LR-N970607, Forwards Response to 970911 & 15 Telcon RAI Re Relief Request for Augmented Reactor Vessel Insp.Procedure Controls Operator Training to Prevent over-pressurization of RPV, Summarized

Forwards Response to 970911 & 15 Telcon RAI Re Relief Request for Augmented Reactor Vessel Insp.Procedure Controls Operator Training to Prevent over-pressurization of RPV, Summarized

ML20210V456
Person / Time
Site:	Hope Creek
Issue date:	09/16/1997
From:	Dawn Powell Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEIN-97-063, IEIN-97-63, LR-N970607, NUDOCS 9709240011
Download: ML20210V456 (6)

Text

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. O.PSEG Public Service Electric and Gas Company P.O. Bov 236 Hancocks Bridge, New Jersey 08038 0236 Nuclear Business Unit SEP 161997 LR-N970607 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 ADDITIONAL INFORMATION REGARDING RELIEF REQUEST AUGMENTED REACTOR VESSEL INSPECTION HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF 57 DOCKET NO. 50-354 i

Gentlemen:

In a letter dated August 29,1997, Public Service Electric and Gas Company (PSE&G) requested relief from parforming the reactor pressure vessel (RPV) examination requirements of the American Society of Mechanical Engineers (ASMF; Boiler and j Pressure Vessel Ccde,Section XI,1983 Edition, through Summer 1983 Addenda, and  !

the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) for the Hope {

Creek Generating Station Specifically, pursuant to provisions of 12 CFR 50.55a(a)(3)(i), and consistent with information contained in NRC Information Notice 97-63, relief was requested from the examinatica of the RPV circumferential shell welds for two operating cycles. During a telephone conference dis :ssion on September 11 and September 15,1997, the NRC Staff requested additional information regarding procedure controls and operator training to prevent over-pressurization of the RPV during cold shutdown, and regarding the results of previous inservice examinations of RPV welds. Attachment 1 contains the requested information.

Shou!d there be any westions concerning this submittal, please do not hesitate to contact us.

Sincerely, I

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David R. Powell Direc -

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Document Control Desk -

LR-N970607 SEP is 1997 C- Mr. H. Miller, Administrator - Region i U.- S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr, D..Jaffe, Licensing Project Manager - HC ,

U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr S. Morris (X24)

USNRC Senior Resident inspector - HC Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Ardic Parkway -

1. PO Box 415 Trenton, NJ 08625

LR N970607 ATTACHMENT 1 ADDITIONAL INFORMATION REGARDING RELIEF REQUEST AUGMENTED REACTOR VESSEL INSPECTION Procedural Controls and Operator Trainina to Prevent Reactor Pressure Vessel (RPV)

Cold Over-Pressurization Events:

Operating procedural restrictions, operator training, and work control processes for the Hope Creek Generating Station provide appropriate controls to rainimize the potential for RPV cold over-pressurization events.

During normal cold shutdown conditions, reactor water level, pressure, and

^ temperature are maintained within established bands in accordance with operating procedures. Operations procedures governing Control Room activities require that Nuclear Control Operators (NCOs) frequently monitor for indications and alarms to detect abnormalities as early as possible. These procedures also require that the Nuclear Shift Supervisor (NSS) be notified immediately of any changes or abnormalities in indications. These procedures also require that changes which could affect reactor level, pressure, or temperature only be performed with the knowledge and at the direction of the NSS. Therefore, any deviations in reactor water level or temperature from a specified band will be promptly identified and corrected. Finally, the status of plant conditions, on-going activities which could affect critical plant parameters, and contingency planning are discussed by Operators at each shift turnove.- This ensures that on-coming Operators are cognizant of activities which could adversely affect reactor level, pressure, or temperature.

A review of industry operating experience indicates that inadequate work management is a potential contributor to a cold over-pressurization event. At Hope Creek, work performed during outages is scheduled by the Outage Management group. Dedicated Senio: Reactor Operators review the outage schedules to avoid conditions which could adversely affect reactor water level, pressure, or temperature. From the outage schedule, a plan of the day (POD) is developed listing the work activities to be performed. These PODS are reviewed and approved by station management, and a copy is maintained in the Control Room. Changes to the PODS require management review and approval. In addition, the detailed cutage schedule receives a thorough i shutdown risk assessment review to ensure defense-in-depth is maintained. l During outages, work is coordinated through the Work Control Center, which provides an additional level of Operations oversight. In the Main Control Room, the NSS is l required to maintain cognizance of any activity which could potentially affect reactor l level or decay heat removal during refueling outages. The NCO is required to provide '

positive control of reactor water level and pressure within the specified bands, and promptly report when operating outside the specified band, including restoration  ;

Lctions being taken. Pre-job briefings are conducted for work activities that have the i potential to cffect critical reactor parameters. These briefings are attended by the l Page 1 of 4 e _

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l LR N970607 Att chment 1 l cognizant individuals involved in the work activity. Expected plant responses and I contingency actions to address unexpected conditions or responses that may be encountered are included in the briefing discussion.

The plant procedure for shutdown requires opening of the head vent valves after the reactor has been cooled to less than 212*F. Procedural controls for reactor temperature, level, and pressure are an integral part of Operator training. Specifically, Operators are trained in methods of controlling water level within specified limits, as well as responding to abnormal water level conditions outside the established limits.

Additionally, NCOs receive training on b ;tle fracture limits and compliance with the Technical Specification pressure-ten ature limits curves. Plant-specific procedures have been developed to provide gu .ce to the Operators regarding compliance with the Technical Specification requirements on pressure-t9mperature limits.

Review of Potential Hiah Pressure iniection Sources:

During normal cold shutdown conditions, RPV level and pressure are controlled with the Control Rod Drive (CRD) and Raactor Water Cleanup (RWCU) systems using a

" feed and bleed" process. The reactor is not taken solid during these times, and plant procedures require opening the head vent valves after the reactor has been cooled to less than 212*F. If either of these systems were to fail, the NCO would adjust the other system to control level. Under these conditions, the CRD system typically injects water into the reactor at a rate less than 60 gpm. This slow injection rate allows the operator cufficient time to react to unanticipated level changes and, thus, significantly reduces the possibility of an event that would result in a violation of the pressure-temperature limits.

The Standby Liqud Control (SLC) system is another high pressure water source to the RPV. The SLC system initiates automatically only if reactor power is not downscale after a low reactor water level or high reactor pressure condition; so SLC system automatic initiation will not occur in cold shutdown. The injection rate of the SLC pump is approximately 46 gpm, which would give the Operator ample time to control reactor pressure in the case of an inadvertent manual injection.

Pressure testing of the RPV is classified as an " Infrequently Performed Test or Evolution," which ensures that these tests receive special management oversight and procedural controls to maintain the plant's level of safety within acceptable limits. Hope Creek practice is to heat up the reactor to hydrostatic test temperature (approximately 170 to 180*F) prior to increasing pressure. During performance of an RPV pressure test, level and pressure are controlled with the CRD and RWCU systems using a " feed and bleed" procedure. The SLC system is normally isolated from the RPV during pressure testing. Increase in pressure is limited to 50 psi per minute. This practice minimizes the likelihood of exceeding the pressure-temperature limits during performance of the test.

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1 LR N970607 Attrchment 1 l

'The probability of the High Pressure Coolant injection and Reactor Core Isolation Cooling pumps contributing to a cold over-pressurization event were addressed in m

PSE&G's previous submittal on this subject dated August 29,1997, i inservice Examination Result,3 During Hope Creek's fourth refueling outage, one circumferential weld and three axial welds were examined ultrasonically. Recordable indications were observed in the circumferential weld and in one of the axial welds.. A summary of the recorded indications is provided in Table 1. Percent of distance - amplitude correction (DAC) curve, length, width, and through-wall dimension are shown for each recorded indication. The largest recorded through-wall dimension,0.07 inches, is bounded by the flaw size distribution used in the NRC's independent assessment of the Boiling

- Water Reactor Vessel and Internals Project BWRVIP-05 Report.

i The indications found at Hope Creek are very small, midplate reflectors. Although they required evaluation, they were well within the acceptance standards in ASME section

. XI, IWB-3510, i

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LR N970607 AttCchment 1 Table 1 Hope Creek Fourth Refueling Outage ISI Recorded RPV Indications Indication %DAC Width Length Through-wall (inches) (inches) Dimension (inches)

- WELD W8 (Top Side)

1. 2 29% 0 0.36 0.03 (Combined Indication) #8 21 % 1.05 0.26 0 23%

29%

33 %

25%

21 %

4 (Combined Indication) #3 58 % 0.24 0.25 0.04

53 %

WELD W8 (Bottom Side)

1. 1 30% 0 0.19 0.02

1. 4 50 % 0 0.04 0.01 WELD WIS - 3

1. 5 21 % 0 0- 0 (Combined Indication) #7 45% 0.65 0.21 0.07 37 %

1. 6 50 % 0.06 0 0.02 Page 4 of 4