ML20117M649

Requests Relief from Specific Requirements Contained in ASME XI,1986 Edition,Section IWA-5250(a)(2),per 10CFR50.55(a)(3)

ML20117M649
Person / Time
Site:	Oyster Creek
Issue date:	09/12/1996
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6730-96-2282, NUDOCS 9609170595
Download: ML20117M649 (5)

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GPU Nuclear, Inc.

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U.S. Route #9 Couth NUCLEAR P**tomc* Bu*388 Forked River, NJ 087310388 Tel 609-9714000 September 12,1996 6730-96-2282 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555

Dear Sir:

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Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Request for Relief R15, Revision 1 By letter dated October 25,1994, the USNRC approved the Inservice Inspection (IL.) Program for the Oyster Creek Nuclear Generating Station. This program was v ritten to meet the 1986 edition of ASME XI, with no addenda. On August 19,1996, GPU Nuclear. Submitted ASME Relief Request R15, addressing leaking ' 'ted flanges. Discussions with the staff have indicated that additions and clarifications Tu!d assist in the review of that submittal.

Pursuant to 10 CFR 50.55(a)(3), this letter is being written to request relief from specific requirements contained in ASME XI,1986 edition, Section IWA-5250(a)(2). This letter supersedes the August 19,1996, letter and replaces it in its entirety. The details and justification of this request are contained in Attachment I.

If any additional information or assistance is required, please contact Mr. John Rogers of my staff at 609.971.4893.

gpichae B. Roche 9609170595 960912 Vice President and Director PDR ADOCK 05000219 G

PDR Oyster Creek MBR/JJR Attachment

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Oyster Creek NRC Project Manager j

Administrator, Region I l

Senior Resident inspector

Attachment I Relief Reauest R15 CODE

REFERENCE:

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ASME Section XI 1986 edition, without addenda.

IWA-5250(a)(2), CORRECTIVE ACTIONS l

APPLICABILITY:

Categories B-P, C-H, D-A, D-B, and D-C. Class 1,2, and 3 Pressure Retaining Components.

i CODE REQUIREMENT:

IWA-5250 (a) (2) "if leakage occurs at a bolted connection, the bolting shall be removed as specified, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100;".

RELIEF REQUEST:

This relief request is intended to authorize alternatives to the removal of bolting at pressure retaining connections when evidence of leakage is detected during system pressure testing.

Relief is requested from compliance with IWA-5250(a)(2), for removal of bolting from leaking bolted connections. GPU Nuclear would implement the alternative of performing corrective measures, as deemed necessary by GPU Nuclear Engineering Staff evaluations, BASIS FOR RELIEF:

Compliance with the ASME Code requirement for the removal of bolting has resulted in undue hardship to the plant without a compensating increase in the level of quality or safety.

Removing a system or subsystem from service or potentially shutting down the plant in order to remove bolting that has not been degraded by leakage is impractical. Experience has shown that equipment can be damaged, personnel exposure can be excessive, and components designed for interference fit make it impossible to remove studs when a bonnet is in place.

The existence of a leak does not justify the use of such potentially extreme measures.

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6730-96-2282 Page 2 4

There are many other factors that must be considered in order to make a responsible and timely decision. Considerations that are important in assessing leakage through pressure retaining bolted connections include: location of the leak. in the plant or system; time in the

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l plant cycle, leaking medium, materials exposed to the leak, Technical Specification limitations, ability to monitor or isolate the leak, and the ability to redirect or capture the leak.

j Numerous industry studies on the degradation and failure mechanisms of bolting in nuclear power plants have been documented. These studies have quantified the experience of bolting failures and identified the primary failure mechanisms associated with bolt degradation. The documents have shown that bolt failures have primarily occurred in pressurized water reactors, in both ambient and elevated temperature environments. The following three causes of bolting failures have been identified and have been evaluated for any possible impact at the Oyster Creek faci lity:

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1. Stress Corrosion Cracking (SCCh This mechanism requires a wet or humid i

environment, high preload stresses, use oflubricants containing molybdenum f

disulfide (MoS ), and/or improper heat treatment of material.

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2. Fatigue: This failure is primarily induced by improper preload' torquing.

3. Borated Water: This is a chemical attack caused by borated water leakage.

f GPU Nuclear has examined the conditions which are directly associated with the failure of l

bolts and evaluated their applicability to Oyster Creek. Records of operating history, j

maintenance procedures, Inservice inspection Program results, and material specifications for susceptibility to corrosion have been evaluated. GPU Nuclear has determined that the present scope of ASME XI NDE examination requirements for post bolted flange leakage to be undesirable when the likelihood of these failure modes is considered with the increase in personnel radiation exposure which would result.

1. SCC; The majority of bolting material installed at Oyster Creek meets ASTM A193, grade B7 specifications, except for the Control Rod Drive (CRD) bolts which are discussed below. This is a chromium-molybdenum material which is considered low strength and generally not susceptible to stress corrosion cracking. All bolting materials have been purchased in accordance with the GPU Nuclear Quality Assurance Plan.

4 Approved lubricants are controlled by procedures. The primary lubricant at Oyster Creek is Chesterton, a nickel based lubricant that does not contain MoS.

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6730-96-2282 Attachment I Page 3

2. Fatinue: Fasteners at the Oyster Creek site are typically torqued to a preload stress of 50% (15 %) of the yield strength. Exceeding this limit requires an Engineering Evaluation. This has been the standard practice at Oyster Creek, and is closely monitored by the Engineering Division.

3. Borated Water: Unlike pressurized water reactors, Oyster Creek does not use borated water in its primary coolant system. The reactor coolant system is frequently monitored for chemical composition and contaminants. No corrosion inducing additives are used or allowed. It is the GPU Nuclear position that chemical corrosion is not the cause of bolt failure in the Oyster Creek Class 1 systems.

Additionally, the atmosphere in the drywell during operation is required by Technical Specifications to be inerted with nitrogen. This starves the bolted connection of oxygen, mitigating the process of both chemical and stress corrosion cracking.

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4. CRD: CRD housing leakage has been primarily noted at Oyster Creek when the primary system was pressurized prior to heat-up and/or CRD Scram Time Testing.

- i This leakage drastically decreased when the vessel metal temperature reached the i

normal operating band and the gaskets and o-rings were properly seated by the -

required scram time tests. This change in leakage has been documented and evaluated by the vendor and found to be acceptable. Subsequent VT-1 examinations of the CRD bolts during normal maintenance evolutions have revealed no degradation caused by corrosion.

During the exchange of CRDs, the bolts are cleaned and ASME XI examinations are performed. GPU Nuclear utilizes these examinations as opportunities to evaluate the bolts for degradation. The sample of bolts that is inspected is a sufficient representation to allow identification of degradation trends. In previous refueling outages since the plant went on line in 1%9, there have been scheduled CRD exchanges. In the 27 years of operation, hundreds of CRD inspections have revealed no reports of CRD bolt failures due to corrosion. GPU Nuclear will continue to inspect the bolts during these periods of opportunity and will also employ alternative j

methods of examination if the need is justified. Although there is a small possibility that one of the eight CRD bolts might fail due to a design flaw, it is highly unlikely that a CRD would separate from its housing flange. As few as three uniformly distributed bolts can support full CRD loading while remaining within the stress limits identified by ASME codes.

The primary coolant system pressure test is done at reduced temperatures. It has been observed that the total amount of unidentified leakage in the drywell decreases significantly for the first few weeks of operation following a refueling outage. Inspections of drywell components made immediately after shutdown for refueling have repeatedly identified minimal or no leakage from the primary coolant system. The leakage found during the low temperature system pressure test is much greater than the leakage identified when the system is at normal operating temperatures.

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6730-E2282 Attachment I Page 4 ALTERNATIVE:

The source of all leakage detected by VT-2 examination during a system pressure test shall be evaluated to determine the susceptibility of the bolting to corrosion and potential failure. This evaluation will consider the following variables at a minimum:

1. Location ofleakage

2. History ofleakage

3. Fastener materials

4. Evidence of corrosion with the connection assembled

5. Corrosiveness of the process fluid

6. History and studies of similar fastener material in a similar environment

7. Other components in the vicinity that may be degraded due to the leakage When the evaluation of the above variables is concluded and the evaluation determines that the leaking condition has not degraded the fasteners, no further action is necessary. However, reasonable attempts to stop the leakage shall be taken.

1 If the evaluation of the variables above indicates the need for further evaluation, or no evaluation is performed, then a bolt in the leakage path will be removed. The bolt will receive a visual VT-1 examination, and be evaluated in accordance with IWB-3140, " Inservice Inspection Visual Examinations". This visual VT-1 examination may be deferred to the next l

l outage of sufficient duration if the evaluation supports continued service. When the removed bolting shows evidence of rejectable degradation, all remaining bolts shall be removed and receive a visual VT-1 examination and evaluation in accordance with IWB-3140.

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