IR 05000443/1987010
| ML20214V336 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/28/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214V312 | List: |
| References | |
| 50-443-87-10, IEIN-87-001, IEIN-87-1, NUDOCS 8706120057 | |
| Download: ML20214V336 (23) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N /87-10 Docket N License N NPF-56 Permit N CPPR-135 Priority --
Category B/C 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: March 10 - May 4, 1987 Inspectors: A. C. Corne, Senior Resident Inspector D. G. Ruscitto, R sident Inspector N. ey, a Reactor Engineer (Examiner)
Approved by- / ' '
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T. C.Els hief, Reactor Projects Section 3C 'Date Inspection Summary: Inspection on March 10 - May 4, 1987 (Report No. 50-443/87-10)
Areas Inspected: Routine inspection by two resident inspectors and one region-based operator licensing examiner of work activities, procedures, and records relative to startup testing and license issuance; post core loading heat-up, hot functional testing and cooldown; maintenance, surveillance and plant operations; and licensee event reports (LER). The inspectors also reviewed licensee action on previously identified items, including 10 CFR 50.55(e) & 21 reports and licensee actions in response to I&E Information Notices; and performed plant inspection-tours. The inspection involved 346 inspection hours by three NRC inspector Results: No violations were identified. One new unresolved item documents the need for follow-up of a generic procedural question on the positioning of RHR crossover line valves, which requires further licensee discussion with the NSSS designer-Westinghous Three additional unresolved items were opened to address questions with regard to the installation and testing of the atmospheric steam dump valves, the startup feed pump and the main steam isolation valves. All three of these separate items relate to a common programmatic issue - proper implementation of the design and design change control program. A potential concern that engi-neering interfaces between the station staff and corporate engineering personnel were not in full evidence, as required by the established licensee design control program, was raise This concern was discussed with senior NHY management and 0706120057 070601 PDR ADOCK 03000443 G PDn
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received immediate licensee attention. This inspection revealed no adverse hard-ware findings as a result of the areas reviewed and components examined. The licensee responsiveness to the documented concerns in the design control area appear to be appropriate and well directed. The disposition to the separate un-resolved items, as documented in this report, will receive separate follow-up in-spection to resolve the individual questions. With respect to the issue involving the RHR valve lineup, licensee engineering and operating personnel had been inves-tigating the problem prior to NRC inspection in this area and appear to be heading toward proper resolution. On the other three items, the inspectors will continue to monitor the effectiveness of the licensee's design and design change control program and any enhancement initiatives that are implemente ,
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DETAILS 1. Persons Contacted E. A. Brown, President, New Hampshire Yankee (NHY)
W. B. Derrickson, Senior Vice President T. C. Feigenbaum, Vice-President, Engineering and Quality Programs W. T. Hall, Regulatory Services Manager D. E. Moody, Station Manager P. M. Richardson, Training Center Manager G. S. Thomas, Vice-President, Nuclear Production J. M. Vargas, Manager of Engineering J. J. Warnock, Nuclear Quality Manager Interviews and discussions with other members of licensee and contractor man-agement, and with their staffs, were also conducted relative to the inspection of items documented in this repor . Plant Status During this report period, the plant enanged operational modes from hot standby to cold shutdown after post core load hot functional and pre-critical testing. The cooldown from Mode 3 through Mode 4 occurred on March 20, 1987 and the plant remained in cold shutdown for the remainder of the reporting '
perio Several events of minor safety significance, either reportable in accordance with 10 CFR 50.72, 50.73 or having testing / schedular impact, occurred during this inspection perio These events are documented below: March 10,1987 - Inadvertent Safety Injection A safety injection (SI) and main steam line isolation (MSLI) occurred at 3:32 a.m. when in Mode 3. All main steam isolation valves (MSIV) were initially closed. An improper tagging sequence caused the MSIV for the
"A" main steam line to open which dropped steam pressure on "A" ' - im generator (SG) sufficiently to cause an SI on low steamline pre eu. This is a rate compensated signal. This event is fully discussed in paragraph Se below under LER 87-00 March 20, 1987 - Feedwater Isolation At 12:24 a.m. while transitioning from Mode 3 to Mode 4, a high water level spike occurred on "A" SG while opening the "A" MSIV (MS-V-86) to initiate cooldown. The swell in the SG was due to a differential pres-sure of about 50 psid across the MSIV. This event is fully discussed in paragraph 5f below under LER 87-01 _ _ -
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4 April 3, 1987 - Emergency Power Sequencer Actuation t
At 11:28 a.m. While restoring vital 125 VDC bus 11A to its normal supply, the bus was de-energized due to operator error. The loss of DC power actuated the emergency power sequencer (EPS) which started the "A" emer-gency diesel generator (EDG). A preliminary review of this event and related discussion of the 125 VDC vital distribution system may be found in paragraph Sg of this report under LER 87-01 April 14, 1987 - SG Blowdown Recovery System Acid Spill
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- 55 a.m. a leak was reported from the steam generator blowdown
, system (SB) caustic fill line drain penetration. This penetration is located on the south wall of the steam generator blowdown recovery room and. drains the acid fill line back to the waste holdup sump. The leakage was identified to be coming from an overflowing sump through the pene-tration and into the yard. The inspector arrived at the scene shortly after the leak was discovered and observed licensee actions which in-cluded caustic neutralization, leak reduction and transfer of water from the sump to a temporary tank and temporary storage drums. The spill was prevented from entering a nearby storm drain. A request for engineering services (RES) was written and will be dispositioned on a priority basis to determine if modifications are required. This event is significant in that during plant power operations, the sump may be potentially con-taminated and this leakage would then represent an unmonitored release path for radioactive liquid. The inspectors will monitor this issue, as required, during future routine inspection April 16, 1987 - Inadvertent Safety Injection At 2:06 p.m. while performing reactor trip breaker (RTB) surveillance, an SI was received in "B" train when the steamline SI block switch was operated according to procedure. Preliminary licensee evaluation sus-pects faulty switch contact This event will be covered in a future inspection report. The licensee will submit an LE April 21, 1987 - Inadvertent Emergency Siren Actuation At about 1:15 p.m. a single emergency warning siren in the Town of South Hampton, New Hampshire was actuated inadvertentl The actuation was the result of an error by an instrument technician while conducting a quarterly suveillance test. The licensee notified the NRC Outy Officer (for information only) of this matter at 2:22 April 29, 1987 - Emergency Power Sequencer Actuation At 11:38 a.m. the essential 4.16 kv. bus from the "B" power train tripped during the conduct of planned relay testing. The subject relays provide second level undervoltage protection, which if actuated in conjunction with a safety injection signal, trip the normal off-site power supply n
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to the bu Upon loss of the bus, the "B" diesel generator started and commenced its sequenced loadin Initial review indicated that all sys-tems operated as designed once the essential bus tripped off lin Since the resultant trip was not part of the planned relay testing, the cause of this event remains under investigation by the licensee, although faulty test equipment has been suspected. The licensee notified the NRC Duty Officer in accordance with the four-hour reporting requirements of 10 CFR 50.72. The inspectors will conduct follow-up inspection upon issuance of the LE . Plant Inspection Tours'
The inspectors observed work activities in progress, completed work and plant status in several areas during general inspections of the plant. They ex-amined work for any obvious defects or noncompliance with regulatory require- l
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ments or license conditions. Particular note was taken of the presence of ;
quality contro12 inspectors and quality control evidence such as inspection l i records, material identification, nonconforming material identification, I l
housekeeping ~and equipment preservation. The inspectors interviewed station .
staff personnel,. craft personnel, supervision, and quality inspection person- !
l nel as such personnel were available in the work area During frequent control room observation periods, the inspectors periodically reviewed control room logs and records including night orders, shift journals, .
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shift turnover sheets, completed Repetitive Task Sheets (RTS), the temporary '
modifications log, weekly surveillance schedules and control board indication Specific note was taken of equipment in " pull-to lock" conditions, equipment tagged, alarm status and adherence to Technical Specifications (TS) Limiting Conditions for Operation (LCOs) and Action Statement During various plant inspections, the inspector reviewed certain work activi-ties in progress to determine whether controls were being implemented in ac-cordance with the noted work requests (WR) and checked certain components to verify proper equipment status in accordance with the noted design coordina-tion reports (DCR). The following items were checked:
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pressurizer relief valve replacement (WR 87W001574)
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main steam isolation valve limit switch repair (WR 87W002169)
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steam generator blowdown leak-off valves (DCR 86/00703)
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containment level transmitters (WR 87W003597 & 3598)
With regard to the inspected level transmitters, the inspector noted that the frequency of surveillance activities for checking silicone-oil level has been increased from quarterly to weekly (reference: repetitive task sheets (RTS)
87RIO2118 & 2119). Also, a request for engineering services (87-RES-0617)
has been initiated to evaluate the replacement of the existing level elements,
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which are currently " stacked" to provide level indication over the required range of height, with one piece elements. This design change would eliminate the observed weeping of silicone-oil from the threaded conduit joints of the existing " stacked" element configuration. When the design change is imple-mented, a return to a quarterly surveillance of silicone-oil fill would be recommended for ALARA and containment entry concerns. Until it is implemented, the inspector confirmed by direct visual inspection of the weeping silicone-oil situation, that the conduct of a weekly RTS is sufficient to assure the requisite level fil With respect to all plant tour inspection items, no violations were identifie . Licensee Action on Previously Identified Items (Closed) Construction Deficiency Report (CDR 84-00-06): Component Sup-ports Spanning Expansion Joints. In letters dated May 4, 1984 (SBN-651),
September 13, 1984 (SBN-713) and July 25, 1986 (SBN-1172), the licensee reported a 10 CFR 50.55(e) design deficiency involving component supports which span structural expansion joints in the containment annulus steel framing. These expansion joints, used to relieve thermal stress in the structural framing, were not accounted for in the design of the subject support The licensee, in response to the identified deficiency, instituted a complete walkdown of all areas of the plant where expansion joints had been installed in the structural members. Potential problems with com-ponents or component supports were then identified for further analysis to determine whether the expected movement could be tolerated by the design. Of a total of 487 component supports potentially affected by expansion joint interaction, 69 supports were evaluated to require modi-fication. Cable tray and conduit supports were predominantly affected (ie: a total of 54 electrical supports) with the remaining 15 suppport redesigns encompassing the I&C, HVAC and piping discipline The inspector spot-checked the Engineering Change Authorizations (a total of 30 ECAs) controlling the redesign of all affected component support He noted, in particular, that two ECAs (01/104656C & 25/11289B) directed the identification of all structural expansion joints by means of sten-ciling "EXP. JT." on the steel members in the field. This served to provide a visual aid to field engineers for any rework or new installa-tion of component supports which might occur after the completed analy-tical revie The inspector discussed the corrective action on this deficiency with licensee engineering personnel, confirming walkdowns of other areas of the plant besides the containment annulus area. He verified QC inspec-
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tion of the rework and surveillance activities of the overall corrective action. Completed engineering change packages were made available to the inspector for revie This CDR is considered close (Closed) Unresolved Items (86-09-05 & 06): Operating procedure review and consistency. NRC follow-up of specific procedural questions was documented in inspection report (IR) 443/86-36, at which time, the on-going and continued implementation of the licensee " consistency review program" for operating procedures was noted. Since that time, the lic-ensee's review of Technical Specification (TS) based surveillance proce-dures has been completed. The licensee has committed to continue the consistency review process for all operating procedures with completion expected prior to commercial operatio The overall procedural adequacy of the operating procedures is not in question, but rather certain administrative aspects of documenting re-ferences, limitations, prerequisites,' precautions and other procedural controls. The inspector has spot-checked the revised surveillance pro-cedures for the purpose of evaluating various component and system operability requirements. No significant problems were identified. The licensee is aware that continued conduct of their consistency review does not substitute for technically accurate and operationally adequate pro-cedural controls. Since the operational history from issuance of the zero power license in October,1986 through hot functional testing has revealed no major operational concerns from a procedural consistency and effectiveness view, the inspector considers these two inspection issues to be resolved and considers their closure appropriate in this repor In order to track further inspection of procedural controls for those operating evolutions which have not yet undergone the consistency review, a new inspection follow-up item has been opened. This is discussed fur-ther in paragraph 8 of this report in conjunction with a specific tech-nical question relating to the subject documented in NRC Information Notice 87-0 c. (Closed) Unresolved Item (86-27-02): Assurance that NRC commitments are tracked. The inspector reviewed the Seabrook Management Manual (SSMM),
Section SM 6.2 (Revision 3) which outlines a method for assuring that procedural requirements, based upon commitments made to NRC documents, guidance, or requests, are not inadvertently altered or removed. In-spection of the Seabrook Action Items List (SAIL), the Production Inte-grated Commitment Tracking System, and the Correspondence Action Tracking Report System revealed a complete and rigorous handling of commitments to the NRC (eg: SER items, inspection items, IE Information Notices).
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The Seabrook Nuclear Production Reporting Program Manual (NPRE) deline-ates a system for identifying commitments made with respect to incoming regulatory correspondence. Also, the Operations Management Manual (0PMM)
provides a mechanism for transmitting general information and instruc-tions to the operators on a timely basis. The inspector, during routine inspections of the control room, has noted the effective use of one of these mechanisms (ie: the Night Order Book) to communicate NRC identified issues to the operators on shif The above programs and procedures constitute licensee measures that ad-dress the generic concern regarding the tracking of NRC commitment Inspection has revealed that these measures have been effectively imple-mente This item is close (Closed) Unresolved Item (86-27-04): Condensate Storage Tank (CST) Ques-tion This two part item dealt with tank suction piping and water tem-perature control The licensee responded to Inspection Report (IR)
86-27 by letter (SBN-1202), dated September 24, 1986. The inspector reviewed the li ensee response and applicable documentation including the FSAR, P& ids and alarm response procedures. He has no further ques-tions regarding the CST suction piping design and temperature maintenance and control. This item is close e. (0 pen) Unresolved Item (86-47-01): Environmental Qualification (EQ) of Raychem Heat Shrink Tubing Splice The inspector reviewed a summary of the licensee evaluation of the Seabrook splice installation criteria and field walkdown results to determine the significance of problems identified at Seabrook relative to the concerns identified in NRC Infor-mation Notice 86-53. While bend radii less than the recommended five diameters was known to be a potential problem, the licensee summary re-vealed other identified variances from the Raychem guidance which had been evaluated. With respect to one of the variant conditions (over-heated splice sleeves), the inspector examined a sample of the field conditions to assess both the licensee field walkdown criteria and their disposition of the acceptability of the subject splice He discussed with the responsible engineering personnel the status of the continuing field walkdown effort, the closecut of request for engineering services (87-RES-0369) addressing the potentially overheated splices, and other splice configurations that were either reworked (eg: 87W002259) or authorized by specific engineering changes (eg: ECA 03/116555B).
He identified no unacceptable splice conditions during this sample in-spection of Raychem heat shrink tubing installation inside Unit 1 con-tainment. He noted that the continuing licensee field walkdowns appeared to provide a thorough coverage of the existing splices, utilizing accept-able inspection criteria. Wyle Test Reports which provide Raychem quali-fication data for the Commonwealth Edison Company have been reviewed by
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NHY engineering personnel and found to generally envelope the Seabrook design and installations. The present licensee position is that the installed Raychem splices do not present a qualification concer These Wyle Test Reports have been made available to the NRC for revie Pending completion of the NRC inspection of this data, this item remains ope (Closed) Violation (86-47-03): Normally Locked Closed Valve Found Mis-positioned. NHY responded to the December 12, 1986 Notice of Violation in letter NYN-87004 dated January 12, 1987. Verification of licensee immediate corrective actions were described in paragraph 7 of NRC Region I IR 443/86-4 Based on identified deficiencies in the locking mechanism for the open valve (CS-V-744), the licensee has modified several locking devices used throughout the plant. The inspector noted several new and innovative locking techniques in use and determined that the licensee has expanded the scope of the corrective action for this deficiency to include other system Based on inspector questions as to what constitutes a properly locked valve, the licensee has developed a definition of a valve lock for in-corporation into station operating procedures. The inspector concluded that this definition is consistent with the guidance quoted in IR 443/
86-4 The long-term corrective action as stated in the NHY response to this violation included an expansion to the " required reading" in the licensed operator and auxiliary operator (AO) training programs. The inspector reviewed this licensee program, which was subsequently modified to in-clude the' subject training in the licensed operator requalification pro-gram. He agreed that this program provided a more effective training mechanism for licensed operators on this topic. " Required reading" con-tinues as the appropriate training device for the dissemination of such information to the A0's. Based on the above corrective action, this item is close g. (Closed) 10 CFR 21 Report (87-88-01): Defective Coaxial Cable. On Feb-ruary 23,1987, GA Technologies, Inc. (GA) submitted a report to the NRC regarding defective coaxial signal cables associated with an in-contain-ment radiation detector. The subject radiation detector is a post-LOCA high range radiation monitor (HRRM) and is manufactured by Sorrento Electronics, a subsidiary of GA. The cable in question is manufactured by the Rockbestos C The NRC Office of Nuclear Reactor Regulation (NRR) requested by letter dated March 11, 1987 that NHY inform the NRC whether this problem was applicable to Seabrook. The licensee responded in a letter (NYN-87035)
dated March 18, 1987 that Rockbestos cable was not used in this HRRM
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, 10 application at Seabrook and further that insulation resistance testing of the installed ITT-Surprenant cable provided satisfactory result No technical questions remain with regard to this issue at Seabroo This 10 CFR Part 21 report is not applicable to Seabrook and is therefore close . Licensee Event Reports (LER) (Closed) LER 86-001-00: Normally Locked Closed Valve Found Mispositione This event was discussed in NRC Region I Inspection Report (IR) 443/86-47 with a Notice of Violation issued on December 12, 1986. Closure of the violation is documented in paragraph 4f of this report. With respect to the LER, the inspector reviewed the licensee's submittal (SBN-1233)
dated November 14, 1986 and had no question This LER is close (Closed) LER 87-004: Containment Equipment Hatch Airlock Valves Inoper-able. The licensee reported this event in accordance with 10 CFR 50.73 in a letter (NYN-87032) dated March 13, 1987. The valve failure was reported as an unusual event and proper notification made to NRC Head-quarters on February 11, 1987. NRC IR 443/87-08 describes the event and includes a Notice of Violation for failure to notify the Commonwealth of Massachusetts within fifteen minutes, as procedurally required, as well as certain other unresolved inspection issues. These open items and follow-up of the licensee corrective action will be the subject of future NRC inspection. During this inspection, the inspector reviewed the licensee's LER submittal and determined that it was both complete and technically accurat On April 24, 1987 NHY submitted a 10 CFR 21 report (NYN-87059) to NRC Region I on this issue. This report will be reviewed in conjunt. tion with
! the IR 87-08 open items. The 10 CFR 21 report item is numbered 87-88-02 for NRC inspection tracking purposes. With immediate follow-up of the unusual event complete and documented and the remaining technical issues being tracked as open inspection items, redundant tracking of the LER is unnecessary. This LER is close (Closed) LER 87-005: Main Control Board Indicators Not Properly Mounted, and LER 87-008: Technical Specifications Daily Log. The inspector re-viewed the NHY reports of these events, NYN-87-034 dated March 16, 1987 and NYN-87045 dated April 6, 1987, respectively. The details of LER 87-008 were previously discussed in paragraph 2h of NRC IR 443/87-0 The inspector determined both LER's to be of a routine nature, techni-cally accurate and appropriately submitted in accordance with 10 CFR 50.73. He had no specific questions on either of these events or the subsequent report Both LER's are close (Closed) LER 87-007-00: Solid State Protection System Auto Shunt Trip Test. As previously discussed in paragraph 2g of IR 443/87-02, an inad-vertent actuation of the reactor trip system occurred on March 4, 1987 during the planned conduct of reactor trip breaker (RTB) testing. The l
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inspector reviewed the LER issued on April 3, 1987 (reference: NYN-87044)
which confirmed, as was previously documented, that operator error caused the event. Because of the train-related positioning of the auto-shunt trip test switches in the back of RTB cabinets, operations personnel mistakenly believed they were testing the "B" trip test switch, when in actuality the "A" trip test switch was pressed. This caused the reactor trip since both trains of RPS logic were aligned to "A" train contro This event was caused, in part, by a human factors element to the auto-shunt trip test panel design and marking. The train-related push buttons which test the RTB shunt (trip) coils are inside the back of the RTB cabinets. When viewed from the front, this equipment has the "A" train RTB above the "B" train RTB. However, in back of the cabinets, the test sections have a left-to-right train relationsnip. Also, once the rear cabinet doors are opened, the train relationship of the test push buttons is not visibly marke Operations personnel have been cautioned about the lessons learned from this event, particularly for future RTB testing. A " Night Order" has been written to document this event for all shifts. Also, the licensee plans, as further preventive action, to place train-related nametags for the test equipment inside the rear doors of the RTB cabinet These licensee actions appear to adequately adress both the evaluation of this event and the measures taken to prevent reoccurrence. This LER is close e. (Closed) LER 87-009-00: Actuation of Engineered Safety Features. As documented in the LER issued pursuant to 10CFR50.73 on April 9, 1987 (reference: NYN-87048), an ESF actuation occurred on March 10, 1987 be-cause of an operator error relative to the removal of a main steam isolation valve (MSIV) from service. Because of the design of the pneumatic / hydraulic control system for the Seabrook MSIV's, it is neces-sary to isolate the instrument air (IA) supply to the air motor on -the valve actuator prior to deenergizing the solenoid valves providing air supply control to each MSIV. This sequence of operations was specified on the tagging order (87-643) issued on March 9,1987 to remove all four MSIV's from service in preparation for downstream repair wor dowever, with respect to the removal from service of MS-V-86, the opera-tors erroneously deenergized the two each "A" train and "B" train sole-noids prior to isolating the IA supply; thus causing the closed MSIV to commence opening. The resultant decrease in steamline pressure below the "S" signal setpoint caused the inadvertent safety injection (SI).
The source of ECCS water via the high-head injection paths to the core was the refueling water storage tank which is maintained by Technical Specifications (TS) above a 2,000 ppm boron concentration. Therefore, no zero power license condition was violated by this even _ _ ____ ________ ____ _ ___ __ ____ -
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Upon receipt of the "S" signal, all systems operated as designed. The licensee reported this event as prescribed by 10 CFR 50.72 & 73 and TS 3.5.2 and initiated timely corrective action with an operations directive j (reference: SS 29956 & 29957) specifying a two-stage tagging process for J removing the MSIV's from service in future plant evolution The inspector reviewed Station Information Report (SIR)87-027, checked the sequence of operations and plant performance and determined that all systems performed as designed. The operator error was appropriately l handled by counseling the individual operators and redirecting the sequence of tagging operations for the MSIV's. This LER is therefore close However, in reviewing the preliminary design change (DCR 0549) under evaluation as a permanent plant modification to prevent this event from recurring, the inspector noted the following inconsistencies:
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Tables 3.9(B)-23 and 6.2-83 of the Seabrook Station FSAR incorrectly specify the " fail safe" category and positions of the subject MSIV' The affected P& ids were similarly incorrec The NHY In-Service Testing (IST) Program incorrectly delineates
" fail safe" test requirements which do not reflect the pneumatic /
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hydraulic control actuator design of the subject MSIV's.
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The MSIV Logic Diagrams (NHY-503667 & 503668) indicate two particu-lar solenoid valves (FY 10A & 10B) per MSIV are "not safety-related" i although energization of these solenoids affects the closure time l of the MSIV' The MSIV surveillance procedure (0X1430.01) for the 18-month valve stroke test includes neither " fail safe" testing, nor any valve performance testing that is not initiated by a main steam isolation signal.
k The inspector confirmed by review of the applicable loop and logic dia-grams that upon initiation of a main steam isolation signal, the MSIV's will perform as designed, taking into account the single-failure cri-terion. He examined the results of the preoperational test (1-PT-13.1)
for the MSIV performance, verifying adequate test performance for the range of conditions expected of the MSIV's during normal operation. One minor administrative error in the documentation of step 6.4.19 was cor-rected by the approval of a PT Supplemental Information sheet, dated May 8, 198 It appears that the inconsistencies noted above relate more to documen-tation and commitment errors than to any concern regarding the safety-related performance of the MSIV's. Discussion with licensee engineering, technical support and licensing personnel confirmed their intent to clarify these inconsistencies in future licensing submittal However,
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the original design intent to modify the air supply solenoid valves in the MSIV actuator design, and thus preclude by design change the possi-blity of recurrence of another event similar to LER 87-009-00 has been rejected (ie: preliminary DCR 0549 was not approved). The operations staff has since revised the major plant evolution (MPE) procedures, such that additional controls are provided to account for the MSIV status during cold shutdown and prior to plant heatup. The inspector reviewed these revisions to 05 1000.01 and OS 10C0.04 and determined that adequate verification steps are provide Further hardware modification to the MSIV's (eg: the air supply solenoids),
while unnecessary for safe operation given the current procedural con-trols, would obviate any over-reliance on operator actions to maintain the valves closed, when required by Mode 3 hot standby or Mode 4 shutdown conditions. This issue merits further discussion. In conjunction with NRC review of the measures taken by the licensee to correct the MSIV documentation inconsistencies listed above, the overall subject of the MSIV design in a " fail safe" application, remains unresolved (443/87-10-01). (Closed) LER 87-010: Inadvertent Feedwater Isolation. The licensee re-ported this event on April 20, 1987 in letter NYN-87055. The cause of the feedwater isolation (FWI) was level swell in the "A" steam generator (SG) as a result of opening the "A" MSIV (MS-V-86) with too great a dif-ferential pressure across the valve. The magnitude of the level swell was sufficient to reach the P-14 setpoint causing the FWI. The inspector reviewed the LER and performed a post-event analysis of chart recordings and data logger readings. Operator actions to restore feedwater were determined to be appropriate. A notification to NRC HQ was made in accordance with 10 CFR 50.7 A related incident occurred on April 1, 1987 when in Mode 5 a FWI signal was generated because of an error in the part of a technician conducting testing on a SG level channel. No FWI occurred, nor was this ESF signal required to be in operation and the event was therefore not reportabl However, the inspector discussed both incidents with operations personnel as examples of minor operational errors causing ESF actuations. The inspector had no further technical questions and no other operational concern LER 87-010 is considered close g. (0 pen) LER 87-011: ESF Actuation - Loss of Power to Vital Bu The loss of power to the vital 4.16 Kv. bus (E-5) was the result of an operator error in restoring 125 VDC vital bus 11A to normal alignment following maintenance on battery charger "A" (EDE-BC-1A). A contributing factor to this incident appeared to be unclear operating instructions provided by a specific note in the procedure (0S 1048.01) being used to restore the normal alignment to bus 11 This LER remains open pending final inspector review of the station information report and revised procedur .
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6. Polar Crane Painting Concerns On June.4, 1986, licensee construction and QC personnel developed a plan that would allow completion of ongoing Unit 1 polar crane coating activities by the following day in accordance with the full QC inspection program. On the second shift of that day, field construction supervisory personnel determined that polar crane painting could continue without the requisite QC approval of the prepared surfaces. This decision was issued to all painting supervi-sors by the UE&C night civil superintendent on June 4, 1986. At the com-mencement of first shift work activities on June 5, 1986, UE&C QA personnel became aware of the directive (ie: a Speed Letter) waiving the requirement for QC pre-inspection of the polar crane areas to be painted. A verbal "stop work" order was issued, Corrective Action Request (CAR 220) was initiated to document a violation of QC hold points, and nonconformance reports (eg: NCR 57/5055 & 5057) were issued to specify the exact locations on the polar crane where painting had proceeded without the proper QC inspection coverag As a result of the above incident, the NRC was contacted by some workers re-garding their concerns on the polar crane painting, the subject lack of QC inspection, and the use of " paint monitors" as coating inspectors. NRC in-spection of the licensee's Employee Allegation Resolution (EAR) program files revealed several similar concerns raised by workers to EAR personne The inspector held discussions with UE&C QA personnel to clarify the sequence of events surrounding the June 4, 1986 incident and confirm that corrective action (eg: CAR 220) was in progress. A meeting was held with licensee con-struction management to verify the scope and timeliness of the required repair work on the polar crane coating, with particular emphasis upon the scheduled conduct of such rework prior to the anticipated fuel load mileston The inspector reviewed those actions taken to resolve the noted NCRs and checked that QA Level II surveillance activities had been conducted for the coating repairs, over and above the QC inspections required by the applicable struc-tural steel coating procedure, IP-103 (Revision 2).
With respect to one stated concern regarding the use of " paint monitors" and the reports filed by them, the inspector determined that the " paint monitors" did not programatically fulfill a quality function; ie - the quality control program did not take credit for the painting checks accomplished by the moni-tors. Thus, the reports filed by the monitors were construction aids and not quality documents governed by 10 CFR 50, Appendix B criteria for QA record It is noted that NRC inspection report (IR) 50-443/86-52 documents review of a similar allegation involving the role of " paint monitors" at Seabrook Station. That NRC inspection did not substantiate the allegatio During inspection 50-443/86-52, the quality of painting inside containment was also investigated. It was noted that "The Unqualified Coatings Log" was maintained to document any questionable paint conditions and thus ensure that the total area of suspect coatings did not exceed the amount which, if it totally peeled or flaked, would adversely affect the flow of water to the ECCS recirculation sump. The inspector verified that relative to the incident on
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the polar crane painting documented above, not only were several hundred square feet of coating repaired per the disposition of the applicable NCRs, but also that an additional area of polar crane coating was added to the
" Unqualified Coatings Log". This measure addresses from a safety-related engineering standpoint any concerns that might be raised regarding paint quality, event though such quality would have been inspected utilizing QC final inspection acceptance criteri Thus, while the concerns raised regarding the handling of the polar crane painting, particularly with respect to the incident of June 4, 1986, appear to have had merit, licensee corrective actions have been both comprehensive and timely. Engineering review, construction rework, QC inspection, QA sur-veillance, and EAR investigation all provided evidence that the licensee properly addressed the problem and that a quality coating was applied to the Unit 1 polar gantry cran No violations were identifie . Interpretation of " Core Alterations" NRC Region I Inspection Report (IR) 443/86-47 described discussion between NHY and the NRC inspectors regarding which components, when moved within the open reactor vessel, constitute core alteration NHY clarified the defini-tion of core alterations as follows:
" Component is further defined as any material that could alter core reactivity to significantly reduce shutdown margin or possesses suffi-cient mass to possibly challenge fuel integrity if mishandled".
The inspector concurs with this clarification in that it is specific enough to provide guidance to operators during refueling activities. With respect to questions raised during fuel loading, neither the positioning of temporary detectors, nor use of the underwater camera would be considered a core altera-tion. Movement of the dummy assembly within the reactor vessel, however, would be considered a core alteration due to its mas The inspector noted that during initial fuel loading of the Unit 1 reactor core, a conservative interpretation of " core alterations", while not program-matically defined, was in effec Inspections conducted at the time of core load revealed no problems in this area. Licensee clarification of the " core alteration" definition appears to reflect a safety conscious attitude on this issue. The inspector has no further questions in this are . Follow-up of NRC Information Notice 87-01 The subject Information Notice (IN), entitled "RHR Valve Misalignment Causes Degradation of ECCS in PWRs" was issued on January 6, 198 The Seabrook FSAR safety analysis, similar to the ECCS analyses discussed in IN 87-01, assumes that flow is provided from at least one RHR pump to all four reactor coolant system (RCS) loops during the injection phase of SI during a LOCA. The RHR
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system configuration documented in the IN corresponds to the Seabrook design, since the standard Westinghouse methodology was utilized in engineering the RHR system in its low pressure ECCS application. The Seabrook RHR crossover line valves, either of which if isolated would degrade the four-loop ECCS injection assumption, are RH-V-21 and RH-V-2 At Seabrook, previous evaluation of the proper positioning of the subject crossover line valves was initiated in response to an NRC unresolved inspec-tion item, 443/84-17-02 (closed in IR 443/85-25) and subsequently revisited in the disposition to CDR 86-00-07 (closed in IR 443/86-46). While no evi-dence that the specific concern raised in IN 87-01 was previously analyzed for Seabrook, the reviews conducted in response to earlier questions have resulted in the proper positioning of valves RH-V-21 & 22 for their ECCS con-figuration while the plant is maintained in Modes 1, 2, and 3. Additionally, the ECCS subsystems, as specified in the Seabrook TS, are not required to be operable in Modes 5 and 6. Therefore, the only situation that remains to be addressed is the question of the proper ECCS configuration of the crossover valves (or other ECCS valves whose isolation would violate the RCS loop in-jection analyses) while in Mode The inspector noted that for routine Mode 4 operation with one RHR train operating, the other RHR train would be operable and lined up in an ECCS con-figuration. While the Technical Specifications (TS) do not preclude Mode 4 operation with an RHR loop out of service, the Seabrook Station operating procedures were revised on April 20, 1987 to ensure that a shutdown RHR train in Mode 4 must be aligned for ECCS operation. However, compliance with TS 3.5.3.1.d for an operable ECCS flow path in Mode 4 is subject to interpreta-tion since the operating procedures (ie: 0S1013.03 & 1013.04) require that the crossover valve associated with a particular RHR train in operation be closed. Thus, if RHR Train "A" is operating in Mode 4, RH-V-22 is procedur-ally closed; and likewise for RHR Train "B" and RH-V-21. This condition would not provide for an operable RHR flowpath to all four RCS loops without manual realignment of the isolated crossover valve. Therefore, the proper ECCS valve lineup for Mode 4 operation remains in questio The inspector reviewed a NHY internal memorandum (ISEG R87-01-002) prepared by the Independent Safety Engineering Group on April 9, 1987 which discussed the need for NHY Engineering to perform an analysis to determine satisfactory ECCS configuration during Mode 4 operations. The Westinghouse Owners Group (WOG) is currently evaluating LOCA conditions during Modes 3 and 4 to address the problem generically and provide guidance as to the correct ECCS lineup in these mode The ISEG has recommended that the normal RHR shutdown cooling configuration be considered and evaluated with respect to the problems raised by closing either crossover valve (RH-V-21 or 22) during Mode 4 operatio Pending final resolution of the concerns presented in IN 87-01 and final dis-position as to the correct lineup for ECCS operability in Mode 4 at Seabrook, this issue is unresolved (443/87-10-02). The inspectors will continue to
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routinely evalute other operating procedures with respect to both technical bases, as was the case with the IN 87-01 concern, and consistency i'tues (reference paragraph 4.b).
9. Maintenance Activities Steam Generator Safety Valves Numerous steam generator (SG) safety valves (SV) experienced seat leakage during hot operations. Following cooldown, the licensee instituted an extensive program for SV rework. Valves were removed and the seats were lapped. Following successful leak tests the valves were reinstalle The inspector periodically monitored licensee progress, observing both work and test activities. He had no questions on the conduct of this SV maintenanc No violations were identifie Battery Chargers On February 2,1987 battery charger "A" (EDE-BC-1A) failed and was re-placed by the portable charger (EDE-BC-1P). On February 10, 1987 an RES (87-185) was written by QC personnel questioning the installation of the portable charger with respect to the wiring, installation procedure and operability. It was subsequently decided that a QA Inspection Report (QAIR) was a more appropriate mechanism to document the noted QC concerns and QAIR 87-148 was initiated on February 23, 198 Since during this February timeframe the portable charger was being used to satisfy T.S. 3.8.2.1 (D.C. Sources) operability requirements, the NRC inspector verified that the battery charger capacity test had been com-pleted on the portable charger within the 18 month periodicity. He also independently questioned the operability of the portable charger with respect to seismic and IE criteri The response to the QAIR was completed on March 31, 198 This response indicated that minor procedural changes were appropriat Additionally, torquing requirements of 30-45 ft-lbs were identified for the terminal bolt connections. Based upon the licensee belief that the " snug tight" practice used to install the bolts was equivalent to about 30 ft-lbs, it was decided that there was no need to retorque the bolts. The in-spector questioned this decision and the lack of a detailed technical evaluation to determine whether the " snug tight" bolt condition was ade-quate from a quantitative standpoint. Additionally, the delay in dis-positioning the QAIR (5 weeks) provided an example of potential timeli-ness problems in the handling of QAIR responses. The repaired battery charger (BC-1A) was returned to service on April 3 without the changes mandated by the QAIR response incorporated into the installation proce-dure (0S 1048.05). However, the system engineer did incorporate the required torque criteria into the work reques s
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. Even though no safety related problems were identified with the final hardware in the inspection of this battery charger replacement and sub-sequent reinstallation, the performance of a safety-related maintenance activity with known corrective action still pending is considered a questionable practice. Therefore, the inspector discussed the handling of this work in a meeting with cognizant personnel from the various lic-ensee departments. NRC concerns were re-iterated and the licensee indi-cated their intent to improve both the timeliness and the technical re-sponse to QAIR issues by establishing:
(1) a method to ensure that QAIRs with operability or safety signifi-cance are given a prioritized dispositio (2) a mechanism to differentiate between QA inspector questions and specific criteria deviations so that answering lower priority in-quiries does not interfere with the disposition of higher priority item (3) measures to require that a formal technical evaluation be conducted in those cases where inspection has not verified that engineering criteria have been me The inspector believes that licensee responsiveness to the NRC concerns are addressed by the above initiatives. Their implementation and the conduct of similar maintenance activities, where QA coverage results in inspection questions, will be the subject of future NRC routine inspec-tion For this particular case involving battery charger BC-1A, no actual hardware deficiencies were noted. It was noted that improvements in the work and QA interface controls are planned. No violations were identifie c. Atmospheric Steam Dump Valve (MS-PV-3002)
The ASDV for the "B" S/G was reworked following leakage identified during pre-critical HFT. The inspector observed disassembly of the valve, its operator and the affected seismic supports. He reviewed the maintenance procedure and work packages being ut.ilized in the conduct of this acti-vit He specifically noted good work practices in the areas of clean-liness control of open system connections, the temporary support of dis-assembled components and the secure storage of parts and tools when not in use. No violations were identifie d. Reactor Protection System The inspector reviewed licensee action on a reactor protection system (RPS) problem first discovered at another nuclear facility. The tech-nical issue related to the use of an ungrounded test signal lead for instrument calibrations. Discussion with the I&C Department Supervisor revealed that NHY was aware of the concern and had initiated a priority
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"E" RES to resolve the proble The inspector concurred with the high prioritization of this RES since this issue must be resolved prior to entry of the plant into Mode 2. He had no further questions at this time and no violations were identifie . Design Control - Component Modifications Atmospheric Steam Dump Valves (ASDV)
l During the conduct of startup testing (ST-55) for the steam dump system, it was discovered that the ASDVs could be opened from the remote safe shutdown panels (RSSP) with the selector switch in the " remote" positio By design, however, control from the RSSP should be disabled in the
" remote" mode. Additionally, the valves could be opened from the main control board (MCB) with the selector switch in the " local" positio This situation was also contrary to the design criteria for the " local /
remote" control of the ASDV Test exceptions were written to document these nonconforming test results and a request for engineering services (RES) was written to resolve both questionable conditions. The response to the RES (87-0191) indicated that procedural and design changes were required to correct the situatio Design coordination report (DCR) 87-0064 was written to address valve behavior while in " remote" operation. A change to abnormal operating procedure 051200.02, Safe Shutdown and Cooldown from Remote Safe Shutdown Facilities, was initiated for operation in " local" contro Inspector
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review of the disposition to RES 87-0191 determined that these design and procedure changes were technically adequate. ST-55 was subsequently completed with satisfactory result While the licensee actions noted above appear to address the technical issues, the inspector expressed concern with the method by which DCR 87-0064 and a previous ASDV modification had been tested. The circuitry which had been inadequately designed, as evidenced by the ST-55 test exceptions, was subsequently modified in June, 1986 as part of a design change to enhance the safe shutdown capabilities of the plant. Design
! change notice (DCN) 65/02728 provided safety grade, manual disabling switches, a safety grade backup air supply and dual train controls for the ASDVs to satisfy 10 CFR 50, Appendix R safe shutdown requirements for potential fires in the essential switchgear rooms. Engineering change authorizations (ECA) 05/112599 and 03/804116 were issued in May and June 1986 respectively to implement these change The inspector reviewed the work requests written to install the new switches and conduct the retesting. He noted that the only retesting specified was general electrical test GT-E-21," Wiring Verification and Functional Checks". Wiring verification includes continuity checks, wiring termination verifications and megger tests, while the functional checks verify correct operation of each electrical device. Although the note on step 6.2.2 of GT-E-21 states, " Verify the operability of the
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20 circuit to function in accordance with the requirements of the intended application", the subject ASDV devices were only tested individually and !
not as a complete circuit. Since each individual device was satisfac- l torily tested by GT-E-21, a " sneak path" in the circuit was identified during ST-55, as discussed above, and not by any planned test of the ECA modifications to the ASDV The licensee stated that it was their intent to check the overall circuit operation during the conduct of ST-55, which had been modified to include ASDV testin However, following a detailed review of the applicable design documents, the inspector raised certain questions in regard to this licensee position and discussed them with licensee engineering and operations personnel. Specifically, as of the end of this report period, it had not been clearly established that the ASDV individual train con-trols had been dynamically tested or that the ASDV " position-maintained" feature had been fully demonstrate Pending additional licensee response to these questiuns and NRC review of these issues, this item remains unresolved (443/87-10-03). Startup Feed Pump The startup feed pump (SUFP) was pre-operationally tested during hot functional testing (HFT) in September 1985. A low suction pressure trip was added to the control circuitry in June, 1986. During pre-critical HFT in February,1987, it was discovered that in certain suction lineups, the pump would trip upon starting due to a momentary dip in suction pres-sur As a result, a design change was implemented to add a time delay in the low suction pressure trip circuitr Initial review of these design changes by the inspector raised several concern Specifically, the adequacy of the change which installed the low suction pressure trip was questioned with respect to what evaluation was per-formed to predict impact of the trip on the system. Additionally, it was not clear that sufficient post-modification testing was conducted following implementation of the original change because the effect of
, the pressure dip had not been identified. Finally, review of the design l change which installed the time delays identified further questions on l the suction piping configuration to be used in testing the time delays l
after installation. The resolution of these NRC questions is required for a determination of the operability of the SUFP after implementation of the current modification and subsequent testing. This issue remains unresolved, pending licensee response to the NRC questions and review of their impact upon SUFP operability (443/87-10-04).
11. Licensed Operator Training Program An inspection to evaluate the effectiveness of the licensed operator training programs was conducted by reviewing the training given the operators to pre-vent or mitigate the effects of abnormal events at the facility. The inspec-
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tor selected for review events relating to the inadvertent initiations of the safety injection and engineered safety systems and problems with NRC notifi-cations following equipment inoperability or inadvertent actuation of the control building air system. Discussions were held with members of the training staff and the following documents were reviewed as part of the in-spection:
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Lesson Plans for the Licensed Operator Requalification Training Program-
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Lesson Plans for the Replacement Operator License Training Program
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Lesson Plans for the simulator portion of the Replacement Operator License Training Program
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Training records of four operators
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Segment examinations for four operators
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Simulator Assessment Exam Report Forms for four operators
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Simulator Examination Report Forms for four operators
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Segment Examination Grade Summary Sheets
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Training System Development Overview Procedure
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Licensed Operator Requalification Training Program Description
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Conduct of Training Analysis
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Evaluation of Training Effectiveness (STC-2012E)
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Report of 1985 Licensed Operator Requalification Training Program
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Licensed Operator Requalification Training Program Evaluation Report (December, 1986)
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Meeting Minutes, Curriculum Development Committee (CDC) January, 1987
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Training Program Evaluation Matrix
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Schedule for 1987 Requalification Operator License Training Program
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Report of Special Task Force Evaluation of Switch Misoperation The Licensed Operator Requalification Training Program consists of five or six phases each year. Each phase consists of a one-week training session, consisting of both classroom instruction and simulator training, for each of the six operating crews. At the end of each week, written and simulator ex-aminations are administered. The written examination covers material which
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was taught in the class room during the preceding week. The simulator ex-amination is either a demonstrative examination or an accident assessment evaluation. The licensee intends to submit these training programs for ac-creditation by the Institute of Nuclear Power Operations (INPO) in March, 198 Proper operation of the safety injection system was taught in Phases 85-1 and 86-2 of the Licensed Operator Requalification Training Program. A question on this subject was included on the weekly examination, given both in 1985 and 1986. The inspector noted improved performance in the 1986 examination in this area by one particular operator who had not responded weli to the 1985 testing. Also, proper operation of the engineered safety system was taught and evaluated in Phase 86-5 and operators were provided further instruction during the Replacement Operator Training Program. The inspector's evaluation of this training resulted in the identification of no training deficiencies which appear to have been contributing factors in the abnormal events selected for review. The inspector noted that this same conclusion was reached in a licensee prepared report (" Report of Special Task Force Evaluation of Switch Misoperation") on this subjec Additionally, the inspector examined other aspects of the Licensed Operator and Replacement Operator Training Programs, to include instruction on how to make proper NRC notifications (eg: December, 1986 simulator training) and on changes to the Control Building Air system (eg: Phase 86-5 training). The inspector's review of the training process revealed that the performance of individual students in both the Replacement and Licensed Training Programs is closely monitore Performance weaknesses appear to be identified early and are addressed by a performance review board. The review board identifies the cause of perform-ance problems and recommends corrective actions, the completion of which is then documented in the individual's training record. The inspector noted that Seabrook Station has never been forced to remove an operator from licensed duties due to a performance proble The inspector also reviewed simulator examination forms, which document in-structors' evaluations of the performance of operators during a simulator examinatio The forms sampled by the inspector were found to contain cri-tical and meaningful evaluations of the operators' performance. Accident assessment evaluations are documented by operators' written responses to questions concerning transients conducted during training in the simulato These evaluations evidenced continued development and appeared to be innova-tive in natur While segment examinations are designed to test mastery of the learning ob-jectives for each phase, they examine only the information presented in the classroom portion of the training during the previous week. Poor performance in an area may result in the information being included in the next requali-fication cycle. The inspector noted that the six examinations given over each phase appear to have some duplication of questions, but that questions are of high quality and provide examination to a sufficient depth of knowledg re
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j 23 The average grades over. the six week period are similar, which is one indica-tion that the examination is not being compromised over the six weeks of the phase. The examination grades also provided evidence that the operators are retaining the information being taught. Since the segment examinations only evaluate information which has been included in the requalification program, the inspector questioned the need to identify other. areas in which retraining is required to upgrade licensed operators. The licensee stated that other-diagnostic tools provided sufficient information about areas that required retraining without an annual written examination. The examinations that have been given were conducted in accordance with the NRC approved training proce-dure The inspector also examined the role of the Curriculum Development Committee (CDC), which is comprised of management personnel from both the Operations and Training departments, in deciding in what areas training will be given
'during the next requalification cycle. Proposed areas of training are de-veloped from instructor observations in the simulator, trainee feedback evaluation forms, supervisory feedback evaluation forms, past plant events, upcoming plant evolutions, NRC and licensee written test results, operator performance evaluations and training deficiency reports. The inspector re-viewed the CDC meeting minutes and verified input from these varied sources of informatio The inspector confirmed that the Seabrook training program has been conducted in accordance with the NRC requirements and has received extensive licensee management involvement. The training provided to prevent the occurrence of the abnormal events appears adequate and inspection confirmed the effective-ness of segment examinations to evaluate the retention of information taught during the preceding week. The inspector noted that the development of future requalification training schedules is a dynamic process which receives direct management oversight at Seabrook Statio He has no further questions on the licensed operator training program. 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 deviation Unresolved items disclosed during this inspection are discussed in paragraphs Se, 8, 10a and . 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 12, 1987 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 inspection.