ML20046A647
| ML20046A647 | |
| Person / Time | |
|---|---|
| Issue date: | 06/07/1993 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Willis G GREENPEACE INTERNATIONAL |
| Shared Package | |
| ML20046A648 | List: |
| References | |
| 2.206, NUDOCS 9307290164 | |
| Download: ML20046A647 (8) | |
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June 7,1993 (10 C.F.R. S 2.206)
Mr. John Willis, Coordinator Nuclear Campaign Greenpeace International 1436 U Street, N.W.
Washington, D.C.
20009
Dear Mr. Willis:
On behalf of Greenpeace International (petitioner) you wrote on March 24, 1993 to the Chairman of the Nuclear Regulatory Commission (NRC) regarding all pressurized-water reactors (PWRs) now operating in the United States.
You request immediate, full inspection of all vessel head penetrations (VHPs) in U.S. PWRs for cracking and publication of the results by-the NRC.
Because you also. request that the NRC. shut down affected reactors, whether the cracking is longitudinal or circumferential, your letter has been referred to the NRC staff for consideration as a petition for enforcement action under 10 CFR 5 2.206.
The
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petitioner further requests that the NRC staff "re-license" reactors which must be closed due to VHP cracking based.on'the r
assertion that repair or mitigation program for such cracks may negatively affect the configuration and effectiveness of safety systems.
The petitioner seeks relief based on allegations that (1) some cracking has been identified in VHPs in-PWRs in. France, Belgium, Switzerland, and Sweden; (2) testing in France revealed incipient.
circumferential cracking of.some VHPs, which could lead to a through-wall break in the primary' pressure boundary without fulfillment of the leak-before-break criterion; and (3) this phenomenon could cause the ejectionof the control rod drive mechanism, with resulting loss of control of the reactor..The bases for the petitioner's request are described-in more detail in " Vessel. Head Penetration Cracking in Nuclear Reactors,"
Greenpeace International and Greenpeace Sweden,-March 1993, which; is attached to the petition.
With regard.to the request for immediate inspections, the NRC staff understands the petitioner's concern for obtaining information on primary water stress _ corrosion cracking : (PWSCC) through' inspection.of VHPs in U.S. PWRs; the.NRC agrees that data developed by inspections will be needed to confirm'the current understanding and evaluation.of this issue.
However, immediate inspection of all VHPs in U.S. PWRs is not now warranted.
This conclusion is based on evaluations of the safety significance of the issue and the potential negative impact of performing I
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Mr. John Willis t071EB inspections before the inspection methods have been fully developed to ensure effectiveness of inspections and to minimize
. personnel exposure, as described below.
Accordingly, your request for immediate relief is denied.
The staff identified PWSCC as an emerging technical issue to the Commission in 1990, after cracking was noted.in pressurizer heater sleeve penetrations at a domestic PWR facility.
At that time,.the staff reviewed the safety significance of the cracking as well as the repair and replacement activities at the affected facility.,
The staff determined that the safety significance of the cracking-was low because the cracks were axial, had a. low growth rate,.and were in a material with a high flaw tolerance (high fracture toughness).
These factors demonstrate that any cracking would result in a leak, well before a possible penetration break.
Nevertheless, the staff issued Information Notice 90-10, February 23, 1990, to inform the industry of the issue.
In addition, the staff met with the Combustion Engineering Owners
Group (CEOG) in February 1990 to discuss a program CEOG' initiated in January 1990 to assess the potential.for, and thefeffects of, PWSCC of susceptible Alloy 600 components in the reactor coolant' pressure boundary.
CEOG also established an inspection program, developed repair methods, and evaluated the effects of reactor:
coolant leakage, including the potential for erosion /corrosionLand wastage of carbon steel vessels from the boric acid contained in the reactor coolant.
Through this program, CEOG. evaluated several instances of. cracking that.had occurred in small pressurizer instrumentation nozzles made of Alloy 600.
CEOG. submitted the detailed findings of the program to the: staff in a proprietary report on February 26, 1992.
The conclusions of the report, in part, follows:
(1)
Circumferential cracking of the heater-sleeves and the instrumentation nozzles is not a credible failure mode because the cracking is due to hoop stresses that will cause c
cracks to have predominantly axial orientations.
(2)
Postulated axial cracks 2 inches in length, which are longer than any cracks observed in the field,- will not exhibit'
. unstable crack growth.
Some PWSCC'may. continue,.which could-
. result in increased gradual leakage with time that can be detected by visual inspection..
(3)
Visual inspection :is the best method _ for detecting a leaking.
sleeve or nozzle,.or for detecting. damage to the pressurizer shell as a result of-boric acid' corrosion, and' scheduled.
detailed'visualiinspection of-the pressurizer lower head should. continue at a fixed interval.
TheLinspection interval
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was determined on the basis of: experimental results from the program.
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Mr. John Willis AM 07'1El The staff has reviewed the report, and finds that its results and the recommended inspections, coupled with field experience, provide a sufficient basis to conclude that loss of structural integrity and ejection of components with respect to pressurizers are highly unlikely.
In December 1991, after cracks were found in a control rod' drive mechanism (CRDM) penetration in the reactor head at_a-French plant, an action plan was implemented to address PWSCCEat all U.S.
PWRs.
The staff met with the Westinghouse Owners' Group (WOG) on' January 7, 1992, the CEOG on March 25, 1992,. and the; Babcock & Wilcox Owners' Group on May 12, 1992, to discussLtheir respective programs ~for. investigating PWSCC of Alloy.600 and to assess the possibility af cracking of CRDM penetrations in their respective plants.
Subsequently, the staff asked the Nuclear.
Utility Management and Resources Council (NUMARC) to coordinate future industry actions because the issue was applicable to~all' PWRs.
Meetings were held with NUMARC and PWR owners on the' issue on August 18 and' November 20, 1992, and March 3, 1993..~ Summaries of the meetings are available in the Commission's Public Document
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Room, 2120 L Street, N.W.,
Washington, D.C.
20555.
In. addition, the Electric. Power Research Institute (EPRI)' is engaging in ongoing research on methods for PWSCC mitigation.. EPRI is also developing a demonstration program to ensure.that. inspections performed on'CRDM penetrations will be highly allable in detecting and measuring flaws because the methods'and examiners will have been tested and qualified.
The WOG conducted a safety analysis on Alloy 600 CRDM penetration cracking to support continued operation of its plants.
That safety analysis, WCAP-13565, was issued in. December 1992 and revised in February 1993..The Westinghouse analysis concludes, and the staff agrees, that,the available data do:not now present a significant safety concern for-Westinghouse plants.
During the meeting on March 3, 1993, the WOG reported that one additional small flaw, 3-mm (0.120-in.) long and-2.25-mm (0.090:in.)? deep,.
was found through metallographic examination of the penetration removed from the French plant.
This indication was:notl axially oriented.
It was oriented about 30 degrees from the horizontal.
The WOG performed an analysir and evaluation of this flaw and-concluded that the stresses in the region of thefflaw areLlow;-
therefore, growth would not be expected..
The-. staff agrees with the conclusions of.this analysis..The staff requested that:the
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WOG supplement its safety evaluation with an assessment of the rationale for the proposed _ inspection schedule and an~ assessment' of the availableLor additional leak detection systems to ensure technical specifications requirements are met.
The-other'PWR owners ': groups.are expected to' submit separate safety-evaluations-for their plants in the near future.
The staff will complete its L
Mr. John Willis VWi07 nil review of the inspection schedule and will issue a safety evaluation following the receipt of the submittals discussed above and will provide the petitioner a copy.
From the review of the information presented in the meetings discussed above, and the review of the WOG and CEOG reports, the staff concludes that CRDM cracking is not a significant safety i
issue at this time.
Nonetheless, VHP inspections should be l
performed in order to verify the analytical evaluations and to confirm that the cracking phenomena, if it is occurring in U.S.
plants, is consistent with the experience in foreign plants.
New information and new events, however, could cause a reassessment of its safety significance of the cracking.
VHP inspections should be conducted in a carefully planned manner.
The staff has safety concerns related to t~rher exposure associated with inspections.
Field experience in foreign countries has shown that personnel radiation exposure could be significantly reduced in a well-pla..ned inspection program, including the use of remotely controlled or automatic equipment.
Currently, the U.S.
industry is developing such equipment for inspection and repair.
Accordingly, immediate inspections could result in the failure to minimize worker exposures.
The currently proposed inspection schedule calls for domestic PWR inspections to begin in the early part of 1994.
This schedule will provide confirmation of existing safety evaluations while providing adequate time to develop and plan inspection methods that will minimize personnel exposure.
The schedule for inspeccions and other industry activities appear in the March 1993 meeting summary.
In summary, the staff concludes that PWSCC in U.S. PWRs does not now represent an undue hazard to public health and safety on the basis of the results from the analyses performed, coupled with the results of the field experience from inspections performed worldwide.
The NRC will assure that VHP inspections to confirm these conclusions are performed in a reasonable time frame and that additional data and analyses are developed and corrective actions taken, as needed, to demonstrate adequate protection of public health and safety over the entire term of licensed operation of U.S.
PWRs.
Based on the low probability of CRDM penetration failure, the low safety significance of CRDM leakage, and in view of the potential for reducing worker exposure to radiation, as set forth above, there is sufficient time available to establish and implement a well-thought-out and well-planned inspection, evaluation, and repair program if needed.
Accordingly, the petitioner's request for immediate relief is denied.
L-t Mr. John Willis-
, AN 07 IEF l
The staff will review your petition in accordance with 10 CFR S 2.206.
I will issue aEfinal decision with regard to your; petition within a. reasonable time.
A copy of the: notice that is being filed for publication with the' Office of the Federal Register;is enclosed for-your information.
Sincerely, OPYgTnal' igned by s
ThomasE. Earle7 Thomas E. Murley,. Director Office of Nuclear Reactor Regulation,
Enclosure:
As stated cc:
All PWR Licensees i
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Rasin, NUMARC J.
Taylor, EPRI L.A. Walsh,-WOG J.'Hutchinson,'CEOG J.
Taylor, BWOG DISTRIBUTION Central File /PDR JTaylor JGoldberg, OGC EDO RF GT8737-JSniezek RWeisman, OGC TMurley HThompson JStrosnider FMiraglia JBlaha WRussell EBeckjord, RES.
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Mr. John Willis The staff will review your petition in accordance with 10 CFR S 2.206.
I will issue a final decision with regard to your petition within a reasonable time.
A copy of the notice that is being filed for publication with the Office of the Federal Register is enclosed for your information.
Sincerely, Thomas E. Murley, Director Office of Nuclear Reactor Re lation
Enclosure:
Notice cc:
All PWR Licenses NUMARC EPRI WOG CEOG BWOG pISTRIBUTION Central File /PDR JTaylor JGoldberg, OGC EDO RF GT8737 JSniezek RWeisman, OGC TMurley HThompson JStrosnider-FMiraglia JBlaha WKoo 1
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Mr. John Willis In summary, the' staff concludes that PWSCC in U.S. PWRs does not now represent an undue hazard to public health and safety on the basis of the results from the analyses performed, coupled with the results of the field experience from inspections performed worldwide.
The NRC will assure phat VHP inspections to confirm these conclusions are perfor ed in a reasonable time frame and that additional data and an yses are developed and corrective actions
- taken, as n eded, to demonstrate adequate protection of public health nd safety over the entire term of licensed operation of U.S.
WRs.
Based on the low probability of CRDM penetration fai re, the low safety significance of CRDM leakage, and in view of the potential for reducing worker exposure to ra ation, as set forth above, there is sufficient time availa e to establish and implement a well-thought-out and well-p1 nned inspection, evaluation, and repair program if needed.
Accordingly, the petitioner's request for immediate relief s denied.
The staff will review your petition in a cordance with 10 CFR S 2.206.
I will issue a final decisi with regard to your petition within a reasonable time.
A opy of the notice that is being filed for pub]ication with t e Office of the Federal Register is enclosed for your inform tion.
Sin erely,
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T omas E. Murley, DiTect,or ffice of Nuclear Reactor l
Regulation J
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cc:
NUMARC CEOG BWOG DISTRIBUTION Central File /PDR JTaylor JGoldberg, OGC EDO RF GT8737 USniezek RWeisman, OGC TMurley
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Mr. John Willis In summary, the staff concludes that PWSCC in U.S.
PWR does not now represent an undue hazard to public health and saf.ty'on the basis of the resuzts from the analyses performed, coupl d with the results of the field experience from-inspections performed worldwide.
The NRC will assure that VHP inspections to confirm these conclusions are performed in a reasonable time fr me and that additional data and analyses are developed and correo_ive actions taken, as needed, to demonstrate adequate protectp'on of public health and safety over the entire term of licensed operation of U.S.
PWR.
Based on the low probability of C M penetration failure, the low safety significance of CRDM leak ge, and in view of the potential for reducing worker exposure to adiation, as set forth above, there is sufficient time available to establish and implement a
well-thought-out and well-pl ned inspection, evaluation, and repair program if needed.
Accordingly, the petitioner's request for immediate relief is denied.
The staff will review your petition in accordance with 10 CFR S 2.206.
I will issue a final decision with regard to your petition within a reasonable time.
A, copy of the notice that is being filed for publication with the' Of fice of the Federal Register is enclosed for your informat'/op.
/
Sincere
//
Thomary' E. Murley, Director Of fice of Nuclear Reactor Regulation cc:
All PWR Licenses EPRI WOG NUMARC CEOG BWOG lt DISTRIBUTION
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Central File /PDR JTaylor/
JGoldberg, OGC EDO RF GT8737 JSnie'zsk RWeisman, OGC TMurley HThompson JStrosnider FMiraglia JBlaha WKoo WRussell EBeckjord, RES JPartlow l
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