactor Engineering Department Supervisor was cognizant of the difference in the coordinate systems since it could affect the fuel loading pattern of the cor He was informed that the fuel loading patterns of the reactor properly incorporate the 90 degree shift in coordinate system The inspector observed portions of the reassembly of cylinder No.14 of diesel fgenerator "A". Cylinders No.2 & 14 were removed to inspect their lineni for indications of microbiologically influenced currosion (MIC)

after sampling of the jacket water system revealed a microbiological in-gress. No unacceptable conditions were note The inspector discussed with the Chemistry Support Supervisor the nature and extent of the sampling program undertaken to detect microbiological ingress into safety related systems. He reviewed chemistry department procedure CH-T-85-22 which outlined the type of testing being conducted, the systems sampled and the interval of sampling. The inspector was also informed that the water supplied to the water treatment plant was being treated with chlorine, and then ozonated after processing, to kill any microbiological organism Furthermore, the licensee is planning to re-place the ozonator with an ultraviolet light sterilizer in the near future since the sterilizer will kill the bacteria without' reoxygenating the de-mineralized water. The inspector questioned whether the inspection of the diesel generator cylinders revealed any indications of MIC and was told that none were detected. However, MIC was determined to be present in the fire water storage tank liners, but analysis showed that it has not advanced far enough to adversely affect the structural integrity of the tank. This problem is documented and being tracked as a nonconforming condition on NCR 82/1131 Corrective action is planned and will be implemented per the final resolution to the in process NC The inspector had no further questions at this time on this licensee identified issu With respect to all of the above independent inspection and pl inspection-tour items, no violations were identifie . Licensee Action on Previously Identified Items (Closed) Inspector Follow-up Item (443/84-19-01): Operator training on radiation monitoring equipment. During cold license examinations conducted in 1984 and 1985, the design and selection of the Radiation Data Monitoring System (RDMS) had not been finalize It was therefore determined that an additional area of training for operator licensing coverage involved use of the radiation monitoring

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systems. The design is now complete and training and examination of licensed operators was conducted as part of the 1985 Operator Requalification Program. The inspector reviewed the training material used in this lesson and discussed the hardware and training with.several licensed operators. Based on these activities, this item is close b. (Closed) Inspector Follow-up Item (443/84-19-02): Training on site specific Radiation Work Permits (RWP). Training in the utilization of the Seabrook RWPs was determined to require additional coverage during the cold licensing examinations.in 1984 and 1985, since site specific procedures had not yet been developed. Subsequently, proce-dures were developed and were incorporated into the 1985 Operator Requalification Program held in May and June of 1985. This area will also be re-addressed in the 1986 Operator Requalification Progra The inspector reviewed the 1985 and 1986 Classroom Objectives and the Radiation Protection Program Manual and has no further questions on this issu This item is closed, c. (Closed) Violation (443/85-01-04): Design interface problem on flow restrictor orifice sizing. The inspector reviewed the licensee's response to the violation (reference: SBN-829, dated June 17,1985)

which stated that the orifice size for all flow restrictors which comprise the ASME Section III Class 1/ Class 2 boundary was intended to be a maximum of 0.375". Westinghouse ( W ) performed an analysis specific to Seabrook to determine if the hole size could be in-crease The results of the analysis were incorporated into the

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plant design. In all cases, orifice sizing greater than 0.375" was allowable, although certain restrictors had tight dimensional toler-ances. As noted in the 443/85-01 inspection report,' reinspection of all affected orifices and repair,.where required, had been conducted prior to the conclusion of the inspection and prior to the conduct of the reactor coolant system hydrostatic tes The licensee and W have concluded that the omission of acceptable design orifice tolerances from the information transmitted to UE&C for inclusion in the applicable reference drawings represents an iso-lated error. The licensee performed a review of their fluid system flow diagrams to determine if any other critical tolerances had not been transmitted to UE&C. With respect to all W supplied items, all critical interface dimensions and notes had been transmitted to UE& The inspector reviewed Engineering Change Authorization 08/104229B which transmitted the design requirements for the restrictors to the installation contractor and also listed the documents to be revise He reviewed all the affected drawings and verified that they had been revised ~as required. The W response to the licensee (reference:

NAH-2710, dated May 31,1985) discussing tr.e error was also reviewed by the inspecto There were no further cuestions on the adequacy of licensee corrective action on this item. This item is close .

7 (Closed) Unresolved item (443/85-20-03): Broken boss on the diesel generator lube oil pump discharge nozzl This problem had been identified at another nuclear site on similar diesel generators sup-plied by Colt Industrie Licensee investigation revealed that the subject lube oil pump discharge tubing' adapter had already been sche-duled for replacement at Seabrook Colt Change Order FS 1395C, dated Janua ry 18,1985. Subsequently, ECA 08/110768A was issued on October 21,1985 to replace the adapter per FS 1395 The inspector confirmed that the required design change was incorpo-rated into the Startup Incomplete Items List (IIL), a work order was issued to implement the change, and the new tubing adapter was instal-led prior to the commencement of preoperational testing of the diesel generators. This item is close e. (Closed) Unresolved Item (443/85-20-04): Completten of structural steel work begun by Perini Power Constructors, Inc. The inspector discussed with-the responsible YAEC Quality Assurance (QA) Engineer the method by which UE&C assures that all incomplete items left by the previous structural steel contractor (Perini Power Construction, Inc.) will be complete He was directed to UE&C Field Structural Procedure (FSP) 153 which governs the installation of structural-steel. All structural steel items which were not completed by Perini are documented on a UE&C Structural Steel Installation Summary (SSIS). Furthermore, it is intended that all structural ~ steel items be completed and inspected prior to turnover of the building to the

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station staff (reference: Administrative Site Procedure, ASP-5).

. If an item is not complete by building turnover, it is placed on the l Startup IIL which tracks the work until completion at a later tim The inspector reviewed an SSIS form addressing incomplete Perini items in the Diesel Generator (D/G) building. Also reviewed was YAEC QA surveillance report Q2.6.31.9628 which checked the status of in-complete Perini items in the D/G building and the cooling tower. The inspector independently verified, as a sample, that the following structural steel installations addressed on the SSIS were complete:

piece mark numbers 11K, 11T and 11V in the D/G "A" diesel fuel stor-age area and piece mark numbers 11X and 11W in the "B" diesel fuel storage area. No discrepancies or incomplete work items were identifie With respect to a prior question on the method of connection (i.e.,

bolts or welds) for specific structural support beams detailed on UE&C drawing, F101567 (Detail F), the inspector reviewed ECA 01/1094338 which clarified the use of bolting for alignment of the structural members only. The welded connections are designed to car-ry the beam loads, including seismic, and no credit has been taken for the bolts in this applicatio The inspector had previously in-

spected the welds in question and nas no questions on their adequacy and no further questions on this unresolved ite This item is close (Closed) Unresolved item (443/85-25-02): Clarification of the gap clearances in the ITT Grinnell piping struts. The licensee identi-fied 25 support points where Figure 211 size A sway struts had been used with Figure 215N, size 1A-3A stiff clamps, resulting in

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questionable gaps in the clamp / strut assembly, as identified by the NRC inspecto The design recommendation, as provided by ITT Grinnell, to install the correct spacer washers to shim the existing gaps was effected by the issuance of ECA 25/807411 The inspector reviewed the design change, noting accountability of the 25 affected supports and reference to a generic ECA (25/807343A) for proper spac-er washer sizin Furthermore, in a related matter involving potential binding of the strut assemblies because-of unacceptable angular deviations, the licensee provided evidence that ali struts on piping systems with normal operating temperatures above 200 degree:, F were monitored dur-ing Hot Functional Testing. Of the 103 strut assemblies checked for unacceptable angularity and binding, eleven discrepant conditions were identified. These nonconforming conditions have been documented on NCR 82/1111A, the disposition of which will determine the engi-neering acceptability of the installed hardware. The inspector re-viewed the NCR and noted accountability for all affected struts where angularity greater than five degrees was identified. The five degree acceptance criteria was noted to have been selected based upon the ITT Grinnell recommendatio In the cases of both strut spacer gap clearances and angularity tol-erances, the licensee has taken steps to identify, disposition, and correct, if necessary, all discrepant condition The inspector de-termined that these measures adequately address the questions and concerns raised by this unresolved ite This item is closed.

4. Licensee Action on Construction Deficiency Reports (CDR) (Closed) CDR (82-00-07): Potential shearing of pinion keys on Limitorque valve operator The inspector reviewed the licensee Fi-nal 10 CFR 50.55(e) Report (SBN-506, dated May 4,1983) and discussed corrective action implementation with YAEC QA engineering personne Westinghouse (W) personnel have inspected all W supplied motor operated valves on-site and determined that the only SB-0 size opera-tors (for which W has elected to replace the pinion keys) had been documented on hr Field Change Notices (NAHM-10523 & NCHM-10518). The

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inspector reviewed both FCNs, noting that replacement of the motor pinion keys in the af fected valve operators was completed in June 198 The inspector also questioned the need for replacement of any pinion keys on Limitorque valve operators procured under the UE&C scope of supply. Licensee inspectica revealed that no UE&C purchased valves were supplied with the SB-0 model operator which requried replacement of the pinion key. Furthermore, an inspection in response to IE In-formation Notice 81-08 revealed that no UE&C valves were identified with Limitorque SMB-4 operators, which was the type having experienced failures at other nuclear plants because mild steel had been used in the motor pinion gear key In general, inspections of Limitorque operators at Seabrook, includ-ing checks of the motor pinion keys, have been routinely implemented under the startup test con +.rols, initially under the relubrication program (General Test - GT-M-102) and more recently under a general inspection of motor operated valves (GT-E-113). Test results have been randomly spot-checked by the inspector during past routine in-spection activitie The inspector has no further questions on the extent of licensee cor-rective action to address this deficienc This CDR is close b. (Closed) CDR (84-00-07): Westinghouse type SA-1 differential re-lays. The initial problems with W SA-1 differential relays were doc-umented in IE Information Notice 83-63. A Westinghouse investigation revealed that certain relays containing silicone-controlled rectifi-ers (SCR) manufactured by S.T.Semicon can cause a random trip outpu Replacement with a Raytheon SCR was the proposed solution. Supple-ment I to IE Information Notice (IN) 83-63 expanded the scope of problems' identified with the SA-1 type relays to include insufficient surge withstand capability (SWC) and internal capacitor failure Although the relays used at Seabrook do not contain any SCRs manufac-tured by S.T.Semicon, it was determined that these relays may be sub-ject to the two more recent problems reported in the IN 83-63 supplement. A Nonconformance Report (NCR 82-255) was written to doc-ument the problem and a Work Request (WR ED2E-0085) was issued to install the manufacturer supplied replacement kits in each of the emergency switchgear units. The work was completed in February,198 The inspector reviewed the relevant NCR and WR, as well as the dif-ferential relay calibration test records, quality control inspection report, materials store house order and control drawings. Based on licensee corrective actions, the inspector determined tha* the hard-ware replacement was properly conducted, as was committed and trans-mitted to NRC Region I in the Final 10 CFR 50.55(e) Report (SBN-843, dated July 10,1985). This CDR is close c. (Closed) CDR (84-00-12): Cracking of the diesel generator exhaust silencer concrete pedestals. The corrective action for this design deficiency was previously reviewed in Inspection Report 443/85-20

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and was the subject of violation 85-20-01. At that time, the uesign details for the replacement support pedestals, issued as the dispost-tion to NCR 82/221G, were reviewed. Inspection of the work revealed that the bolting work on the pedestals was incomplete, however, the licensee's final report to the NRC stated that the deficiency had been correcte The inspector reviewed the licensee's letter to the NRC (reference: SBN-835 dated July 10,1985) which clarified their final report on this CDR. Also reviewed was Nonconformance Report 82/616A which described the field work needed to correct the deft-cient bolting on the pedestals. All bolted connections for both die-sel generator exhaust silencer pedestals were verified by the inspector to be complete. To verify completion of the reworked pedestals, as redesigned, the inspector spot-checked the dimensions of the support pedestal cover and stiffener plates and the installa-tion of foundation grouting. All work was found to agree with the design details. In addition, the inspector reviewed certain QC inspection records related to the grouting installation and verified these documents to be accurate and compelt The corrective actions undertaken by the licensee as a result of vio-lation 85-20-01 will be reviewed during a future NRC inspection, This CDR is closed.

d. (Closed) CDR (85-00-01): Gould molded case circuit breakers. The licensee stated in their Final 10 CFR 50.55(e) Report to the NRC, Re-gion I. (reference: SBN-877, dated October 9,1985) that the reason the 40 ampere Gould molded case circuit breakers failed the instantaneous overcurrent trip test was due to the use of an incorrect time-current response curve as supplied by Gould for test acceptance criteri The licensee stated that all the circuit breakers passed the instan-taneous overcurrent trip test when retested using the values obtained from the correct time-current response curve. Commitment was made that Gould would test all safety-related (Class IE) circuit breakers to the requirements of NEMA AB-2 prior to shipment to Seabroo The inspector reviewed Nonconformance Reports (NCRs) 82/223 & 82/350 which addressed the corrective actions taken on the breakers and confirmed compliance with the commitments made to the NRC. He also discussed with the responsible Startup Test Engineer the testing con-ducted on these breakers, the testing requirements, and the results of the instantaneous overcurrent testing which were found acceptable relative to the correct time-current response curv Licensee corrective actions on this item address the discrepancy doc-umented in the deficiency report. This CDR is closed.

e. (Closed) CDR (85-00-05): Safety Injection (SI) limit switch bracket integrity. As documented in Inspection Report 443/85-20, previous NRC review of corrective action en this CDR revealed inconsisten-cies between the final as-built configuration of the subject limit switch brackets and the design details. This was identified as a

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violation (85-20-02) and will be reviewed during a future NRC inspectio The inspector reviewed Westinghouse (W) calculations performed to substantiate the licensee position that the subjects brackets, as installed, are capable of performing their safety-related function and therefore a reportable. deficiency does not in fact exis Included in the inspector review were design calculations for the

" Analysis of External Limit Switch Bracket 9580-022 with Two NAMC0 Limit Switches, EA170 or EA180", Revision 0-SC8065 and Revision 1-SC806 The inspector verified that the detailed W calculations either model or envelope the as-built limit switch bracket configurations, as confirmed by field inspection. Actual weld details, critical sec-tions, and part sizes and thicknesses were confirmed to have been considered in the stress calculation The results revealed that existing bracket configurations will withstand at least 100,000 fully reversed load cycles of 4.59 acceleration, considering component dead weight and seismic and trip load Since the procurement specifica-tion indicates acceptance requirements to a minimum 2.lg loading, the reviewed calculations substantiate the licensee position that this item is not a reportable deficiency (reference: SBN-830, dated June 18,1985).

This CDR is close . Fire Protection Activities ('Jnit 1) - The inspector witnessed the application of Pyrocrete flame retardant material on several structural steel beams in the train "A" essential switchgear roo He verified, as a sample, that the material was applied to th? specified depth and questioned the craft regarding the Pyrocrete mixing proces Repair activities to previously applied Pyrocrete were also observed to be in progress in the train "B" es-sential switchgear room and in the areas acove the control roo The inspector spot-checked sections of the control building fire-proofing application to the design details of UE&C drawing F101360 (Revision 1). The inspector confirmed that fireproofing material was installed on the structural steel members at elevation 50', where a three-hour fire barrier was required by design. In the area above the control room, the inspector noted that NCRs had been issued to document and repair areas (e.g. , elevation 93') where, at the edges of structural beams, the thickness of the existing fireproofing mate-rial was found to be deficient. The inspector questioned the mea-sures taken by the licensee to assure that this problem was confined to the local areas being repaired (referen:e: Blue Sheet 096).

The licensee provided documented evidence of fireproofing " low-spot" inspections by the original fireproofing c:rtractor (EF Byrnes), as

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conducted by Byrnes QC personnel prior to the transfer of work to subsequent fireproofing contractors, Lundeen, Inc. and UE&C respectivel Additionally, the responsible UE&C engineer was interviewed and it was determined _that the repair work above the control room, as noted by the inspector, was being accomplished primarily because of the need to remove and reapply fireproofing to accommodate the installa-tion of other structural attachments to the subject beam Thus, while the questioned NCRs (e.g., 59/5417) were dispositioned to require repair, the engineering determination was that if the beams had not been repaired, the localized deficient areas were not signi-ficant enough to affect the performance of the fireproofin The inspector had no further question on the fireproofing application and repair activitie No violations were identifie While reviewing the Seabrook fire protection program " Evaluation and Comparison to Branch Technical Position (BTP) APCSB 5 5-1, Appendix A" (Revision 3, dated December, 1985), several apparent tire zone incon-sistencies were noted by the inspecto The control red drive motor generator (MG) set enclosure was identified as a separate fire zone on one drawing and not on others. In addition, the horizontal (floor)

boundaries in the "A" train electrical chase to the containment pene-tration area were not consistent between drawings. The inspector discussed these inconsistencies with licensee engineering and licen-sing personnel and determined that drawing errors were in evidence, but that the physical fire protection criteria had not been violate In fact, the MG set enclosures are part of the "A" train essential switchgear fire zone and the "A" train electrical chase has its fire zone defined by the concrete enclosure boundarie Licensee personnel clarified for the inspector the fact that the reviewed document was a response to the NRC Branch Technical Position on fire protection and would not be used for plant fire response activities. Since the drawing errors were minor and affected neither the plant fire zones, nor the intent of the BTP response as a licensing document, the inspector had no further questions on this issu No violations were identifie . Emergency & Startup Feedsater Pump Redesign and Rework (Unit 11 The inspector witnessed turbir.e driven emergency feedwater (EFW) pump casing removal and resealing being conducted under the supervision of the vendor (Terry) representative. The casing had been previously removed to allow access to various fasteners which had been assigned modified torque values in a recent field change necessitated by cold start transients observed during system testin The vendor repre-sentative not only supervised the operation, but instructeJ the main-

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tenance personnel in proper techniques, work practices and the turbine design feature During Hot Functional Testing (HFT) of the EFW system, while trouble-shooting high bearing temperatures on the turbine-driven EFW pump (reference: NRC inspection report IR 443/85-31), the mechanical seal on the pump was disassembled. One of the machined surfaces on the pump shaft, upon which the seal cartridge set screw adjustment is made, was found to be tapered rather than flat as required by desig The tapered set screw is believed to have. caused the cartridge to back away from the casing as the set screws were adjusted, creating a

,.otential for mechanical seal leakage. As corrective action, both the motor-driven and turbine-driven EFW pump rotating assemblies were removed and returned to the vendor (Ingersoll-Rand) for remachinin A Nonconformance Report (NCR 82/1084A) was written to document the proble The inspector reviewed the NCR, pertinent drawings and the pump tech-nical manual and discussed the repair with the system test engineer (STE) and the vendor representative. He questioned the licensee re-garding the potential for a similar deficiency on the nearly identi-cal startup feed pump (SUFP). Licensee investigation revealed that although the pumps and seals (Durametallic) are similar, the actual configuration of the SUFP seal does.not have seal cartridge set screws. After reviewing the SUFP seal drawings and verifying actual seal configuratSn, the inspector had no further question Following remachining, the pump rotating assemblies were reinstalled under vendor supervisio Testing of these pumps will be delayed until completion of the design change in progress as a result of wa-ter hammer and recirculation flow problems encountered during HF The EFW problems and the subsequent redesign and rework and were documented as a test exception and part of an unresolved item in NRC IR 443/85-30, and will require hot retesting of the EFW system prior to initial criticality. NRC review of licensee corrective action and witness of retest conduct is planned during future inspection activ-ities and is required to close the unresolved item (85-30-01) on HFT system verificatio b. The startup feed pump (SUFP) lobe oil cooling system was redesigned to make it independent of the secondary component cooling water (SCCW) system by installing a tap off the SUFP discharge as the cool-ing medium. While examining this modification in the field, the in-spector questioned the adequacy of the piping supports on the branch line which taps off the discharge of the SUFP, flowing through the lube oil cooler and into the pump suctio A review of the design change (DCN-680260A) indicated that the mod-ification had not been reviewed by the piping support group as re-quired by Power Discipline Instruction Memo PDIM-01 Reevaluation of the as-built configuration of the redesigned lube oil cooling

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system by licensee engineering personnel indicated that additional piping support was required to dampen vibratio Engineering Change Authorization (ECA) 25/807277A was subsequently written and the re-quired support point installe A procedural memo, originally issued on August 6,1985 by UE&C site engineering to all piping support personnel (SM#12672A) concerning change documents affecting supports, was reissued on December 24,1985 (SM#13789) to reemphasize compliance with design chaage review re-quirements already in effec Based upon the fact that: (a) a pro-gram already existed to ensure multidisciplinary review of design changes, (b) the licensee was prompt in reissuing the above memo, and (c) the seismic support of the subject system is- not a design require-ment, the inspector determined the lack of support group review, in this example, constituted an isolated case for which licensee correc-tive action was both complete and comprehensive. He has no further questions at this tim Additionally, while reviewing the operating procedure associated with SUFP startup (051035.02), it was noted that the procedure still ref-erenced the ;1d cooling water supply (SCCW) instead of the new modi-ficatio Since SUFP operation was in progress, the inspector questioned the process by which design changes are incorporated into operational procedure changes. Following discussion with the Engi-neering Services Department Supervisor, the inspector reviewed the Nuclear Production Design Control Program Manual (Revision 3, 11-21-85) which will be implemented at the time of final system ac-ceptance turnover or 90 days prior to fuel load whichever comes firs This document establishes procedures for design change con-trol and includes the method by which operating procedures are updat-ed following design changes. The inspector noted that this process is implementeo and controlled using Design Coordination Reports (DCRs). In addition, the inspector confirmed that after design closecut both the licensed operator and non-licensed operator training organizations will receive controlled copies of approved final design packages and revised design documents to allow adjust-ments in the training curriculum. He had no further questions on the planned program of procedural controls for disseminating design change information to the necessary operations personne No violations were identifie . Preoperational resting (Unit 1)

The inspector reviewed and witnessed portions of Preoperational Test (PT)

1-PT-32 (Revision 0) for the 120 VAC Vital Instrument Power Supply. Spe-cifically, he reviewed sections 3,4,5 and G.1 - 6.10 of the PT. The in-spector discussed the test conduct with the responsible Startup Test Engineer (STE) to include information on the testing apparatus. The

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l inspector examined traces of the waveform for the AC output from inverter 10, recorded during steps 6.6 and 6.7 of PT-32 and determined evidence of a valid waveform, given the testing acceptance criteria for levels of dis-tortion. All current field changes and test exceptions were also re- !

viewe No unacceptable conditions were note ;

The inspector discussed with the Engineering Services Department Supervi-sor a problem with certain terminal board components in inverters 1A,8,C &

0, which the licensee identified during surveillance testing and reported to NRC, Region I per the requirements of 10 CFR 50.55(e) (reference: CDR 85-00-21 dated December 26,1985). He examined the subject terminal boards in inverters IA and IC and noted the nature of the deficienc The in-spector was informed that the deficient terminal board in inverter 10 had been replaced prior to testing, but that supplier problems prevented re-placement of the other terminal boards at that tim The inspector con-cluded that the PT-32 test results could not be adversely affected by terminal board deficiencies, still in evidence, on inverters IA, B & Replacement of these terminal boards and any other corrective action will be followed-up in a future inspection upon licensee closecut of CDR 85-00-21.

No violations in the conduct of PT-32 testing activities were identified.

As noted in paragraph 2 of this report, during a plant inspection-tour, the inspector observed post-HFT measurements being taken in conjunction with the conduct of PT-43 on the upper internals assembly by the NSSS sup-plier (Westinghouse). These keyway measurements were taken on the gap between the pads and the up'per core plate. The keyway mates with slab-sided pins in the cere barrel allowing radial expansion and some axi-al movement, while preventing transverse movement of the upper core plate.

The keyways have stellite faced pads and have exhibited some evidence of pad / plate gap enlargement at other facilities with the potential for less than a full fit in the keyway. An interview with the vendor representa-tive revealed that the gap enlargement at Seabrook appears to be accept-able based upon a preliminary evaluation of the measurements.

The inspector also witnessed indexing tests of the spent fuel pool bridge crane using a dummy fuel module. This test verifies that the indexed po-sitions for both North-South and East-West movement precisely locates the suspended element above the proper cells and allows lowering of the ele-ment into the cell without binding. Additionally, calibration testing of the digital load cell was witnesse This cell has replaced the gauge type (0111on) load cell used previously in similar testing activities.

The inspector also witnessed calibration of the pit-side fuel transfer system pressure gauge and pressure switch, as such calibration was being conducted in preparation for pre-operational testing of the fuel transfer and handling equipment. The calibration data sheet was reviewed with the technicians and test equipment was also inspected for calibration.

Through interviews, the inspector determined the technicians to be knowl-edgeable concerning calibration techniques and procedures. Phase I

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testing of the cask handling crane was observed, as well as portions of PT-34.1, the Fuel Handling. Equipment preoperational test. The inspector reviewed PT-34.1 through step 6.5, which was in progress. The one docu-mented field change had been properly entered into the test package, the chronological log and test data were determined to be current, and the one noted test exception was adequately documente In preparation for fuel receipt, the inspector confirmed that operations department personnel are being given training on the utilization of the fuel handling equipment. The inspector observed operator training given in the Fuel Storage Building by one of the training center instructor Practical experience was provided in the form of startup checks and opera-tion of the spent fuel bridge crane and the new fuel elevator using the dummy fuel assembl The instructor appeared to be well prepared and knowlegeable about both equipment and procedures. The inspector verified that such training is mandatory for all operators and has no questions regarding training or operator personnel readiness for fuel receipt in this regar Additionally, as follow up to some testing utilizing dual temperature in-dicators (TI) on the main control board (MCB), the inspector questioned whether the markings on the electrical termination side of these instru-ments presented a correct view of the association of the MCB cables and terminations with their respective tis. A startup quality control request for information (RFI 99/112251A) was generated to clarify the design re-quirements for such MCB dual indicator markings. Subsequently, a design change (ECA 05/113442A) was issued to add to the MCB Specification (170-1)

information on the method of mounting dual indicator identification tags so as to provide correspondence between the cable terminations and instru-ment tag marking The inspector had no further questions on this issu No problems were identified with any testing activities as they related to the MCB dual indicator tagging syste With respect to.all of the above preoperational testing activities, no violations were identified.

8. Piping and Isometric Drawingl P & ID) Controls During recent NRC inspections, the inspectnrs identified some discrepancies between the p & 10 details and the as-built location or orientation of piping and components. In all cases, the equipment has been determined to have been properly installed, but the question as to whether the P & 10 discrepancies constitute drawing errors has remained. While some P & ID questions have involved labeling and positioning errors which the licensee agrees require drawing revision, other examples have been more of an operational nature, such as, the location of valves relative to shield walls and the order of branch lines off a piping heade _ _ _ _ _ _ . _ _

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In these examples, the licensee has stated that functional design capabi-lity of the affected system has not been adversely affected and therefore no drawing revision is r.eede However, discussion with New Hampshire Yankee (NHY) engineering management personnel has revealed that the operations staf f has prepared their own set of P & ids, distinct from the UE&C design P & ids, which are intended

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for in plant operations utilization and training. On these drawings, which have been produced on a Computer Assisted Drawing (CAD) system, such operational questions as valve location relative to shield walls and pip- ,

ing branch line order have been correcte Thus, it appears that two sets of P & ids will be available at Seabrook -- the UE&C design drawings, in-cluded in the FSAR, which are functional representations of the piping and instrumentation systems, and the NHY CAD version presenting a more accu-rate operational view of the same piping system The inspector discussed with NHY licensing and engineering personnel what impact two distinct and different sets of P & ids may have on Seabrook for future engineering, licensing, and operations. The licensee is currently investigating the option of submitting the NHY CAD version of the P & ids to the NRC in a future revision of the FSAR, thus making them the official design drawings and combining the functional and operational purpose of the P & ids into one set of print Pending further licensee progress in this direction and additional NRC review into the potential impact, i f any, of ouplicate and different sets of piping and isometric drawings at Seabrook, this issue is unresolved (443/85-35-01).

- TMI Actic- Plan Status

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During this period, inspection to verify compliance at' Seabrook with the TMI Action Plan requirements of NUREG-0737 was initiated. The inspector reviewed documentation associated with these iems, to include applicable FSAR and SER sections and correspondence from the licensee to the NRC Of-fice of NRR providing information and commitments as to Seabrook's compli-ance. Hardware inspection of affected components and systems has also begu The inspector noted that while the commitments for NUREG-0737 conformance had generally been met by the licensee, implementation of certain committed programs had not proceeded sufficiently for the inspector to confirm acceptability. Two action items in particular (I.C.5 and I.C.6) lacked implementation progress such that a final evaluation of the licensee's attendant programs, using licensing criteria, could not be completed and documented in the Seabrook SER without further NRC confir-matory action. A future supplement to the SER will document the necessary Confirmatory Action Item l Further NRC inspection of these confirmatory action items, as well as all TMI Action Plan items, requiring follow-up, is planned and will be con-ducted over the course of future NRC inspections of Seabrook activitie ;

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1 Loss of Radiation Shielding Water From the Baron Concentration Monitor The inspector was informed by the licensee on December 20,1985 that the ,

heater control mechanism of the boron concentration monitor (BCM) failed high. This failure caused all the shielding water in the BCM to boil off, increasing radiation levels in the area above those permitted in an unre-stricted area as defined in 10 CFR 20.105 (Note: the BCM contains a four curie americium-beryllium neutron source).

The inspector reviewed a record of a radiation survey of the local area surrounding the BCM that was conducted immediately after the failure was discovered. The survey data indicated radiation levels on contact with the BCM as high as 25 mrem / hour with levels decreasing to about 1 mrem / hour at a distance five feet away. No external contamination of the BCM was detecte The inspector noted that, as a precaution, the licensee had roped off the area within approximately five feet of the BCM and had designated it a radiation area prior to the time of this inciden The BCM has been the subject of prior NRC inspection (reference: Inspection Report 443/85-25 dated November 15,1985) at which licensee controls in the area of the BCM were confirmed. The licensee's records indicated that no one had cntered the radiation area since the last time the BCM radiation levels were known to be normal (i.e., contact BCM readings at approximately three mrem / hour). Since craft were known to have been working on scaf-folding'approximately ten feet from the BCM, a radiation survey of that area was conducted and indicated the radiation levels there at about one mrem / hour, which is less than the limit for unrestricted areas as defined by 10 CFR 20.105. Based upon the radiation levels in the area of the BCM, the marking of the radiation area prior to the incident, and the fact that no personnel were known to have entered that area, the licensee determined this event to be nonreportabl Based upon the survey data recorded, the inspector concurred with this determinatio The licensee has since removed the neutron source from the monitor and confirmed that it is intact. This information confirmed the accuracy of the contamination survey conducted on the exterior of the BCM immediately after the inciden Since the loss of shielding water and resulting increase in radiation lev-els surrounding the BCM had been previously questioned by the inspector in report 85-25, the inspector checked the results of the survay data with the estimates of the radiation levels based upon analysis of such an event (as listed in the licensee's Systems Training Manuals). The radiation levels detected were less than those that had been estimated, confirming conservatism in the licensee's prior analysi The inspector had no further questions on this inciden No violations were identified as a result of this even :l

1 Preservation and Protection Activities (Unit 2)

.The inspector reviewed NHY Administrative Site Procedures, ASP-6 on the

" Preservation, Storage and Preventive Maintenance of Seabrook II Equipment, Materials and Structures", and ASP-7 on an " Atmospheric Corro-sion Monitoring Test Procedure" for Unit 2 and discussed the implementa-tion of both procedures with the responsible NHY and UE&C personne Plans for the continued preservation / protection of Unit 2 components were confirmed and plans for a long term study to evaluate the licensee's op-tions with regard to the future.of Seabrook 2 were discussed with NHY managemen The inspector noted that the licensee status of Seabrook Unit 2 is presently categorized as " indeterminate", but that licensee efforts to maintain the option to resume construction have been ongoing. In this regard, the inspector confirmed that Unit 2 equipment, which is identified as nonconforming, is being tracked in a manner which allows for future correction of the nonconforming conditions and closure of the open NC Similarly, construction deficiency reports (i.e. ,10 CFR 50.55(e) reportable)

are being tracked by a computer coding system, which distinctly tracks the Unit 2 CDRs, even when the comparable Unit 1 items are correcte The inspector also reviewed an engineering memorandum discussing licensee action to address a previous NRC question (reference: Blue Sheet 56) in-volving protection of the Control Building Air (CBA) west intake piping, which is routed thru the Unit 2 construction area to the Unit I control room and which is required for Unit 1 operation. In this memorandum (File TF B4.2.7, A16.2.7) various options to completion of the remaining work in the Unit 2 area are discussed, with engineering recommending that con-struction adopt the option to backfill only that portion of the Unit 2 area (between the turbine building and radiation control area tunnel) so as to provide seismic integrity to the Unit 2 turbine building and protect the CBA west air intake pipin The inspector has no questions regarding licensee engineering, handling and tracking of the Unit 2 and common work items required to support Unit 1 operation. With respect to the Unit 2 preservation and protection pro-grams reviewed during this inspection, no violations were identifie . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or devi-ations. An unresolved item disclosed during this inspection is discussed in Paragraph ~

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13. Management Meetings At periodic intervals during tne course of this inspection, meetings were held with senior plant management to discuss the scope and findings of this inspection. An exit meeting was conducted on February 4,1986 to dis-cuss the inspection findings during the period. During this inspection, the NRC inspectors received no comments from the licensee that any of their inspection items or issues contained proprietary information. No written material was provided to the licensee during this inspection.