ML20077D112
| ML20077D112 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/25/1994 |
| From: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C321-94-2384, NUDOCS 9412080059 | |
| Download: ML20077D112 (6) | |
Text
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s GPU Nuclear Corporation Nuclear
=st:388 Forked River. New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
November 25,1994 C321-94-2384 US. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 ASME Section XI Request for Relief R14 Ily letter dated October 25,1994, USNRC approved the Inservice Inspection (lW program for the Oyster Creek Nuclear Generating Station. This program was written to meet the 1986 edition of ASME XI, with no addenda. GPUN letter C321-94-2209 dated November 22,1994 was submitted to docket ASME relief request R13. The USNRC approved relief request R13 on November 23,1994. Subsequent to that approval, maintenance in the plant has revealed the need for additional relief.
Pursuant to 10 CFR 50.55a(g)(5)(iii), this letter is being written to request relief from specific requirements contained in ASME XI,1986 edition, Section IWA-5250(a)(2) due to the impracticality of compliance with the Code. This section of the Code refers to corrective action requirements relating to leakage from bolted connections. The details and justification of this request are contained in Attachments I and 11 to this letter. As further assurance of stud integrity, GPUN is pursuing the tooling and expertise to proof test, if practicable, a stud on each valve by subjecting the stud to a force equal to 70% of yield (30% above normal operating stress) with a tensioner.
As restart from refueling outage 15R is presently scheduled for November 27,1994, expedited approval of this relief is requested by November 25,1994. If any further information or assistance is required, please contact Mr. John Rogers of my staff at 609.971.4893.
Jol a J. Ba m
' ce President and Director CSOC03 Oyster Creek JJil/JJR j
Attachment cc:
Administrator, Region I 94120e0059 941125 8
Senior Resident Inspector f
"O" 0500 $ 9 f
,I Oyster Creek NRC Project Manager GPU Nuclear Corporahon is a subsday ct Genera! Pubac UhWes COWo'abon j
t ATTACHMENT I REl.IEF REQUEST R14 r
CODE
REFERENCE:
ASME Section XI 1986 edition, without addenda; IWA-5250(a)(2), Corrective Measures CODE REQUIREMENT:
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 CODE RELIEF REQUEST:
o For leakage found on Class I system bolted connections during the performance of the Class I system pressure test in refueling outage 15R, relief is requested as specified for the following bolted connections:
A.
Main Steam Isolation Valve V-1-7, Body to Bonnet B.
Mr.in Steam Isolation Valve V-1-8, Body to Bonnet PROPOSED ALTERNATIVE EXAMINATION FOR REQUESTS:
A visual inspection of the exposed portion of the bolting above and below the raised face flange, and an ultrasonic inspection of the bolting in the area of wetting have been perfomled.
Additionally, visual inspections of the adjacent valve body and bonnet have been performed to detect any indication of corrosion in the leakage path.
BASIS FOR RELIEF REQUEST During the performance of the system pressure test at the end of the current refueling outage 15R, leakage was detected from the bolted body to bonnet flange on main steam isolation valves V-1-7 and V-1-8. As committed in the original relief request, R13, an attempt was made to remove a bolt (more precisely a stud) from the path of leakage and visually inspect it for corrosion. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of effort and an expenditure of over 350 mrem of exposure, only a single nut could be removed from one stud on each valve in question with some minor damage to one stud. The stud on each valve could not be moved. Repeated attempts to loosen the stud on valve V-1-7 met with failure and funher efforts will in all probability result in damage to the stud with the resultant loss of capability to perform the required inspection.
Valve V-1-7 had prcviously been completely disassembled in refueling outage 13R and all 24 studs were scheduled for replacement as a target of opportunity. Even when all insulation, upper valve parts, and interference had been removed, only two of the 24 studs could be removed and replaced. The studs in question are not typically removed. Additionally, the l
thread engagement of these studs is about four inches into the valve body which is in excess of the standard thread engagement.
i j
l 1
C321-94-2384 Attachment I Page 2 An identical main steam isolation valve was disassembled during the present refueling outage.
No corrosion was found on any of the removed studs. However, removal of the 24 studs resulted in a damaged stud hole, seven damaged nuts, and eighteen studs being mechanically damaged, discarded, and replaced.
The intent of the ASME inspection is to: 1) detect corrosion; and 2) verify the integrity of the mechanical flange. The reasons why corrosion is not expected at Oyster Creek were docketed to the USNRC in GPUN letter C321-93-2041 dated February 4,1994. That methodology has been summarized and included as Attachment II to this letter. This lack of corrosion at Oyster Creek has been once again validated this outage, as no Code inspection performed on any Code component re.aled any corrosion in excess of that allowed by ASME Section XI,1986 edition, paragraph IWA-3100.
It is worthy of note that although these two valves have had leakage at the end of the last two refueling outages (while the systems wt at less than operating temperature), drywell inspections conducted at the beginning and end of the last two outages (at operating pressure and temperature) have revealed no leaks. Should the leakage develop during the next cycle, the resultant increase would be measured in the unidentified drywell leak rate which is limited by Technical Specifications. Exceeding the limit will result in a plant shutdown.
Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested to not removc a stud for inspection on the two referenced main steam isolation valves due to the impracticality of bolt removal. Any further attempts to remove the bolts in the leak path would in all probability result in damage to the studs and valve body. In lieu of stud removal, a visual inspection of the exposed threads, both above and below the raised face flange have been performed. An additional UT along the length of the stud has also be performed to check for excessive wastage or damage to the stud caused by the attempted removal Finally, a visual inspection of the valve body and bonnet in the path of the leakage has been performed to detect any indication of corrosion.
SUMMARY
GPUN requests that the specified relief requested from the Code requirements be granted, as i
the major factors which can result in bolt corrosion discussed in GPUN letter C321-93-2041 dated February 4,1993 (and summarized in Attachment Il to this letter) are: 1) not applicable to Oyster Creek (such as chemically accelerated corrosion); 2) under control by existing procurement and examination programs (such as the quality assurance requirements for procurement and the existing ISI requirements for examination); 3) not existent at Oyster Creek as evidenced by extensive operating history, controlled maintenance procedures, and Inservice Inspection records compiled since 1969. Incidentally, during this outage visual inspection of two cap screws from flanges in the recirculation system have revealed no corrosion in excess of ASME limits.
.l
~
C321-94-2384 Attachment I Page 3 Addition $ illy, unsuccessful attempts were made to comply with ASME Section XI,1989 edition,1990 Addenda. Further attempts would be impractical and would in all probability result in damage to the studs and valve body while needlessly increasing the radiological exposure of the technicians with no increase in the integrity of the reactor pressure boundary.
l
)
I
ATTACHMENT II 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. These documents have shown that bolt failures have primarily occurred in pressurized water reactors (PWR), at both ambient and elevated temperature environments. The following thrse causes of bolting failures have been identified and have been evaluated for any possible impact at the Oyster Creek facility.
1.
Stress Corrosion Cracking (SCC): This mechanism requires a wet or humid environment, high preload stresses, use of lubricants containing molybdenum disulfide (MoS ), and/or improper heat treatment of material.
2 2.
Fatigue: This failure is primarily induced by improper preload torquing.
3.
Ilorated-Water: This is a chemical attack caused by borated water leakage.
GPUN has examined the conditions which are directly associated with the failure of bolts and evaluated their applicability to Oyster Creek. Records of operating history, maintenance procedures, Inservice Inspection program results, and material specifications for susceptibility to corrosion have been evaluated. GPUN 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 A 193, grade B7 specifications. This is a chromium-molybdenum material which is considered generally not susceptible to stress corrosion cracking. All bolting materials have been purchased under nuclear quality control program or have been tested in accordance with the GPUN QA Plan.
Approved lubricants are controlled by procedures. The primary lubricant at Oyster Creek is Chesterton, a nickel-based lubricant that does not contain MoS.
2 2.
Fatigue: Fasteners at the Oyster Creek Site are torqued to preload stresses of less than 50% of the yield strength. This has been the standard practice at Oyster Creek, and is closely monitored by the Plant Engineering Department, Mechanical Section.
C321-94-2384 Attachment II Page 2 3.'
nnrated Wnten Unlike pressurized water reactore, Oyster Creek does not use borated water in its primary coolant system.. The reactor coolant system is pure demineralized water and is frequently monitored for chemical composition and contaminants. No corrosion inducing additives are used or allowed. It is the GPUN position that chemical corrosion is not a cause of bolt failure in the Oyster Creek Class 1 systems.
4.
Envirnnment-The atmosphere in the drywell during operation is required by Technical Specifications to be inerted with nitrogen. This starves the bolted connections of oxygen, mitigating the process of both chemical corrosion and stress corrosion cracking.
SUMMARY
The major factors which can result in bolt corrosion are: 1) not applicable to Oyster Creek (such as chemically accelerated corrosion); 2) under control by existing procurement and examination programs (such as the quality assurance requirements for procurement and the existing ISI requirements for examination); and 3) not existent at Oyster Creek as evidenced by extensive operating history, controlled maintenance procedures, and Inservice Inspection records compiled since 1969.
.