IR 05000443/1996011
| ML20147E478 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/07/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20147E475 | List: |
| References | |
| 50-443-96-11, NUDOCS 9703140003 | |
| Download: ML20147E478 (30) | |
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' U. S. NUCLEAR REGULATORY COMMISSION-
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REGION I
i Docket No.:
50-443 License No.:
NPF-86 Report No.:
50-443/96-11 i
Licensee:
North Atlantic Energy Service Corporation i
Facility:
Seabrook Generating Station, Unit 1
' Location:
Post Office Box 300 l-Seabrook, New Hampshire 03874
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i Dates:
November 30,1996 - January 13,1997 i
I Inspectors:
John B. Macdonald, Senior Resident inspector David J. Mannai, Resident inspector i
Leonard Prividy, Reactor Engineer, Region I
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Joseph Colaccino, Mechanical Engineering Branch, NRR Mark Holbrook, Contractor, Idaho National Engineering Laboratories Accompanied by:-
Javier Brand, Resident inspector Intern j
Approved by:
John Rogge, Chief, Projects Branch 8 Division of Reactor Projects i
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9703140003 970307 i
PDR ADOCK 05')O0443 G
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EXECUTIVE SUMMARY Seabrook Generating Station, Unit 1 NRC Inspection Report 50-443/96-11 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 7-week period of resident inspection; in addition, it includes the results of the Generic Letter (GL) 89-10 motor operated valve
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l inspection conducted between September 16 and November 15,1996.
Operatio'ns:
e initial responses to design-bases discrepancies identified in the emergency feedwater systera have been appropriate. Required notifications were timely and accurate. Operabil;ty evaluations properly referenced design-end licensing-bases j
information. Additionally, probabilistic risk assessment speciahst review which
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determined that the design-bases discrepancies were of negligible risk consoquence was a good initiative. (Section O2)
Maintenance:
Licensee Event Report, LER 96-05, " Missed Surveillance Requirement," properly
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addressed the reporting criteria required by 10 CFR 50.73. The failure to implement
~ an accelerated surveillance frequency for the "D" safety injection accumulator fill -
i and drain isolation valve in accordance with in-service test (IST) program requirements for component performance in the alert range has been dispositioned.a non-cited licensee identified violation, consistent with Section Vll.B.1 of the NRC
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Enforcement Policy. Licensee personnel identified this event and initiated prompt <
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and effective corrective actions. (Section M1.1)
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New fuel receipt inspection activities were performed in a carefully controlled manner according to station procedures with excellent coordination between involved personnel. (Section M1.2)
The licensee addressed work control and maintenance restoration aspects of a previously identified NRC unresolved item regarding a temporary chainfall hoist that had been left installed following completion of a work activity. However, the item remains open pending licensee evaluation of follow-up inspector question concerning potential seismic interaction between the chainfall and the containment sump recirculation suction valve encapsulation tank cover that it was attached to.
(Section M8.1)
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Enaineenna:
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More information will be needed to close NRC review of the design assumptions j
used by North Atlantic in the Seabrook MOV program. (Sections E1.2 and E1.3)
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Notwithstanding the positive measures taken by North Atlantic to assure design-basis capability, including an acceptable resolution of a degraded voltage issue, assumptions regarding valve f actor, stem friction coefficient, and load sensitive behavior were not fully justified for untested MOVs. -(Section E1.4)
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North Atlantic established and implemented an adequate MOV trending program as recommended by GL 89-10. (Section E1.5)
- North Atlantic committed to complete a major revision of MOV program procedure
ES1850.OO3 by January 31,1997, in response to inconsatencies between current q
practices and other program descriptions. However, the resident inspectors have y
determined that-the program revision had not been completed prior to issuance of-
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this inspection report. (Section E1.6)
Two followup items were opened regarding a change to the GL 89-10 program
scope and technical justifications for certain motor-oprated valve design assumptions. (Sections E1.2 and E1.3) An unresolved item concerning motor -
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operated valve grouping criteria was closed. (Section E8.7)
Plant Support:
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Radiological protection program controls were observed to have been properly I
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. implemented. Workers were noted to be complying with established procedures
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and controls. (Section R1.1)
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4 A fire protection program surveillance.was effectively performed. involved a
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individuals were knowledgeable of assigned responsibilities, good. communications
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Assurance of Quality i
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A licensee task force that was formed to verify or restore the design-bases configuration control of pressure transmitters installed in the station continued to -
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l perform effectively. Identified discrepancies were properly documented and t
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evaluated. (Section 07.1)
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TABLE OF CONTENTS
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EX E C U TI VE S U M M A RY............................................., ii i
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TA B LE O F CO NT ENTS.............................................. iv
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~l. Operations
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Conduct of Operations......................................... 1 01.1 ' General' Comments (71707)................................ 1 l~
Operational Status of Facilities and Equipment........................ 1 l
02.1 Feedwater System Pipe Break Analysis........................ 2
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O2.2 Emergency Feedwater Auto-Start Design Deficiency
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Quality Assurance in Operations................................. 4
~ 07.1 - '(Update) Transmitter Configuration Control (IFl 50-443/96-10-01)..... 4 ll M ainte n a nce.................................................. 6 M1 Conduct of Maintenance
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, M1.1.(Closed) LER 96-005-00,. Missed Surveillance Requirement.......... 6 l-
.M 1.2 New Fuel Receipt Inspection................................ 7
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i M8.1-(Update) URI 50-443/96-02-02,.Chainfall Holst Attached to Safety-l Related Equipment
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Ill. Eng ine ering................................................... 9
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El Conduct of Engineering
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E1.1 Generic Letter 89-10 Motor-Operated Valve Program Review (Tl 2515/109)
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E1.2 Summary Status of Generic Letter 89-10 MOVs.................. 9 l
E1.3 MOV Sizing and Switch Settings............................ 12 E1.4 Design-Basis Capability
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E1.5 MOV Failures, Corrective Actions, and Performance Trending
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E1.6 MOV Program Administration.............................. 17 E7 Quality Assurance in Engineering Activities......................... 18 L
E8 Miscellaneous Engineering issues
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E8.1.(Closed) Inspection Report 50-443/91-81, Section 2.7............ 18 E8.2 (Closed) Inspection Report 50-443/91-81, Section 2.10........... 19
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j E8.3 (Closed) Inspection Report 50-443/91-81, Section 3.0............ 19
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E8.4 (Closed) Inspection Report 50-443/91-81, Section 2.2............ 19 E8.5 (Closed) Inspection Report 50-443/91-81, Section 2.4............ 19 E8.6 (Closed) Inspection Report 50-443/91-81, Section 2.4............ 20 E8.7 (Closed) Unresolved Item 50-443/94-11-02......
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E8.8 (Closed) Inspection Report 50-443/91-81, Section 2.5............ 20 E8.9 (Closed) inspection Report 50-443/91-81, Section 2.6............ 20 E8.10 (Closed) inspection Report 50-443/91-81, Section 2.7............ 21 E8.11 Review of Updated Final Safety Analysis Report (UFSAR)
C o m m itm e nt s......................................... 21 IV. Plant Support
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R1 Radiological Protection and Chemistry Controls
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R 1.1 General Comments
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S1 Conduct of Security and Safeguards Activities....................... 22 S 1.1 General Comment (71707, 71750).......................... 22 F1 Control of Fire Protection Activities.
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F1.1 Fire Detection Equipment Surveillance.........
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Manageme nt Meetings.............................
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X1 Exit M e eting Sum m a ry........................................ 23 PARTIAL LIST OF PERSONS CONTACTED....................
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INSPECTION PROCEDURES USED..................................... 25 LIST O F ACRO NYM S U SED.........................
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Report Details Summarv of Plant Status The facility operated at approximately 100% of rated therm 6! power throughout the
' inspection period with routine minor power reductions performad to suppc.1 instrument calibrations and turbine valve testing. On December 12,1996, a high differential pressure signal on the "C" circulating water (CW) travelling screen initiated automatic closure of the associated CW pump discharge valve. Operators tripped the "C" CW pump and reduced reactor power to approximately 86% to maintain CW differential temperature discharge limits. The differential pressure indication was determined to be due to instrument malfunction. Corrective actions were completed and the reactor was returned 100% of
rated thermal power on December 13,1996. The primary to secondary coolant leakage rate remained less than 0.05 gallons per day throughout the inspection report period.
l. Operations
Conduct of Operations 01.1 General Comments (71707)
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Using inspectior. Procedure 71707, the inspectors conducted frequent reviews of ongoing pla% operations, in general, routine operations were performed in accordance with station procedures and plant evolutions were completed in a deliberate manner with clear communications and effective oversight by shift s
supervision. Control room logs accurately reflected plant activities and observed -
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shift turnovers were comprehensive and thoroughly addressed questions posed by a
the oncoming crew. Control room operators displayed good questioning perspectives prior to releasing work activities for field implementation. The inspectors found that operators were knowledgeable of plant and system status.
O2 Operational Status of Facilities and Equipment a.
Insnection Scoce (71707,37551)
On October 9,1996, the NRC issued a letter to power reactor licensees titled
" Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Adequacy and Availability of Design Basis Information." The letter requires licensees to provide information to the NRC within 120 days of receipt that; describes processes that establish, control, and maintain design-bases and configuration control; provides the rationale for concluding design-bases requirements are translated into operating, maintenance, and test procedures; provides the rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases; describes processes for the identification of problems and
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implementation of corrective actions, including actions to determine the extent of l
problems, actions to prevent recurrente, and reporting to the NRC; and, assesses l
the overall effectiveness of current pro: esses and programs in concluding that plant
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configuration is consistent with the design bases.
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The resident inspector staff discussed response status with engineering department management, attended several plant management project meetings, and evaluated the technical adequacy of the initial corrective actions to several discrepancies identified during the licensee letter review and response. Two discrepancies identified in the design-bases review of the emergency feedwater system (EFW)
were selected for in-depth inspection. _The inspectors reviewed applicable sections of the Updated Final Safety Analysis Report (UFSAR), Technical Specifications (TS),
station procedures, potential implications for support programs such as licensed m
operator training, and subsequent operability and safety evaluations.
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02.1 Feedwater System Pipe Break Analysis b.
Observations and Findinas On December 6,1996, during the development of design coordination report, DCR 96-034, the licensee determined that a non-conservative time assumption of initial EFW flow to the intact steam generators (SGs) following a feedwater line break accident existed in UFSAR Section 15.2.8, "Feedwater System Pipe Break." The UFSAR assumed delivery of the minimum required EFW flow of 470 gallons per minute (gpm) to the intact SGs within 77 seconds of receipt of a low-low SG level signal. This was based on a 2 second instrument loop process time plus 75 seconds forthe steam turbine driven EFW pump to achieve rated speed, flow, and discharge pressure. However, the existing analysis in the UFSAR and supporting calculations did not consider the 20-second stroke time for the automatic closure of the control and isolation valve to isolate a faulted SG upon receipt of a feedwater.
.line break signal. Design bases flow to the intact SGs can not be achieved in the,.
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most limiting condition of single EFW pump operation until an assumed faulted SG is
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isolated. The licensee made a one-hour report to the NRC Operations Center (NRC OC)in accordance with the requhments of 10 CFR 50.72 for a condition outside the design basis of the plant. Additionally, ACR 96-1337 was initiated to document this issue, and a preliminary operability evaluation was performed.
The licensee procured the services of the Yankee Atomic Electric Company (YAEC)
Transient Analysis Group (TAG) to evaluate this condition. The TAG reviewed the current Seabrook analyses for feedwater and steam line breaks and performed a sensitivity case demonstration in which a beyond design bases failure of an EFW control and isolation valve in the closed position to an intact SG was assumed in.
addition to the existing design-bases limiting conditions. As documented in YAEC memo, SBP 96-122, dated December 6,1996, the reviews concluded that for either a feedwater or steam line break that the previously unanalyzed 20-second time delay for the isolation of the faulted SG had a negligible impact on the existing UFSAR accident analyses.
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.C_o_nclusions The inspectors reviewed UFSAR Section 15.2.8, ACR 96-1337, associated YAEC memorandum, and discussed the issues with licensee engineering management.
The inspectors conclud,d the license effectively identified the existence of a non-t
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conformance in the UFSAR regarding feedwater line break analysis. Immediate I
responses were appropriate. Proper NRC notifications were accomplished. initial
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operability evaluations were performed and design-bases calculations and analyses were properly used in the evaluation of this issue. Additionally, affected documents and calculations were revised to reflect the EFW control and isolation valve 20-second closure time following receipt of a feedwater line break signal.
02.2 Emergency Feedwater Auto-Start Design Deficiency b.
Observations and Findinas On December 12,1996, during the 10 CFR 50.54(f) review effort, the licensee identified a scenario that would preclude the automatic initiation of the EFW system. The EFW system consists of a motor-driven (MDEFW) pump and a steam turbine-driven (TDEFW) pump. Automatic initiation of the MDEFW pump is actuated by the "B" train of the solid state protection system (SSPS). Steam is provided to the TDEFW pump from the "A" SG through main steam valve, MS-V-393, whose automatic initiation is actuated by the "A" SSPS train. Steam is also provided to the TDEFW pump from the "B" SG through main steam valve, MS-V 394, whose automatic initiation is actuated by the "B" SSPS train. The steam lines from the
"A" and "B" SGs discharge to a common steam supply line to the TDEFW pump.
Main steam valve, MS-V-395, located in the common steam supply line receives its automatic initiation actuation signal from both the "A" and "B" SSPS trains. The specific automatic initiation failure scenario. involves a feedwater or steam line break
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in the "A" SG in conjunction with a failure of the "B" SSPS train, such that the MDEFW pump will not automatically start and steam admission to the TDEFW will not automatically occur because the "A" SG would be faulted and the failure of the
"B" SSPS would prevent the automatic opening of MS-V-394, thereby precluding the "B" SG from supplying steam to the TDEFW pump. The licensee made a one-hour report to the NRC Operations Center (NRC OC) in accordance with the requirements of 10 CFR 50.72 for a condition outside the design basis of the plant.
Additionally, ACR 961338 was initiated to document this issue and a preliminary operability evaluation was performed.
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The scenario identified by the review team was inconsistent with the EFW system design bases as described in UFSAR Section 6.8.1.h which states that the EFW.
system is capable of autornatically initiating flow upon receipt of an actuation signal. Additionally, UFSAR Section 15.2.8 assumes that one or both EFW pumps willinitiate and supply EFW flow to at least two SGs within 96 seconds of the receipt of an actuation signal. However, within the scenario identified by the review team, operator action would be necessary to actuate the EFW system. The licensee operability evaluation concluded that existing emergency operating procedure (EOP) direction would ensure that control room operators would identify that the EFW system did not automatically initiate, and would take appropriate actions to manually actuate the system within ten minutes. Actuation within ten minutes is necessary to ensure existing accident analysis parameters of; maintaining adequate inventory in the intact SGs; avoiding the pressurizer from becoming solid I
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with water; and, maintaining reactor coolant system decay heat removal capability.
Licensee review of past simulator performance and in-plant EOP usage indicate operators typically complete the verification of EFW operation within approximately four minutes. Currently the licensee is developing a design change that would address this design vulnerability.
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_ Conclusions The inspectors reviewed UFSAR Sections 6.8.1.h and 15.2.8, as well as ACR 96-1378 and the associated operability evaluation. Additionally, the inspectors discussed this scenario with engineering department management, on-line maintenance personnel, and probabilistic risk assessment engineers. Initially, the licensee immediate actions were appropriate. The NRC 00 was properly notified and an operability evaluation was initiated. The inspectors questioned if additional controls had been placed on the removal of the MDEFW pump for on-line maintenance purposes and if the potential for TDEFW overspeed and trip during manual actuation had been considered. The licensee indicated that the minimal incrementalincrease in the contribution to core damage frequency from this scenario would not have uniquely warranted increased on-line maintenance controls but that the MDEFW would not be removed from service for purely elective maintenance until the start of the upcoming refueling outage. Additionally, the licensee concluded that the existing nonsafety-related heat tracing of the TDEFW steam supply lines was of sufficient capacity to ensure saturated conditions were maintained within the supply piping until manual actuation of the TDEFW pump would be accomplished. At the conclusion of the report period,.the licensee was-J developing surveillance criteria for the heat tracing.
The immediate licensee corrective actions to the above design-bases discrepancies have been appropriate. Final correction of the design vulnerability will be tracked as an inspector followup item. (IFl 50-443/96-11-01)
Quality Assurance in Operations 07.1 (Update) Transmitter Configuration Control (IFI 50-443/96-10-01)
b.
Observations and Findinas NRC Inspection Report 50-443/96-10 documented that during an independent inspection of the service water (SW) pump house, the inspectors noted that plastic foreign material exclusion (FME) shipping caps were installed in the spare conduit j
ports on several SW system pressure transmitters. This NRC observation was l
brought to the attention of plant management. ACR 96-1002 was generated and a
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dedicated task force was established to perform an inspection and issue a report.
The ACR evaluation was expanded by the licensee to inspect and verify or establish as necessary the proper configuration of a population of approximately 551 currently installed process transmitters (Rosemount and Foxborough). The licensee
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concluded this action was appropriate due to the diversity of transmitter models and designs that could be installed in various service applications.
The inspectors have met with the task force leadership on several occasions to verify project scope, assurance of quality, incorporation of potential lessons learned into proper program processes, and to maintain status of inspection progress and observations. At the conclusion of this inspection report period, the task force had completed approximately 378 transmitter field inspections. Essentially all of the harsh environment transmitters, with the exception of transmitters inside containment, have been inspected, inspection plans for transmitters located inside containment are under development and completion of the physical inspection is anticipated prior to start-up from the refueling outage scheduled for May and June 1997. ACR's have been generated for each discrepancy found. The task force had still not identified any transmitter discrepancy that affected instrument operability or environmental qualification. However, several minor deficiencies continued to be identified such as loose junction box door clamps, missing qualification tags, field model number discrepancies, and missing end caps for instrument drain line tubes.
Deficiencies are traceable to a work request, ACR or EWR The inspector verified that items with the potential to affect operability, environmental qualification, or calibration were reviewed by operations, system engineering or environmental qualification program personnel, as appropriate.
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Conclusions The licensee has maintained sound control of the inspection activities using the -
work request process for transmitter inspections and coordinating inspections to
~ ensure the protected train transmitters are not disturbed. The deficiencies identified represent conditions that do not conform with vendor specifications, however the conditions did not affect instrument operability or qualification. The deficiencies either have been corrected or are in the work planning process. Additionally, documentation of inspection status and results was very good. The dedicated task force has provided positive verification of the design-bases configuration j
requirements for a significant sample population of transmitters currently installed in j
the statiors. The inspector will review the final task force report and any
subsequent corrective actions will continue to be tracked by an inspector follow-up
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item. (Update IFl 50-443/96-10-01)
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ll. Maintenance M1 Conduct of Maintenance M1.1 (Closed) LER 96-005-00, Missed Surveillance Requirement a.
Inspection Scone (62707.61726J The inspectors reviewed Licensee Event Report (LER) 50-44 3/96-05, " Missed Surveillance. Requirement," dated August 8,1996, to evaluate the technical and programmatic aspects of the event and to verify that reporting requirements were fulfilled.
b.
Observations and Findinas On July 15,1996, the licensee determined that the American Society of Mechanical Engineers (ASME)Section XI surveillance interval for SI-V-53, Safety injection Accumulator "D ' Drain / Fill Connection had been exceeded. The normal frequency was quarterly stroke time testing, however recently the frequency had been accelerated to monthly due to a slight increase in the valvo's stroke time during the last quarterly surveillance test.
Technical Specification surveillance 4.0.5 requires that in-service testing of ASME Code Class 1,2, and 3 valves be conducted in accordance with Section XI of the.
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Boiler and Pressure Vessel code. Accordingly, quarterly stroke time testing of SI-V-53 is required. SI-V-53 is a normally-closed, fail-closed valve which is stroke time
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tested to verify closure and preserve SI accumulator integrity.
Previously, on June 5,1996, with the plant operating in Mode 1, at 100% of rated thermal power, SI-V-53 was stroke-time tested as part of post-maintenance retest.
The measured str 'ke time of 2.26 seconds in the close direction, exceeded the alert range limit of 2.'s seconds, but was less than the required action limit of 2.8 seconds. The Ucensee concluded that the maintenance on the valve should not have affected the valve's stroke time. The IST engineer placed the valve on the accelerated alert surveillance frequency of once every 31 days rather than the normal quarterly interval. The increase was annotated on test documentation by
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the IST engineer and forwarded to the planning and scheduling technician for I
schedule updating. However, the reviewing planning and scheduling technician did not notice the annotation and did not enter the revised alert frequency into the schedule.
However, on July 15,1896, another planning and scheduling technician identified the error and notified the IST engineer. SI-V-53 was stroke-time tested (1.34 seconds in the close direction) later that same day, however the surveil lance interval had been exceeded by two days. The licensee determined the cause of the event was personnel error since the change in surveillance frequency was not entered into the appropriate work control documents and procedural inadequacy in that the change in surveillance frequency required no verification. Licensee
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corrective actions included revision of the work control procedures to require independent confirmation that the work schedule has been revised to reflect IST results. The licensee concluded there were no adverse safety consequences as a result of this event. The completed LER evaluation indicated this is the first occurrence of a missed surveillance because a component was not adequately entered into an accelerated surveillance frequency due to IST results.
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Conclusions The inspector concluded that the LER properly documented the event and contained the appropriate reporting requirements of 10 CFR 50.73. - The corrective actions ~
were appropriate to address the causes of the event. The inspector verified the described corrective actions had been completed. The inspector review also determir ed that the licensee safety analysis was adequate. The licensee-identified and corrected violation is being dispositioned as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
M1.2 New Fuel Receipt inspection M
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Insoection Scone (62707.60705)
g On January 10,1997, the inspector observed station personnel perform new fuel receipt and inspection activities using Work Request WR 96 WOO 1025, and station procedures MS 0515.09, New Fuel Off loading, and RS 0722, New Fuel and Core Component inspection. Four new fuel assemblies, two in each of two separate
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casks were inspected.~The inspector observed receipt inspection activities in the-fuel storage building, reviewed the procedures, held discussions with Reactor Engineering and Maintenance personnel and performed independent inspections of
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the new fuel assemblies and their shipping casks.
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Observations and Findinas The inspector found that the new fuel shipping casks inspection, movement of the new fuel, and the inspections performed for each new fuel assembly, were done according to procedures in a carefully controlled manner. Specifically, the two
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separate and independent inspections performed by Reactor Engineering personnel l
were meticulously done. Also proper foreign material exclusion (FME) techniques l
were observed and e iforced by all personnel involved.
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Conclusions The inspector determined receipt inspection activities were performed correctly using the procedures. New fuel assemblies were thoroughly inspected visually for manufacturing defects, bowing, debris and damage. Maintenance, Health Physics l
and Reactor Engineering personnel worked well together, and properly documented l
their activities in the work package. The inspector had no further questions.
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l M8 Miscellaneous Maintenance issues M8.1 (Update) URI 50-443/96-02-02, Chainfall Hoist Attached to Safety-Related Equipment a.
Insoection Scone (62707)
NRC Inspection Report No. 50-443/96-02 documented the NRC identification of a chainfall hoist that was connected to the encapsulation tank for containment recirculation sump suction valve CBS-V-008, that was located in the mechanical penetration area of the primary auxiliary building. The item was considered unresolved, pending inspector review of a completed licensee Adverse Condition Report evaluation. The inspector reviewed the completed ACR 96-154 and held discussions with appropriate maintenance department personnel, b.
Findinas and Observations
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The ACR evaluation concluded the chainfall remained attached following performance of modifications to the valve in the November-December 1995
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l refueling outage. The top portion of the encapsulation tank had to be removed to perform the work. The chainfall hoist was used to accomplish this task. Contractor personnel no longer onsite performed the work and thus were not interviewed. The associated work package contained no specific or general direction for return of lifting equipment to proper storage locations.
. (The. apparent cause determination concluded the workers did not have proper
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direction for removal and storage of the chainfall. The field closeout form, MA
' 3.08, provided no direction for storage of permanent lifting equipment.
Additionally, the maintenance supervisor and system engineer did not discover the chainfall during walkdowns. Further, the condition was not discovered during subsequent supervisory plant walkdowns.
Corrective actions included:
Initiation of a request for engineering services RES96-264 to evaluate
installation of a permanent pad eye to store the chainfall.
Addition of this ACR to the pre-outage contractor training program.
- Revision of MA3.08, " Field Work Closecut Form," to include removal of
installed rigging / proper storage of permanently installed rigging.
Communication of this ACR to the management team.
- Communication of Maintenance Management and Unit Construction
Supervision expectations to supervisors regarding job closcout for significant/ complex evolutions.
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Conclusions
The inspector concluded the corrective actions taken or planned were adequate.
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The inspector reviewed plant records and verified the adequacy of completed
corrective actions and that planned actions were contained in the action item j
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tracking system. The ACR background information indicated the chainfall was stack l
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'and appeared to impart no stress on the valve encapsulation tank. The inspector d
. questioned why the ACR did not qualitatively or quantitatively evaluate the potential stress. The licensee indicated the ACR would be revised to address this
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j outstanding inspector question. Consequently, this item remains open pending
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licensee analysis of any potential adverse seismic interactions with the chainfall and l
encapsulation tank head not previously considered. This item remains open.
(Update URI 50-443/96-02-02)
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E1 Condui:t of Engineering E
E1.1 Generic Letter 89-10 Motor-Operated Valve Program Review (Tl 2515/109)
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Introduction and Puroose
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On June 28,1989, the NRC issued Generic Letter (GL) 89-10, " Safety-Related l
Motor-Operated Valve Testing and Surveillance," requesting licensees to establish a
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program to ensure that switch settings for safety-related motor-operated valves a
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(MOVs) were selected, set, and maintained properly. Seven supplements to the GL have been issued to provide additional guidance and clarification. NRC inspections
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of licensee actions implementing the provisions of the GL and its supplements have been conducted based on the guidance provided in NRC Temporary instruction
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i 2515/109, " Inspection Requirements for Generic Letter 89-10," which is divided j
into Part 1, " Program Review," Part 2, " Verification of Program implementation,"
and Part 3, " Verification of Progrem Completion."
i The NRC conducted the Part 1 inspection at Seabrook in December 1991 as l
documented in NRC inspection Report (IR) 91-81. A Part 2 inspection, which was conducted in May 1994 and documented in NRC IR 94-11, included an update of
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j the open items developed during the Part 1 inspection. The purpose of this Part 3 i
inspection was to closecut the NRC staff's review of the GL 89-10 program at i
Seabrook.
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E1.2 Summary Status of Generic Letter 89-10 MOVs e
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Insoection Scope
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Generic Letter (GL) 8910 requested that licensees notify the NRC in writing within i
30 days after the MOV design-basis reviews, analyses, verifications, tests, and
inspections have been completed. In letter NYN-94106, dated September 16,
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1994, North Atlantic Energy Service Corporation (North Atlantic) stated that differential pressure testing of 9 additional MOVs was needed to complete the GL 89-10 prograrn at Seabrook consistent with the guidance of GL 89-10, Supplement 6. North Atlantic committed to complete this testing during refueling outage (RFO)
04 and provide a completion report to the NRC within 30 days. North Atlantic submitted Seabrook's GL 89-10 design-basis closure report to the NRC by letter NYN-96002 dated January 15,1996. The inspectors reviewed the closure report,
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Seabrook MOV program procedure ES1850.003, " Motor Operated Valve Performance Monitoring," and other engineering documents to verify program completion, b.
Observations and Findiras The Seabrook program inc!uded 118 MOVs of which 60 were dynamically tested.
North Atlantic's methods for demonstrating MOV design-basis capability included (1) valve-specific dynamic test at, or near, design-basis conditions; (2) valve-specific test, linearly extrapolated to design-bacis conditions; (3) in-plant information obtained from dynamic tests on similar MOVs; and (4) calculations performed of adequate actuator capabili,ty for MOVs that were not practicable to test. The inspectors reviewed special test packages and engineering evaluations for the following MOVs.
RC-V-122 PORV 456A block valve RC-V-124 PORV 456B block valve RH-V-2 2 Residual heat removal train A discharge cross-connect SI-V-138 -
Charging pump supply to reactor coolant system cold legs While the closure report contained important information regarding the Seabrook
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MOV program, it did not contain adequate detail regarding the MOV available thrust margins, especially for untested MOVs. The inspectors obtained clarifications essential to a complete understanding of the report to evaluate the program for closure. In addition, other significant issues developed during the initial onsite inspection. North Atlantic responded by providing the following additional engineering documents on September 25,1996, for in-office review at Region 1:
Engineering Evaluation SS-EV-960019, " Valve Set Up Adequacy for Motor
Operated Valves That Are impractical to Dynamically Test," Revision 0, dated September 24,1996
Engineering Evaluation SS-EV-960020, " Evaluation of Differential Pressure Conditions for Motor-Operated Valves," Revision 0, dated September 24, 1996 Calculation C-S 1-86206, " Basis for Motor-Operated Valve Rate of Loading
Allowance," Revision 0, dated September 23,1996
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Commitment Change Request 96-04, " Deletion of Thermal Barrier Loop
Motor Operated Valves from Generic Letter 89-10 Test Program," dated September 23,1996.
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During a conference call with North Atlantic personnel on October 10,1996, the inspector discussed the following comments regarding these documents:
Additional documentation was needed to understand North Atlantic's justification of the valve factors assumed for establishing mwy MOV torque switch settings for untested MOVs. With the exception of tha pressurizer power operated relief valve (PORV) block valves, the inspector considered that North Atlantic had not presented in its engineering evaluations adequate applicable industry data to support MOV program closure. The inspector questioned twenty-one MOVs in which the available thrust margins ranged from approximately 13% to 38%, as provided in Engineering Evaluation SS-EV-960019.
- North Atlantic's justification for the assumption of a 0.15 stem friction coefficient was weak for the following reasons: (1) North Atlantic had acquired stem friction coefficient data for only 16 MOVs in the program; (2)
only eleven data points were acquired under dynamic conditicns; and (3) six data points exceeded the assumed stem friction coefficient value of 0.15.
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After revising the thrust calculations of the untested MOVs to address the inspectors' comments, North Atlantic concluded that the reactor coolant
thermal barrier component cooling water MOVs (3-inch 1500#, Velan flex-wedge gate valves) had very low (3-10%) thrust margins, with one MOV (CC-V395) having a negative thrust margin of -8.1 %. These revised
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calculations assumed a valve factor of 0.5, a stem friction coefficient of 0.15, an allowance of 20% for load sensitive behavior, and properly accounted for degradea voltage, motor performance temperature effects, torque switch repeatability, and diagnostic equipment uncertainties. North Atlantic decided to delete these MOVs from the Seabrook GL 89-10 program. This action was documented in Commitment Change Request 96-04, which included a 10 CFR 50.59 safety evaluation. North Atlantic noted that these MOVs were described in the updated final safety analysis report (UFSAR) as being nonsafety-related, and their operation was not credited in any UFSAR accident analyses. The MOVs are deenergized during normal operation, with their breakers open at the respective motor control centers.
This GL 89-10 program scope change will be reviewed by the NRC under followup item IFl 50-443/96-11-02.
Based on the above comments regarding untested MOVs at Seabrook, the inspector requested North Atlantic to review 21 MOVs to assess all available options to evaluate their thrust margins, such as: (1) additional conservatisms in differential pressures or ambient temperatures used in their calculations; (2) the impact of using a stem friction coefficient value greater than 0.15; (3) future increase of the MOV torque switch setting; or (4) potential MOV modifications, in addition to these l
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margin improvement efforts, North Atlantic was requested to provide more detailed technical justification of 5 valve factor assumptions. North Atlantic responded to these comments on October.31,1996, in Section 1 of Engineering Evaluation SS-EV-960019. NRC assessment of this engineering evaluation is discussed in Section E1.3.
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Conclusions More detailed justifications will be needed to complete NRC review of various technical assumptions used by North Atlantic in the Seabrook MOV program.
E1.3 MOV Sizing and Switch Settings a.
Insoection Scope The inspectors reviewed valve documentation that established the thrust requirements for MOVs in the Seabrook GL 89-10 program. These documents included thrust calculations and test evaluation packages associated with the selected MOVs. North Atlantic's methods for determining minimum thrust requirements were summarized in the Generic Letter 89-10 Design Basis Closure Report. The purpose of this review was to assess North Atlantic's justifications for assumptions used in MOV thrust calculations which form the basis for determining the design-basis requirements. North Atlantic's final justifications for these assumptions were included in Engineering Evaluation SS-EV-960019, which the inspector discussed during telephone conferences with the licensee on November 4 and 5,1996.
~ ~ North Atlantic's thrust calculations typically utilized the standard industry equations.
Mean seat or orifice diameter measurements were used (depending on the valve vendor) to calculate valve seat area. Valve factors were based on in plant test results, industry test results, or on engineering judgment. A stem friction coefficient of 0.15 was used for determination of actuator output thrust capability.
During valve setup, a bias margin was included to cover diagnostic equipment uncertainty and torque switch repeatability.
Valve Factor and Grouoina North Atlantic divided its MOVs into valve groups based on valve manufacturer, valve type, ANSI pressure class rating, actuator size, and motor size. North Atlantic attempted to use in-plant data to justify valve factors for non-dynamically tested MOVs. However, North Atlantic did not have sufficient site-specific test results to cover all valve groups adequately. Some examples were noted in which valve factor data did not establish adequate justification for program assumptions, including:
Group 10 consisted of four, eight-inch Westinghouse gate valves. A valve
factor of 0.5 was applied to the open direction and a valve factor of 0.6 was applied to the close direction. For the open direction, test results from 6 other
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valves were considered. The highest valve factor obtained from this testing was 0.57. Howaver, the licensee chose the 0.50 open valve factor on the basis of low differential pressure. The close valve factor of 0.60 appeared to bound the available data and was considered to be adequate.
Group 16 consisted of four,12-inch Westinghouse gate valves, which are the
RHR suction isolation valves that isolate the residual heat removal (RHR) system from the reactor coolant system hot legs. Initially North Atlantic applied a valve factor of 0.5 to the open and close direction. However, in Engineering Evaluation SS-EV-960019, the licensee changed the open valve factor to 0.53 and the closed valve factor to 0.56 after the review of other plant data for similar valves. The inspectors noted that valves RC-V-22 and RC-V-87 had close thrust margins of approximately 15%.
Load Sensitive Behavior North Atlantic's MOV switch setting methodology did not specify a margin for load sensitive behavior. This issue originally was noted in NRC Inspection Report 50-443/91-81 and discussed further in Inspection Report 94-11. Instead of setting aside specific margin, the licensee relied on engineering judgment that the design-basis differential pressures were conservative enough to account for the uncertainties associated with load sensitive behavior. The inspectors did not consider this approach to be acceptable ber ause the conservatism contained within the differential pressure assumptions was not quantified on a valve-specific bas.
-North Atlantic conducted a preliminary review of its load sensitive behavior data,;.
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and formalized a 20% allowance for load sensitive behavior in Calculation C-S-1-86206.
The inspecars also noted that North Atlantic did not include a load sensitive behavior margin in the open direction. Licensee engineers stated that an open direction margin was not needed because the torque switches typically are bypassed during the first 25% of valve travel. However, changes in stem friction coefficient which cause load sensitive behavior may still occur in the open direction.
Given the small amount of open direction dynamic stem friction coefficient data available at Seabrook, the inspectors did not consider North Atlantic's position to be justified.
Stem Friction Coefficient I
North Atlantic's thrust calculations assumed a ste o friction coefficient of 0.15.
The NRC expressed concern regarding to North Atlantic's justification for this value in inspection Report 50-443/94-11. In Engineering Evaluation SS-SE-960019, North Atlantic maintained its position that the 0.15 value was appropriate for MOVs at Seabrook. In light of ttle limited amount of data on which to base meaningful
= statistical results, North Atlantic also used qualitative. arguments based on Electric Power Research Institute (EPRI) Report TR-102135, "EPRI MOV Performance Prediction Program STEM / STEM-NUT LUBRICATION TEST REPORT," to support its assumption. The NRC generally has not accepted EPRI data as a sole basis for GL l
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89-10 program closure. While the data derived from EPRI testing is meaningful for various lubricants (Seabrook uses Fel-Pro N-5000), the testing did not simulate the different environmental conditions that pertain to specific plant valves. Thus, the friction coefficient data cannot be applied generically to the MOVs at Seabrook.
The inspector concluded that the licensee's justification of the 0.15 stem friction coefficient was unacceptable for GL 89-10 program closure, and also potentially nonconservative for low-margin MOVs that have not been tested dynamically.
North Atlantic committed as part of the periodic verification program to perform dynamic testing of 15 additional MOVs during the next two refueling outages in order to obtain additional site-specific stem friction coefficient data, c.
Conclusions The combination of insufficiently justified assumptions for valve factor, load sensitive behavior, and stem friction coefficient raised questions regarding design-basis capability for some untestable MOVs. North Atlantic's technical approach associated with these assumptions was not acceptable, and will be reviewed under followup item IFl 50-443/96-11-03 prior to GL 89-10 program closure.
E1.4 Design-Basis Capability a.
Inspection Scope The inspcctors reviewed dynamic test evaluation packages and associated test reports for the selected MOVs. The purpcse of this review was to assess North '
Atlantic's efforts to establish design-basis capability for all MOVs in Seabrook's GL 89-10 program.
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Observations and Findinas Dearaded Voltaae As described in section 8.3.1.1b.4(b) of the Seabrook FSAR, the degraded grid relay provides a second level of undervoltage protection for Class 1E equipment. It monitors the voltage at 4160 volt safeguard buses E5 and E6, and has a setpoint of 3933 volts. Energization of the degraded grid relay does not separate the E5 and E6 buses from offsite power unless there is a concurrent safety injection (SI) signal.
If the degraded grid relay is activated without an SI signal, then offsite power to the buses is maintained, and only a control room alarm is received.
During the first week onsite, the inspectors reviewed several MOV calculations in which North Atlantic used minimum voltage values that were not based on the supply voltage at the degraded grid relay setpoint. This was contrary to the NRC I
position discussed in Supplement 1 of GL 89-10. The inspectors observed that the licensee also had identified the inconsistency in an MOV program self-assessment that was conducted in June 1995. In the self-assessment report executive summary, North Atlantic stated that "..the electrical voltage regulation calculation used a non-conservative degraded relay setpoint value as recommended in GL 89-
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10. This concern was discussed at length with Engineering and Licensing. North Atlantic Energy Service Corporation's position is the same as was taken in response to a previous audit observation (Audit No. 93-A08-02) and that has previously been accepted by the NRC in response to the Electrical Distribution System Functional Inspection assessment."
r After several telephone conferences and a second onsite visit from October 2-4, 1996, North Atlantic developad electrical calculations containing revised minimum MOV voltages under deoraded voltage conditions. These calculations used a starting voltage of 3%3 volts at the E5 and E6 buses to determine the MOV terminal voltages. The results of these calculations were summarized in Engineering Evaluation CS-EV-960021, Revision 0, dated October 4,1996, " Evaluation of Motor-Operated Valve Performance at the Degraded Grid Relay Setpoint." Using the new degraded grid relay setpoint of 3933 volts, the limiting value for the voltage at the MOV terminals was recalculated based on case studies using the DAPPER computer program from electrical calculation 9763-3-ED-00-02-F, Revision 5,
" Voltage Regulation" (aka Calculat;cn 02). Calculation 02 provides three cases for evaluating minimum available MOV voltage: (1) Full Load Case; (2) Diesel Generator Loading Case; and (3) Simultaneous Start Case. The MOV terminal voltages for the Diesel Generator Loading Case were unaffected. The Full Load Case resulted in MOV terminal voltages that were higher than those used in the current MOV capability calculations. However, the Simultaneous Start Case resulted in lower MOV terminal voltages for seventeen MOVs, where the terminal voltages declined by about 12-14 volts. For conservatism, North Atlantic selected a 20-volt reduction
- to revise the thrust calculations of the seventeen MOVs to assess their available thrust margins. North Atlantic concluded that all MOVs in the Seabrook program were capable of performing their safety functions with E5 and E6 bus voltage at the 3933 volt degraded grid relay setpoint.
The inspectors reviewed Engineering Evaluation SS-EV-960021, including Attachment 2, "Available Margin for the Simultaneous Start MOVs," which provided the basis for North Atlantic's conclusion that adequate capability existed for the seventeen MOVs under degraded voltage conditions. Attachment 2 included the new minimum voltage values to be incorporated into a revision of Calculation 02 for the seventeen MOVs and provided the derated thrust capability based on the new minimum voltages. Attachment 2 also included appropriate allowances for diagnostic equipment inaccuracy to determine the available thrust margin. North Atlantic willinclude this new information into updated MOV design basis calculations with the followup action tracked by Action Request 96030539. The inspectors concluded that North Atlantic provided a prompt and thorough response l
to an issue that had near-term operability implications for many MOVs.
Marain Imorovement The inspectors did not consider that an appropriate valve factor basis had been established by North Atlantic to provide adequate thrust margins for all Seabrook MOVs. As noted in Section E1.3, for example, some of the valves in Group 16 (12-inch Westinghouse gate valves) had low apparent thrust margins, with valves RC-V-l L.
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22 and RC-V-87 having margins in the closed direction of approximately 15%.
These valves were among the 21 MOVs selected by the inspector for North Atlantic to evaluate margin improvement measures to assure MOV design basis capability.
In Engineering Evaluation SS-EV-960019, North Atlantic agreed to incorporate certain margin improvement measures as part of the Seabrook MOV program.
i Specifically, for GL 89-10 MOVs with less than 10% thrust or torque margin or less
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than 1000 lbf. of thrust margin, an adverse condition report (ACR) will be initiated and a corrective action plan developed. The remaining valves in the test group will
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also be evaluated as part of the ACR. Valves that have less than 10% available margin will be scheduled for margin :.nprovement no later than during the next scheduled refueling outage.
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Conclusions Notwithstanding the positive measures taken to assure design-basis capability, including an acceptable resolution of the degraded voltage issue, North Atlantic's assumptions regarding valve factor, stem friction coefficient, and load sensitive behavior were not fully justified for untested MOVs.
E1.5 MOV Failures, Corrective Actions, and Performance Trending i
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Inspection Scoce Item (h) of GL 89-10 requested licensees to include a monitoring and feedback
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' effort in the MOV program to establish trends in MOV operability. The inspectors-
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. reviewed Seabrook's MOV trending practices as described in procedure d
ES1850.003, " Motor Operated Valve Performance Monitoring." Corrective actions
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l concerning recent problems with spray additive tank valves CBS V38 and CBS-V43
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were included in the review.
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Observations and Findinas j
The inspectors noted the following requirements included in procedure ES1850.003-i MOV failures which result in declaring the equipment inoperable require formal e
evaluations in accordance with the Station Operating Experience Manual. The failures are included in annual system performance reports.
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MOV performance is trended to ensure adequate switch settings throughout the j
e life of the plant. Trended parameters, such as peak unseating load and l
maximum load, are recorded on Forms ES1850.003C and D. MOV thrust margins are calculated, evaluated, and trended by the MOV system engineer.
Valve CBS-V38 experienced a demand failure during a routine in-service test in
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August 1996. The licensee attributed the failure to a loose electrical connection at
the motor terminals. North Atlantic also concluded that the closed torque switch
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was set too high, and the setting was lowered as part of the corrective actions.
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i The redundant valve, CBS-V43, was diagnostically tested in October 1996. While the valve cycled satisfactorily, its available thrust margin was less than 10%. In accordance with the recent thrust margin improvement program (see Section E1.4),
a gear change modification was implemented to upgrade the valve's performance capability, c.
Conclusions North Atlantic established and implemented an adequate MOV trending program as recommended by GL 89-10.
E1.6 MOV Program Administration a.
Insoection Scoce-i The inspectors reviewed the governing MOV Program procedure 1850.003 i
throughout the inspection and observed how the various implementing procedures
t were controlled to fulfill program requirements, b.
Observations and Findinas The inspectors noted that Revision 1 of procedure ES1850.003 was issued over two years ago and required a major update to reflect current MOV Program status.
For example, Figure 10.1 included the MOV differential pressure test groups, and-this information differed significantly from that included in the Seabrook GL 89-10
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design-basis closure report. North Atlantic committed to complete a revision to -
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procedure ES1850.003 by January 31,1997.
During the review of open items from prior MOV inspections, the inspectors observed that several procedures had not been updated as expected.
e The Rotork maintenance procedure, LS 0569.27, " inspection / Preventive Maintenance of Rotork Valve Actuator," should have been updated after NRC Inspection 94-11 to included a caution note regarding the impact of the limit switches if the valve is operated manually (see Section E8.10). Maintenance and operating personnel were trained using an updated lesson plan and training module shortly after the conduct of NRC Inspection 94-11. However, the j
maintena. ice procedure was not revised until this inspection.
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t e During NRC Inspection 94-11, North Atlantic agreed to enhance MOV Jynamic test procedures to provide specific sign-off steps that require completion of an engineering evaluation of test data prior to returning a valve to service. Also, acceptance criteria were to be specified in the body st the procedure, although this information was provided to field personnel by the MOV engineer using Form 1850.003E (see section E8.8). The MOV test procedure changes were made during this inspection.
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There were no adverse safety consequences as a result of the procedure update issues.
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Conclusions North Atlantic committed to completa a revision of procedure ES1850.003 by January 31,1997, due to inconsistencies with program closure documents and current practices.
E7 Quality Assurance in Engineering Activities E7.1 Independent Assessment and Manaaement Suncort a.
Insoection Scope The inspector reviewed a report of an independent assessment conducted of the Seabrook MOV program. This review was conducted during the latter part of the first onsite week of inspection, b.
Observations and Findinas An independent assessment of the Seabrook MOV program to evaluate its readiness for closure was conducted in June 1995. The assessment team was led jointly by experienced North Atlantic and Yankee Atomic quality assurance personnel assisted by three personnel: a Yankee Atomic quality assurance auditor, an engineer from
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another nuclear facility, and a contractor personnel (Vectra). As discussed in Section E1.4, the team challenged the licensee regarding the degraded voltage issue. This issue was not fully resolved until this inspection.
E8 Miscellaneous Engineering Issues The inspectors reviewed the following MOV program issues that were discussed in Inspection Report 50-443/91-81. Section numbers refer to inspection Report 50-443/91-81.
E8.1 (Closed) Inspection Report 50-443/91-81 dection 2.7: Justify extansion of motor-actuator inspection and lubrication interval beyond vendor recommendation.
Limitorque Corporation Bulletin SMBI-82D recommended inspection and lubrication of motor-actuators at least every 18 months. As prescribed by Section 3.5.5.4 of Station Procedure ES1850.003, North Atlantic performs preventive maintenance, including lubrication, of GL 89-10 program MOVs every other refueling outage. The licensee's justification for this deviation from the vendor's recommendation was not documented in the GL 89-10 program. The licensee MOV coordinator provided the
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inspectors with MOV performance reports that documented the preventive maintenance activities that have been performed on safety-related MOVs during the last several years. North Atlantic also evaluates site-specific and industry component failure information, and compares as-found diagnostic test data with initial calibration data. Finally, as documented in Evaluation 95-TSEV-0002, the I
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l licensee has attempted to evaluate main gearbox lubricant in response to a concern
regarding color and consistency. However, lubricant sample testing has not been j
performed due to radiological contamination of the sample. The inspector l
concluded that the licensee has not thoroughly verified the effectiveness of the i
current maintenance and lubrication interval. This issue will be reviewed as part of
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the followup item (IFl 96-11-03) concerning stem friction coefficient discussed in Section E1.3.
E8.2 (Closed) Inspection Report 50-443/91 "1, Section 2.10: Disposition of Limitorque Corporation Maintenance Upda+.,s 88-2 s. 4 00-1. These maintenance updates involve potential hydraulic Lcking of motor-acator spring packs. North Atlantic
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initially evaluated this iuue in response to industry Significant Event Reports 30-86 l
and 20-87. Based on a review of actuator serial numbers, the licensee determined that its actuators have internal grease relief paths. MOVs susceptible to hydraulic lock have not been modified unless the condition has been observed during normal surveillance or diagnostic testing. Modified springpacks were installed on containment sump isolation valves CBS-V-14 and CBS-V-8 after valve CBS-V-14 experienced the phenomenon. In addition, the licensee implemented procedure changes requiring maintenance personnel to check for excess grease and to clean the spdng packs during motor-actuator maintenance. Seabrook ICTS Action item OE11426 tracks the issue. The inspectors concluded that North Atlantic adequately addressed the potential for hydraulic locking of spring packs at Seabrook.
E8.3 (Closed) Inspection Report 50 443/91-81, Section 3.0: Control of switch positioning to preclude short stroking. Walkdowns conducted during Inspection No.
50-443/91-81 identified that several MOV control switches in the control room
.were fixed position switches, that did not return to the neutral position when released by the operator. A concern was identified that valves may be damaged due to excessive seating tnrust generated during short stroking if manual operation is attempted with a co.nrol switch that is in the closed position. The inspectors reviewed station operating procedure OSO90.0, " Manual Operations of Remote MOVs." The procedure contained a list of MOVs that could be short stroked during power restoration following manual operation, and instructs operators to position the valve between 50% and full open prior to restoration of power. Similar caution notes also were added to procedures LS0569.01 and LS0569.05, and Operations Department Lesson Plan Number L1641C. These actions adequately addressed the NRC concern.
E8.4 (Closed) Inspection Report 50-443/9181, Section 2.2: Review of Westinghouse weak link analysis methodology. This item involved completion of North Atlantic's review of the seismic analyses used to establish MOV structural limits. The inspectors verified that the licensee completed its review of the Westinghouse methodology, and that appropriate structural limits were incorporated into MOV l
switch settings.
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E8.5 (Closed) inspection Report 50-443/91-81, Section 2.4: Validate stem friction coefficient assumption. The licensee assumed a stem friction coefficient of 0.15 to determine actuator thrust output. In New Hampshire Yankee letter NYN-92058,
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dated April 30,1992, the licensee justified the assumption using industry information, but did not discuss the results of plant-specific data. As discussed in Section E1.3, the inspector's found North Atlantic's justification for the 0.15 friction coefficient value to be unacceptable for GL 89-10 program closure. This item will be tracked under followup item IFl 50-443/96-11-03.
l E8.6 (Closed) Inspection Report 50-443/91-81, Section 2.4: Margin for load sensitive behavior. The licensee had not included a specific margin for load sensitive behavior in its evaluations of MOV design-basis capability. _ As discussed in Section
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E1.3,- North Atlantic did not adequately resolve NRC concerns regarding treatment i
of load sensitive behavior. This issue will be tracked under followup up item IFl 50-443/96-11-02.
l E8.7 (Closed) Unresolved item 50-443/94-11-02: The licensee's dynamic test plan did not meet the valve grouping recommendations (30% or a minimum of two MOVs)
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discussed in Supplement 6 of GL 89-10. The inspectors reviewed North Atlantic
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report NYN-96002, " Generic Letter 89-10, Design-Basis Closure Report," dated
January 15,1996, and discussed with the licensee two groups that appeared not to meet the recommendations of GL 89-10. The valves in Giaup 7 (auxiliary steam
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system) were not practical to test under design-basis conditions, and only one of.
the two MOVs in Group 17 experienced differential pressure under design-basis conditions. The inspector concluded that North Atlantic's dynamic test plan was consistent with the recommendations of the generic. letter supplement,
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' E8.8 (Closed) Inspection Report 50-443/91-81, Section 2.5:- Develop dynamic test
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procedure guidance and acceptance criteria for evaluating MOV design-basis-capability. During the Part 1 inspection in 1991, the NRC noted that adequate
procedures had been established to obtain data for signature analysis, but that
' acceptance criteria for the diagnostic data had not been established. As l
documented in NRC Inspection Report 50-443/94-11, the NRC observed that North Atlantic had improved the evaluation portion of its procedures, but still did not
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provide adequate detailed guidance for evaluating design-basis capability prior to
returning an MOV to service. During the current inspection, while reviewing a recent dynamic test package for valve RH-V 22, the inspectors noted that the test procedures did not clearly require completion of the engineering evaluation of the test data prior to returning the valve to service. North Atlantic personnel agreed with the observation and provided documents showing that the MOV had been fully evaluated before it was returned to service. The inspectors concluded that adequate administrative controls existed to ensure that engineering evaluations of dynamic test data are performed prior to returning MOVs to service.
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t E8.9 (Closed) Inspection Report 50-443/91-81, Section 2.6: Periodic verification of j
design-basis capability. During the current inspection, North Atlantic changed
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procedure ES1850.OO3, "MOV Performance Monitoring Program," to update the required MOV diagnostic test frequency to once every three refueling outages. The j
- of diagnostic testing if necessary. The inspectors found the procedure to be procedure change also allows the MOV System Engineer to increase the frequency
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consistent with the recommendations in GL 89-10. North Atlantic has not yet
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determined to what extent dynamic testing will be performed. This issue will be reviewed by the NRC under GL 96-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves."
E8.10 (Closed) Inspection Report 50-443/91-81, Section 2.7: Revise Rotork operator p'ocedure and training module to caution against inadvertently changing limit switch settings. During the Part 1 inspection, the licensee agreed to revise its procedures to include a caution when manually operating Rotork actuators beyond the limit switch settings. The revision was needed because manual operation of the actuator could reset the switch setting and affect remote MOV operation. The licensee also
' stated that training would be enhanced to ensure that maintenance and operations personnel were aware of the potential for inadvertently affecting the switch settings. The inspectors verified that the licensee revised training modules and lesson plans to caution against inadvertently changing Rotork limit switch setpoints.
However, the inspector identified that procedure LS0569.27, " Inspection /PM of Rotcrk Valve Actuator," had not been changed. The MOV coordinator gave the inspector a copy of a proposed procedure change, dated September 20,1996, that included a caution against changing Rotork limit switch setpoints when manually operating valves. The inspector concluded that the procedure change adequately addressed NRC concerns regarding manual operation df Rotork motor-actuators.
E8.11 Review of Updated Final Safety Analysis Report (UFSAR) Commitments A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a a
special focused review that compares plant practices, procedures, and/or parameters to the UFSAR descriptions. While performing the inspections documented in this report, the inspector reviewed the applicable portions of the UFSAR that related to the areas inspected and verified that it was consistent with the observed plant practices, procedures, and/or parameters.
IV. Plant Supoort R1 Radiological Protection and Chemistry Controls R 1.1 General Comments a.
Insoection Scone During the inspection period the inspector toured the radiologically controlled area (RCA) on several occasions to observe radiological controls practices.
b.
Observations and Findinas The Seabrook Station radiological controls technicians at the RCA checkpoint were attentive and provided assistance to radiation workers to assure proper work practices were used when radiation workers signed in and out of the RCA. The l
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inspector determined that radiation area postings were proper and well marked and survey results were current and posted properly. All personnel observed were properly wearing dosimetry while in the RCA. A sampling of high radiation area doors identified no discrepancies with locking or posting requirements.
c.
Conclusions The inspector determined that Seabrook Station was properly implementing the
- station radiological controls program requirements in the areas inspected.
Radiological controls personnel were knowledgeable of station procedules and provided good oversight of radiation workers. Department managers were observed in the field observing and supervising department personnel.
S1 Conduct of Security and Safeguards Activities S1.1 General Comment (71707, 71750)
The inspectors observed security force performance during inspection activities.
Protected area access controls were found to be properly implemented during random observations. Proper escort control of visitors was observed. Security officers were alert and attentive to their duties.
F1 Control of Fire Protection Activities F1.1 Fire Detection Equipment Surveillance b.
Observations and Findinas (71707,71750)
On December 11, the inspector observed portions of primary auxiliary building (PAB) trip actuating device operational test (TADOT) for smoke detectors and manual pull stations (procedure OX 0443.92 and RTS 96R04392A001).
The inspector held discussions with involved fire protection personnel who performed the evolution and directly observed the satisfactory test of several PAB smoke detectors. The smoke detectors were tested using the magnet tester. The smoke detector test is satisfactory if the detector alarm LED energizes, proper detector address is displayed on the PAB fire detection panel FP-CP-559 and the alarm resets automatically at CP-559.
c.
Conclusions l
The inspector concluded the surveillance was satisfactorily performed. Involved
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personnel were knowledgeable of the fire detection equipment operation and associated surveillance procedure. The testing was well controlled and sound communications and coordination with control room operators was observed. Fire protection personnel properly documented the test results. The inspector had no further questions.
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V. Manaaement Meetinas
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'1xit Meeting Summary Tne licensee was informed of the purpose and scope of the MOV inspection documented in Section ill of this report at an entrance meeting conducted on September 16,1996. Findings were discussed periodically with the licensee-throughout the course of the inspection. The inspectors met with the principals listed below on September 20 and October 4,1996 and a final exit meeting with the licensee was conducted on November 15,1996 to discuss preliminary MOV inspection findingL The licensee acknowledged the preliminary findings and conclusions, with nu c.v. aptions taken. The bases for the inspection conclusions did not involve proprietary information, nor was any such information included in this inspection report.
The resident inspectors presented the results of the remaining top:cs documented in this inspection report to members of licensee management, following the conclusion of the inspection period, on March 7,1997. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
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I PARTIAL LIST OF PERSONS CONTACTED Licensee W. Diprofio, Unit Director *
G. Kline, Technical Support Manager *
R. White, Design Engineering Manager *
J. Peterson, Maintenance Manager J. Grillo, Operations Manager" B. Seymour, Security Manager W. Leland, Chemistry and Health Physics Manager P. Brown, Principal Engineer *
R. Faix, Engineering Supervisor *
P. Searforce, MOV System Engineer *
G. Sessler, Project Engineer *
J. Sobotka, Licensing Engineer *
NRC
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Albert W. DeAgazio, Project Manager T. Scarbrough, EMEB/NRR*
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INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726:
Surveillance Observation IP 62707:
Maintenance Observation IP 64704:
Fire Protection Program IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 73051:
Inservice inspection - Review of Program *
IP 73753:
Inservice inspection IP 83729:
Occupational Exposure During Extended Outages IP 83750:
Occupational Exposure IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Fe-ilities IP 92902:
Followup - Engineering IP 92903:
Followup - Maintenance IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors Tl 2515/109 inspection Requirements for Generic Letter 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-443/96-11-01 IFl Review Final correction of EFW design vulnerability 50-443/96-11-02 IFl Review MOV program scope change for thermal barrier cooling valves 50-443/96-11-03 IFl Justify technical assumptions regarding valve factor, stem friction coefficient, and load sensitive behavior Qlosed 50-443/94 11-02 URI Motor-operated valve grouping criteria not consistent with GL 89-10 Supplement 6 LER 50-443/96-05 LER Missed surveillance requirement Discussed /Uodated 50 443/96-10-01 IFl Pressure transmitter configuration control task force
50-443/96-02-02 URI Chainfall hoist attached to safety-related equipment
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LIST OF ACRONYMS USED ACR Adverse Condition Report ASME American Society of Mechanical Engineers CAS Central Alarm Station CBS containment building spray EDG Emergency Diesel Generator EFW Emergency Feedwater EOP Emergency Operating Procedure FME Foreign Material Exclusion gpd gallons per day gpm gallons per minute LCO Limiting Condition for Operation MDEFW '
motor-driven emergency feedwater pump MOV motor operated valve MPCS Main Plant Computer System NSARC Nuclear Safety and Audit Review Committee NSARC OS NSARC Operations Subcommittee psig pounds per square inch gauge QC Quality Control RES Request for Enpinsering Services RHR Residual Heat Removal SG steam generator SIR Station Information Report SORC Station Operations Review Committee SW Service Water TDEFW turbine-driven emergency feedwater pump TS Technical Specifications UFSAR Updated Final Safety Analysis Report WR Work Request
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