IR 05000443/1986020

Insp Rept 50-443/86-20 on 860401-0523.No Violation Noted. Major Areas Inspected:Waste Process Bldg,Records,Rhr Vaults Turnover Preparations & Controls for NUREG-0737 Items

ML20199L119
Person / Time
Site:	Seabrook
Issue date:	07/03/1986
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199L085	List: IR 05000443/1986020 ML20199L081
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.3, TASK-1.B.1.2, TASK-1.C.2, TASK-1.C.3, TASK-1.C.4, TASK-1.C.7, TASK-1.D.2, TASK-2.E.3.1, TASK-2.F.2, TASK-3.A.1.2, TASK-TM 50-443-86-20, IEB-79-13, IEB-79-17, IEB-81-02, IEB-81-2, IEB-83-03, IEB-83-3, NUDOCS 8607090315
Download: ML20199L119 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-20 Docket N License N CPPR-135 Priority -- Category A/B Licensee: Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit 1 Inspection at: Seabrook, New Hampshire Inspection conducted: April 1 - May 23,1986 Inspectors: A. C. Cerne, Sr. Resident Inspector D. G. Ruscitto, Resident Inspector R. S. Barkley, Resident Inspector J. A. Schumacher, Reactor Engineer C. F. Holden, Jr. , Sr. Resident Inspector, Maine Yankee R. A. McB arty, R ctor Engineer Approved by: /. _

7 N T. C. Elsas hief, Reactor Projects Section 3C Date Summary: Inspection on April 1 - May 23,1986 (Report No. 50-443/86-20)

Areas Inspected: Routine inspection by four resident inspectors and two region-based inspectors of work activities, procedures, and records relative to the waste process building and its systems; primary auxiliary building, control building and RHR vaults turnover preparations; the review and witness of preoperational testing activities; review of training programs and the follow-up of licensee scheduled activities and controls for TMI Action Plan items. The inspectors also reviewed licensee action on previously identified items, including 10 CFR 50.55(e) reports, and performed plant inspection-tours. The inspection also included observation of the techniques proposed to be used for NDE examination of reactor coolant piping welds. The inspection involved 439 inspection-hours by six NRC inspector Results: NRC witnessing of pre-operational testing has disclosed no major problem Review of selected TMI Action Plan Items, Construction Deficiency Reports and lic-ensee response to IE Bulletins and Circulars revealed no safety concerns. No violations were identifie gDR ADOCK 05000443 PDR

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DETAILS Persons Contacted G. S. Thomas, Vice-President, Nuclear Production (NHY)

J. DeVincentis, Director of Engineering (NHY)

G. F. Mcdonald, Construction QA Manager (YAEC)

D. E. Moody, Station Manager (NHY)

L. A. Walsh, Operations Manager (NHY)

P. M. Richardson, Training Center Manager (NHY)

R. E. Cyr, Maintenance Manager (NHY)

J. J. Warnock, Nuclear Quality Manager (NHY)

R. E. Guillette, Assistant Construction QA Manager (YAEC)

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

D. G. McLain, Startup Test Group Manager (NHY)

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

Interviews and discussions with other members of the licensee and contractors management and staff were also conducted relative to the inspection items documented in this repor Licensee Personnel in Attendance at Management Meeting on May 22, 1986 (See Paragraph 13)

W. B. Derrickson, Senior Vice President (NHY)

W. P. Johnson, Vice President / Director of QA (NHY)

G. S. Thomas, Vice President, Nuclear Production (NHY)

J. DeVincentis, Director of Engineering (NHY)

G. F. Mcdonald, Construction QA Manager (YAEC)

Plant Inspection Tours The inspectors observed work activities in progress, completed work and plant l status in several areas during general inspections of the plant. They examined

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work for any obvious defects or noncompliance with regulatory requirements I or license conditions. Particular note was taken of the presence of quality

! control inspectors and quality control evidence such as inspection records, material identification, nonconforming material identification, housekeeping and equipment preservation. The inspectors interviewed craft personnel, supervision, and quality inspection personnel as such personnel were available in the work area The inspector questioned the procedure used to repair a damaged floor drain system (WLD) line which was embedded in the seismic, concrete wall of the #2 RHR equipment vault. The line had been inadvertently cut during concrete core drilling operation The repair required a 12" diameter core bore on both sides of the wall to access the damaged area of the pipe. Inner and outer patches were welded to the line in accordance with Non-Conformance Report (NCR)

73 0731A. The inspector reviewed this NCR as well as the pipe repair with l acceptable findings.

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While inspecting the containment building at the O'-0" elevation, the inspec-tor questioned the color coded identification of the containment air handling units (CAH-AC-1A,18,1C,10,1E & 1F). They were identified by red nameplates, indicating that these units are safety-related Train A equipment. Following licensee inspection as documented in QA surveillance report Y-1538, it was confirmed that the CAH units are non-safety related and were improperly labele Startup work requests (WR) CAH-347 and CAH-350 were issued to remove the old tags and install the proper labelin With regard to all of the above indepen8ent inspection and plant inspection-tour items, no violations or unresolved safety concerns were identifie . Licensee Action on Previously Identified Items (Closed) Unresolved Item (443/85-15-07): Welding Qualification Records Program Weaknes This item concerns incorrect qualification record The licensee committed to revise Weld Procedure Specification (WPS) S-FWP-300, train welders on the use of the procedure, and audit welder qualification records to verify that those records were correct. The in-spector reviewed welder training records, the revised WPS, and audit report SA962CS465, and found that the licensee's commitments were met, and the actions were completed. Based on the inspector's review, this item is considered close (Closed) Violation (443/85-15-02): Breakdown in Control of Contractor Interfaces With Regard to Support Stiffener Addition. The inspector re-viewed the licensee's document package which contained a revision to Field General Construction Prccedure (FGCP) No. 27 entitled, "UE&C/ Con-tractor Interface on Release of Equipment and Tanks". Procedure revi-sions were reviewed and found to adequately define the requirements for stiffener plate installations and the associated interface requirements between contractors assigned responsibility for such installations. In addition, the modification sheet to Engineering Change Authorization (ECA)

05/102323A properly included the required stiffener addition. Applicable drawings necessary to support the stiffener addition were included. A

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field inspection of the affected support stiffener found the installation to be in accordance with the EC The inspector also reviewed approximately thirty ECA's associated with stiffener / support installation These changes were found to contain the documents necessary for both installation and proper contractor in-terface. This violation is close (Closed) Violation (443/85-20-01): Corrective Action Failure on Diesel Generator Pedestal Cracking. Interim procedure changes to Field Piping Procedures (FPP) No.1 and No.4 were reviewed and found to adequately address inspection requirements for site QA personnel. Project Policy Statement No. 27 titled, " Reporting of 10 CFR 50.55(e) Information to the NRC" was revised to ensure equipment and/or engineering activities are properly verified prior to closecut of Construction Deficiency Re-

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ports (CDR). In addition, a memorandum from the Assistant QA Manager was sent to all construction QA personnel defining their reponsibilities with respect to verification of hardware before reporting that a CDR has been completed. The inspector visually inspected the subject diesel generator pedestals and found them to be installed in accordance with specified standards. This violation is close (Closed) Violation (443/85-20-01): Inadequate Design Change Control and Corrective Action on Limit Switch Bracket Rework. The inspector reviewed the licensee's document package which included a directive from Site Engineering that provides guidelines to eliminate any cause of confusion and prevent potential rework when welding a non-safety-related item to a safety-related ite These guidelines state the conditions under which the safety related block on NCR forms shall be used and when required QA inspections shall be performed and documented. Also, contractor per-sonnel have been instructed via an interoffice memorandum not to accept verbal instructions to clarify conflicts on design documents. Design changes will be submitted to Engineering which will ensure adequate in-formation is available to preclude conflicts. A visual inspection of the subject limit switch bracket rework found the installation satisfac-tory and in accordance with design documentation. This violation is close . Licensee Action on IE Bulletins, IE Circulars and Construction Deficiency Reports (Closed) IE Bulletin (IEB 79-17): Pipe Cracks in Stagnant Borated Water Systems at PWR Plant United Engineers and Constructors (UE&C) letter SBU-953345 dated September 4, 1985 lists measures taken to minimize the initiation of intergranular stress corrosion cracking at Seabrook. Ad-ditionally, the licensee has established a supplemental examination pro-gram plan (SEPP) for preservice inspection (PSI)/ inservice inspection (ISI) which includes welds in Class 2 systems that fit the stagnant borated water conditions. The inspector reviewed the SEPP and confirmed it is in place and that it is being implemented as part of the facility PSI progra Based on the above this item is close (Closed) IE Bulletin (IEB 79-13): Cracking in Feedwater System Pipin A licensee memo dated October 18,1985 lists recommendations by the Opera-tional Engineering Supervisor (OES) regarding feedwater line cracking as discussed by IEB 79-13. These include the volumetric non-destructive examination (NDE) of feedwater lines at the steam generator nozzle-to pipe weld and adjacent areas and the development of an augmented ISI program for NDE of those welds and areas. The facility ISI program is dependent upon the date of issuance of the operating license, and therefore has not yet been established. The inspector reviewed ultrasonic examination (UT) records to confirm that the examinations were completed following hot functional testing (HFT) and prior to fuel load in accordance with the OES recommendation. The areas of examination agreed with the OES recommendations and with the IEB requirements. Based on the above, this item is considered closed.

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c. (Closed) IE Bulletin (IEB 83-03): Check Valve Failures in Raw Water Cooling Systems of Diesel Generators. The inspector reviewed the Piping and Instrumentation Diagram (P&ID) details for the service water (F805019 and F805033) and diesel generator cooling water (F202103) systems. He confirmed that the only check valves in the pipe lines supplying cooling water to the diesel generators (DG) exist in the discharge piping of the service water (SW) and cooling tower water pumps. These valves are in-cluded in the Code Valve Test List, FSAR Table 3.9(B)-23, and will be periodically exercised under the Seabrook Valve Inservice Testing (IST)

Program in accordance with the ASME Code Section XI, IWV-3414 and 3520 code requirements, as applicabl IEB 83-03 specifically focuses on check valve failures in the DG cooling system where valve integrity would be challenged, not continuously, but only as the DG are placed in service. There are no such check valves in the DG cooling system at Seabrook. Since the SW system valves are normally fulfilling their safety function with the SW system in operation, the check valves are continuously monitored for proper opening or seating by observation of the system flow rates or by pump testing. Thus, re-duced flow rates caused by valve integrity problems would be identified as they occu The inspector discussed licensee action on this bulletin with both the responsible YAEC Senior Engineer and the NRC Technical Contact from the Office of I He determined that no further action on IEB 83-03 is re-quired with respect to the Seabrook DG cooling water system desig This bulletin is close d. (Closed) IE Circular (IEC 78-16): Limitorque Valve Actuators. Licensee actions to respond to the concerns raised by this circular were previously reviewed during the NRC Region I 443/85-31 inspection. At that time, a Limitorque recommendation that the valve actuator not be used to manu-ally seat a valve further than its electric motor had the capability of initially seating it was still under consideration by the license During this inspection, the inspector reviewed an Operations Department Instruction, ODI.10 regarding " Checking Valve Positions and Circuit Breaker Alignment" and noted that the use of " unreasonable force" in operating valves was procedurally prohibited. Also, a recent request by the Operations Department for training on Limitorque valves specifically indicated that operators would not be allowed to use mechanical extension assists in attempting to seat leaking valve These instructions will be added to ODI.1 The problem discussed in this circular was addressed to Seabrook in a letter from Limitorque Corporation to UE&C, dated August 31,1978. The licensee has now taken action on all of the Limitorque recommendation This circular is close .

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6 (Closed) IE Circular (IEC 80-21): Regulation of Refueling Crews. This circular highlighted existing regulatory requirements relevant to the assignment, qualifications, training and responsibility of fuel handlers and their supervisors. Operations Procedure OS1000.19, Rev. 00 entitled

" Refueling Operations" was reviewed by the inspector. He noted that the requirements of IEC 80-21 were spelled out in the procedure as well as Technical Specification 6.2.2d. (Proof and Review copy). This item is close IE Bulletin (IEB 81-02, 81-02 Supplement); CDR 80-00-04, CDR 81-00-04, CDR 81-00-07: Failure of Gate Type Valves to Close Against Differential Pressure IEB 81-02 was issued in April of 1981 and its supplement was issued in August 198 These bulletins described testing which indicated that certain gate-type motor-operated valves (M0Vs) failed to close under simulated service conditions. These valves are used in various safety related applications. Valve nominal sizes range from three inches to eighteen inche The licensee reported these deficiencies in the three 10 CFR 50.55(e)

reports listed below:

80-00-04 Westinghouse 3 Inch MOVs 81-00-04 Westinghouse 4 Inch MOVs 81-00-07 Westinghouse 6-18 Inch MOVs Corrective action involved open and close setting adjustments in some torque switches. Other valves were changed from torque control closure to limit control closure. In some cases, torque switch spring packs and operator gear ratios were changed. These modifications were conducted under Westinghouse (W) Field Change Notice (FCN) NAHM-10516 Valve and operator nameplates and closure tags were modified by FCN 10521A implemented by ECA 08/101358 All valves were subsequently tested in accordance with General Mechanical Test Procedure GT-M-05 and General Electrical Test Procedure GT-E-3 This problem was the subject of NRC Headquarters IE Vendor Inspection Reports (IR) 99900033/82-02 and 99900033/83-01 of the Westinghouse Elec-tric Corporation Electro-Mechanical Division (W-EMD) in Cheswick, PA in 1982 and 198 The 83-01 report identified one valve (two locations)

which was listed on the W-EMD site report but not on the Seabrook 10 CFR 50.55(e) report. This inconsistency was corrected in a revised final 10 CFR 50.55(e) report by the licensee (SBN-695 dated June 22,1984).

The inspector independently conducted field inspections of ten MOVs of various sizes and operator types. A review of ECA 08/101358A indicated that four valves, previously identified as requiring modification, had not had any changes mad Discussion with the licensee indicated that

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these four valves had their design pressure ratings reduced and therefore no longer required modificatio The field inspection also indicated that three valves had operators replaced. Certain nameplate information was missing from these valves. Discussions with licensee Startup Engi-neers confirmed that two valves (1-CBS-V51,1-CBS-V53) had their opera-tors replaced under WR CBS-0252 and CBS-0558 respectively. The missing nameplate data on 1-SI-V-77 will be stamped under WR SI-9 The inspector reviewed the relevant ECAs, FCNs and Test Data Sheet He questioned the difference in specified (TP-14) closing times of CBS-V-49 and CBS-V-53. Both are identical 6", safety injection pump suction valves. The valves were tested in accordance with GT-M-05 using TP-14 valve closing times. The licensee verified that this inconsistency had been previously documented by providing NCR 82-580A which identified the discrepancy and provided W confirmation that the stroke time was accept-abl It was noted that six valves had torque switch adjustments which did not agree with those listed in TP-1 These torque switch settings were evaluated as acceptable under ECA 99/109784A which had W concurrenc This ECA adjusted the TP-14 setting The inspector had no further question The above items are close . TMI Action Plan Requirements (NUREG 0737)

Licensee commitments in response to the requirements of the TMI Action Plan have been reviewed by the NRC staff as documented in the Safety Evaluation Report (SER). Staff evaluation of some TMI items has not been complete Further inspection of these items may be warranted in the future if the evaluation results indicate significant changes are necessary. During this

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inspection, the licensee's actions in implementing several commitments were inspected as noted below, and the following TMI action items were close I.A.1.3 Shift Manning l

l The requirements of NUREG 0737, Item I.A.1.3 were modified by a July 31, l

1980 letter from D. Eisenhut, Director of Licensing, USNRC to all power i reactor licensees and applicant These requirements were promulgated l with minor changes in Generic Letter (GL) 82-12 issued in June 1982.

l The scope of these documents include measures to prevent fatigue which could reduce the ability of the operating crew to maintain the reactor in a safe conditio The Proof and Review version of the Technical Specifications (TS) and

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the Seabrook Station Management Manual (SSMM) implement the requirements of GL 82-1 In addition the inspector reviewed the TS for conformance to GL 82-16 j

entitled "NUREG 0737 Technical Specifications" and found them to be in j full agreement.

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8 I.B.1.2 Independent Safety Engineering Group Each licensee must establish an onsite Independent Safety Engineering Group (ISEG) to perform independent reviews of plant operations. The organization and implementation of the ISEG was reviewed in NRC Region I IR 50-443/86-0 I.C.2 Shift Relief and Turnover

Procedures must be provided for shift relief and turnover to ensure that each oncoming shift is made aware of critical plant status infc+mation and system availabilit The Operations Management Manual (0PMM), Chapter 4 entitled " Shift Relief and Turnover" specifies shift relief and turnover procedure Specific-ally, it addresses the responsibilities of shift personnel, the turnover process and the checklists which are filled out for turnove The inspector reviewed Revision 0 of the OPMM. Additionally, he reviewed a memorandum from the Operations Manager on the subject of effective dates for OPMM policies. It was noted that the procedures for Chapter 4 became effective when the manual was issued in August 1985. Completed shift turnover checklists in the control room were sampled and the opera-tors were found to be familiar with their use and the requirements of the OPM d. I.C.3 Shift Supervision Responsibilities Procedures must be provided to assure that duties, responsibilities and authority of the shift supervisor and control room operators are properly defined.

, The SSMM provided general responsibilities for the Shift Superintendent (SS), Unit Shift Supervisor (USS), Supervisory Control Room Operator (SCRO), Control Room Opet ator (CRO) and Auxiliary Operators (AO).

The OPMM, Chapter 3 entitled, " Shift Operations" addresses several areas of responsibility for shift supervisio Specifically, it spells out the responsibilities of the USS with respect to plant safety, control room priorities, control room emergencies and equipment status. This responsibility is further reinforced by a memorandum from the Vice President, Nuclear Production (VP-NP), issued annuall The responsibilities of the SRO, CR0 and A0 for plant shutdown, alertaess, monitoring of plant indications, emergency actions and control manipula-tions are also covered. Specific requirements for actions on a plant trip, notification of management and regulatory agencies, control room i conduct and communications are given. .

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The inspector reviewed the above manuals and the most recent VP-NP memo and noted that although the SSMM was in its second revision (2/5/86),

it had not yet been approved by the VP-N I.C.4 Control Room Access Procedures must be provided to establish the authority and responsibil-ities of the person in charge with respect to access to the control room and establish a clear line of authority and responsibility in the control room in the event of an emergenc Access within the protected area is controlled by access authorization procedures described in Security Procedure SP2.2. This necessarily limits access into the control building. Access into the area "at the controls" within the control room is further limited to those personnel

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on official business and permission to enter must be obtained from the SCR0 or USS for all personnel other than the shift complement. These requirements are addressed in OPMM Chapter 3, paragraph 1.7 entitled

" Control Room Conduct". Lines of succession for the control room are specified in OPMM Chapter 1, paragraph 2.3 entitled " Control Room Command Function". The inspector reviewed the above documents and interviewed on-shift operators concerning access control, supervision and lines of successio I. NSSS Review of Procedures Operating license applicants are required to obtain reactor vendor review of their low power, power ascension and emergency procedures.

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The review conducted by Westinghouse Electric Corporation (W), Nuclear Operations Division was documented in April and May 1986. Following discusiion with W, the licensee resolved all comments and modified the procedures appropriately. The inspector reviewed W 1etters NAH-3038, NAH-3061 and NAH-3077 as well as the licensee response to these comments (STD86-0010, 0011 and 0014). He verified on a sampling basis that the changes agreed upon had been entered in the final approved revision of the procedures. Emergency procedure review by W will be covered by future inspection of TMI Action Plan items I.C.1 and I. I. Safety Parameter Display System This item requires all applicants to install a safety parameter display system (SPDS).

The SPDS design for Seabrook was the subject of a May, 1986 audit by the NRC Office of NRR. This audit included detailed examination of hardware and software in place and planned. The results of this audit will be published in a future Safety Evaluation Report (SER).

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10 II.F.2 Instrumentation for Detection of Inadequate Core Cooling The licensee is required to install instrumentation and develop proce-dures to provide easily interpreted indication of inadequate core cooling (ICC).

The station has installed three related systems of instrumentation to display core cooling parameter The first is the reactor vessel level instrumentation system (RVLIS)

which consists of two redundant independent trains that monitor reactor vessel water level under static and dynamic conditions. The full range reading is used during natural circulation conditions. The dynamic head reading provides indication of level based on the pressure drop across the core for any combination of Reactor Coolant Pumps (RCP). Indication is provided in the control room by two pressure indicators (one dynamic head and one full range) per train (0-120%). In addition, Train A indi-cation is provided on the MCB plasma display while Train B indication is located on the plasma display at the Shift Technical Advisor (STA)

station. Backup indication may be read by thumbwheel selection for each train on the ICC Monitors (ICCM) located behind the Nuclear Instrumenta-tion (NI) Cabinets in the Control Room (CR). The plasma displays provide system, graphics, parameter trending, diagnostic and alarm status cap-abilities as well as indicatio The second system involves the core subcooling margin (SCM) monito The subcooled margin calculations use auctioneered high quadrant average thermocouple (TC) core exit temperatures and wide range Reactor Coolant System (RCS) pressure. The subcooled margin values are routed to both the plasma displays and analog indicators on the main control boar As with the other parameters associated with ICC, SCM may be read on the ICCM as a backu The third sytem for monitoring core cooling status is the incore TC sys-tem. The system consists of 29 TCs per train whose output is directed to the ICCM where the data is directed to the plasma display, control board indicators and the main plant computer system (MPCS). The plasma display provides bulk average core exit temperature (CET), thermocouple mapping, trending, quadrant and individual element data and diagnostic capabilitie The analog displays for each train consist of a tempera-ture indicator (0-2300 F) for the third highest valid TC temperature, referred to as the hot channel CE Individual CETs may also be read on the ICCM or MPC Hardware inspection of the RVLIS was conducted and documented in NRC Region I IR 50-443/85-25. Initial inspection of procedures relating to RVCS and the ICC instrumentation indicate complete integration into the

, Seabrook Specific Emergency Procedures. These procedures will be ex-I amined in detail in future procedures inspection . - _ .

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1 II.E.3.1 Emergency Power Supply for Pressurizer Heaters

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The pressurizer heater banks and associated controls necessary to estab-lish and maintain natural circulation at hot standby conditions shall

be provided with the capability to be powered from the onsite emergency ,

power source. Additionally, they shall be connected to the emergency buses in a manner that will provide redundant power supply capabilit !

The heaters' motive and control power interfaces with the emergency buses shall be through nuclear safety-related (Class 1E) electrical device The inspector reviewed UE&C Drawings F-310013 and F-310014 and confirmed i

that the licensee has connected the group A backup pressurizer heaters j to emergency bus E-52 and the group B backup pressurizer heaters to emergency bus E-62. FSAR Section 8.3.1.1.c.7(d) was reviewed to verify that either bank of backup heaters is sufficient for natural circulation 4 at hot standby conditions. Previous NRC inspection had checked that the

. heater cables to the pressurizer were installed in accordance with the

! proper train designation and color coding. The diesel generators have

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sufficient capacity to power these heaters while supplying their maximum

, estimated accident load. Additionally, the inspector confirmed that the

! motive and control power connections to these buses are Class 1E elec-

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trical devices and were denoted as vital circuits on the UE&C Conduit

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and Cable Routing Program (CASP).

l 1 While reviewing the identification tagging of the backup heaters on the i MCB, the inspector questioned the non-safety-related color coding on the

! heater controls. Discussion with the System Test Engineer (STE) revealed that although the power and controls for the heaters are safety-related, the heaters themselves are not. Therefore the black labels are in fact

correct as installe III.A.1.2 Upgrade Emergency Support Facilities

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Each licensee is required to have an onsite Technical Support Center (TSC), Operational Support Center (OSC) and Emergency Operations Facility (EOF). These facilities were used during the Emergency Drill which was conducted on February 26, 1986. NRC Region I IR 50-443/86-10 on Emer-

gency Preparedness evaluated these facilitie Additionally, these

facilities were included in the Emergency Plan Implementation Appraisal  ;

documented in IR 85-32 and IR 86-1 ,

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i Findings: As a result of the above review of TMI Action Plan Requireme'ts, the inspectors verified that equipment, as installed, is in accordance with NUREG-0737 and licensee commitment Licensee programs inspected have been

implemented in accordance with requirements. All the above items are closed.

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12 Pre-operational Testing (PT)

1-PT-11 Containment Recirculation Sump Operability Demonstration.

The inspector reviewed 1-PT-11, Revision 2 including six Test Procedure Changes and verified that it satisfied the testing commitments outlined in  ;

the FSAR. Two inspectors witnessed Section 6 of the test which involves the combined running and sump switchover of the Train B RHR and CBS pump Test-ing activities were witnessed in the control room, RHR vault and containmen They noted that the procedure was adhered to and that an adequate number of qualified people were available during the conduct of the test. An inspector i reviewed Request for Information (RFI) 99/116158A which provided technical

' documentation for the determination that available net positive suction head (NPSH) was greater than the predicted value in FSAR Table 6.2-78.

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The inspector reviewed the operating procedures for RHR startup and CBS opera-tion and verified the calibration of the test gages in us No concerns were

identified.

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7. Training Review A review of the New Hampshire Yankee (NHY) training programs was conducte Licensed Operator Training and non-licensed Auxiliary Operator Training are conducted under the responsibility of the Training Center Manager (TCM). All

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other general and speciality training is the responsibility of the Training

Department Manager who, unlike the TCM, is on the staff of the Seabrook Sta-

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tion Manager. The inspector conducted program reviews, record reviews and

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interviews during the course of the inspectio i l He inspected each of the operator training programs conducted by the Training

, Center including a documentation review of completed courses and observation of portions of training sessions in progres General training includes lessons for maintenance technicians (mechanical,

electrical and instrument and control), radiation workers and chemistry tech-I nicians. Examples of workers requiring specialty training are crane opera-

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tors, fork lift operators and welders. The General and Specialty Training l Program was reviewed for compliance with the Final Safety Analysis Report

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(FSAR) Chapter 13 and American Nuclear Society (ANS) Standard 3.1 - 1981.

General Employee Training (GET) was verified to include Site Specific, Emer-

! gency Plan, Industrial Safety, Fire Prevention and Security Training as well

as Prenatal Radiation Guidelines. Records were reviewed for a random sample of personnel. No concerns were identified.

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8. Demonstration of Ultrasonic Examination of Cast Stainless Steel Reactor Coolant Loop Piping Welds On April 15,1986 at the Seabrook Unit 1 facility, the Westinghouse Inspection Service Group demonstrated the ultrasonic examination system and method to be used at Seabrook for preservice inspection of reactor coolant loop piping welds. The demonstration was witnessed by licensee personnel, the NRC (NRR and Region I), and NRC consultant The equipment used for the demonstration consisted of the following:

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Sonic MKI instrument

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Immersion type transducer using a liquid filled soft shoe 41 refracted longitudinal beam nominal angle, 1 inch diameter with a frequency of MHz

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Test blocks containing artificially induced flaws System sensitivity was established from side drilled holes in the Seabrook 1 calibration block SB-RC- The system successfully detected the artificial flcws contained in the Westinghouse test blocks, but a similar attempt using blocks provided by Pacific Northwest Laboratories was inconclusiv When demonstrated on Seabrook 1 Reactor Coolant Loop pipe welds, the system displayed a high noise level in one case, which may preclude a meaningful ultrasonic examination of the weld. In other cases the noise level was sig-nificantly lo;2 The system appears to be capable of detecting deep cracks in the cast material, but it is unclear that shallow cracks producing low amplitude reflections in the cast material will be found with any degree of reliabilit The consensus, after observing the demonstration and discussing the results with Westinghouse personnel, is that the technique will provide the best pos-sible examination of the cast material with the equipment currently available.

l Any requests for relief on future examination in which weld penetration is j not assured will be the subject of review by the NRC Office of NRR. The de-

! tailed results of the above demonstration will be the subject of a future report issued by NRR.

I 9. Walkdown of Piping Systems The inspector conducted a walkdown of major piping systems within the waste process building (WPB), evaluating as-constructed conditions against design detail .

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The liquid waste (WL), floor drains (WLD), boron recovery (BRS), gaseous waste (WG), waste solidification (WS), spent resin sluice (RS) and WPB airhandling (WAH) systems were traced. Specific attention was given to the systems asso-ciated with the primary drain tank degasifier (1-BRS-TK-67) including support-ing auxiliary systems such as primary component cooling (PCC), auxiliary steam (AS) and condensate (ASC) and nitrogen (NG). The waste test tanks (1-TK-63A,B) and associated components, specifically pumps, piping demineralizer, filter and valving, were inspected for consistency with the P&ID. Similar inspection was conducted on the floor drain tanks (1-WL-TK59A,B) with added examination of the floor drain filters (1-WL-F-54A,B). Small bore instrumen-tation, drain and vent lines were specifically checke The inspector con-ducted a detailed inspection of the WL evaporator (1-WL-EV-4) concentrating on instrumentation and controls, conduits, junction boxes and wiring. It was noted that construction of most systems was essentially completed although testing of several systems, as well as individual components, remained to be complete These system walkdowns revealed no safety concern . IE Information Notice 85-62 This notice provided all licensees with the backup commercial telephone num-bers of the NRC Operations Center. The inspector verified that the correct numbers were listed in the control room. This item is similar to open item 85-32-14(b) which was closed in NRC Region I IR 50-443/85-1 . Atmospheric Dump Valve Modification The design of the main steam (MS) atmospheric dump valves (MS-PV-3001, 3002, 3003 & 3004) at Seabrook was changed from electrohydraulic actuated operation and control to an air operated system with safety-related backup, high pres-sure nitrogen gas bottles. An explanation of this modification, certain de-sign details, and marked-up revisions to the FSAR were submitted to the NRC Office of NRR by the licensee (SBN-1042) on May 10,1986. During this inspec-tion, the inspector examined the in process and as-constructed modification work on MS-PV-3001 & 3004, reviewed the governing ECA design information, and evaluated both the engineering and field work in line with the FSAR commit-ments and additional information provided to NR Specifically, ECA 05/113130E was reviewed and eight other ECAs, concerned with instrument tubing and gas bottle support details, were spot-checked. The

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inspector confirmed that the modification work was being controlled by a

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Startup Test Department (STD) Work Request, WR-MS-1452 and that the safety-related sections of the work (ie: ANSI B31.1, Seismic Category 1) were being handled in accordance with the appropriate quality assurance controls (ie:

QAS-1 program). Physical configuration and component details were checked against the installation details (eg: M508116) and the control loop diagrams (eg: M506585). The inspector noted proper electrical train separation and logic for the various solenoid valves used in the design to control air flow for the operation of the atmospheric dump valve __

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While examining the support details for the safety-related, high pressure gas bottles, the inspector noted the use of high-strength bolts and weld design and quality in line with AISC design and AWS criteria. He questioned the. lack of plate washers over certain slotted holes where A-325 bolts had been in-stalled. Subsequently, the licensee issued ECA revisions for all affected supports to install plate washers over the long-slotted holes in accordance with AISC criteria for high-strength bolts. The inspector had no further questions regarding the design and construction of the subject support While the overall design modification for the air operated control system for the MS atmospheric dump valves is still under review by NRR, this inspection confirmed that installation, in process controls, and quality assurance acti-vities associated with this modification work are in line with licensee com-mitments and the appropriate codes and standards. The inspector discussed the subject work with Engineering and QA personnel and received satisfactory answers to all his question No unacceptable conditions were identifie . Limitorque Valve Motor Operator Wiring During an Environmental Qualification (EQ) audit conducted by the NRC Office of NRR during the last inspection period, it was noted that certain jumper wires in some Limitorque valve motor operators were unmarked on their short length and therefore could not be identifie In responding to the NRR audit (reference: SBN-998, dated April 10,1986), the licensee committed to evaluate all unidentified wiring and replace it, if necessar During the current inspection period, the inspector reinspected two valves within containment (CGC-V-14 & 28), containing unmarked wirin For CGC-V-28, the unidentified wires appeared to be vendor supplied because of their simi-larity to other marked vendor wires. In CGC-V-14 the unmarked jumper wires appeared to be field installe The inspector then reviewed several documents from the work packages of the subject valves to determine if the wire re-placement, in the case of the field installed material, could be traced by recor While several work requests for troubleshooting, rework and reinspection by the STD documented controlled activities with respect to CGC-V-14 & 28, none of them clearly identified the installation of the subject jumper Also, previous STD inspections of Limitorque motor operators (reference: GT-E-113)

had checked for unacceptable wiring, but had used " similarity" and not posi-tive identification, as acceptance criteri Both UE&C procedure FEP-505 (Revision 1) and the Wiring Systems Notes & Typical Details (M300230) allow for the installation of jumpers with qualified NSIS wire (code DAOP). However, for the questioned field wires, neither the wire codes, nor record of the jumper installation could be verifie .

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Subsequently, in light of both the NRC inspection / audit and IE Information Notice (IN) 86-03, discussing the EQ deficiencies in Limitorque operator wiring, the licensee issued a Nonconformance Report (NCR 82/1286C) calling for a reinspection of all Limitorque motor operators located in a harsh en-vironmen Acceptance criteria for the different wire types are provided, along with traceability requirements. The disposition to the NCR requires that any wire, either field or vendor supplied, for which identification cannot be confirmed, shall be replace The inspector discussed the licensee's corrective action program on the un-identifiable jumper wires with Engineering and Licensing personnel and with the NRR EQ Audit Team Leader. Using wire " similarity" for the acceptance of unmarked wires was considered rejected unless other documented evidence of qualification data could be provide The licensee's current prcgram for addressing IN 86-03 includes reinspection of all affected valves and replacement of unmarked wires, in accordance with the criteria of NCR 82/1286C. This corrective action is consistent with the licensee's commitment to NRR in their response to the EQ Audit (reference:

SBN-998). Such corrective measures, as necessary, apply to both vendor sup-plied and field installed wir The inspector has no further questions on this issu No violations were identifie . Management Meetings At periodic intervals during the course of this inspection, meetings were held with senior plant management to discuss the scope and findings of this in-spection. An exit meeting was conducted on May 23,1986 to discuss the in-spection findings during the period. During this inspection, the NRC inspec-tors 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 inspectio On May 22,1986, a meeting was held in the Region I Office in King of Prussia, Pennsylvania by mutual licensee / Region I agreement to discuss the Seabrook project status and schedule. New Hampshire Yankee Management personnel (see

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paragraph 1) presented information on startup activities, operational readi-

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ness, th'e current status of the Employee Allegation Resolution (EAR) program activities, follow-up of NRC open items, licensing and emergency preparedness status and organizational changes. Both the licensee and NRC management agreed that such meetings were beneficial on a periodic basis to provide for information interchange and for a consistent understanding of both regulatory developments and current project status.

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