ML20003C926

Responds to 810223 Request for Info Re Costs of Operating Facility W/Radioactive Crud Buildup & Reactor Vessel Embrittlement.No Cost Info Available.Forwards Pages from NRC 1979 Annual Rept Re Reactor Vessel Embrittlement

ML20003C926
Person / Time
Site:	Point Beach
Issue date:	03/03/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:	Mendl J WISCONSIN, STATE OF
References
NUDOCS 8103180808
Download: ML20003C926 (5)

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NUCLEAR REGULATORY COMMISSION y

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WASHINGTON, D. C. 20555 March 3, 1981 Cocket No. 50-266 t-g Oy)\\

4 Mr.-Jerry E. Mendl g,

Wisconsin Public Service Commission p$

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s, Hill Farms State Office Building (A '3 Madison, Wiscon?'n 53702 4

Dear Mr. Mend 1:

This is in response to your letter of February 23, 1981, wherein you asked about the possible costs of operating a facility with radioactive crud buildup and reactor vessel embrittlement. This response reflects the clarifications offered by both you and Mr. Peter Anderson of Wisconsin's Environmental Decade, Inc. in a conference call on February 23, 1981.

The question about crud stems from the knowledge that Commonwealth Edison Company has requested permission to decontaminate the reactor coolant system at their Dresden Nue' ear Power Station, Unit No.1 (a dual cycle boiling water reactor). Information related to this decontamination operation (including costs), and the reasons for it, are contained in NUREG 0686,

" Final Environmental Statement Related to Primary Cooling System Chemical Decontamination at Dresden Nuclear Power Station, Unit No.1" (copy enclosed).

As discussed in Section 2.3 of this document, the reason for the request for decontamination at Dresden-1 is a special case.

It should not be inferred from this that each nuclear plant will need to undergo this evolution after an operating period of about 20 years.

Such is not the case. The crud levels at Point Beach are typical for pressurized water reactors. Although the possibility of a full-scale chemical decontamination at Point Beach over its lifetitae cannot be ruled out, there is nothing to indicate at this point that it would be necessary. We should add that the steam generators at Surry and Turkey Point are going to be or have been replaced without the need for chemical decontamination of the entire reactor coolant system.

Rather, local decontamination has been used to reduce occupational radiation exposure. These measures would probably be employed by WEPC0 for either steam generator replacement or tube sleeving. No specific cost information is available, but the nature of the operation leads us to conclude that the cost is small.

As to your question about reactor vessel embrittlement, we learned from you that the question stems from a discussion of Unresolved Safety Issues presented in the NRC Annual Report for 1979. We have enclosed the pertinent pages of this report for your convenience. The origin of this issue and its plan for resolution are summarized therein. Point Beach Unit 1 is one of the 20 older operating plants included in this summary.

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- Generically, this issue is now satisfactorily resolved.

Its resciution will be issued as NUREG-0744 to be titled, "Rastlution of Reactor Yessel Material Toughness Safety Issue". This document is scheduled for publica-tion on April 15, 1981 for public comment.

As for Point Beach Unit 1, we fully expect that the pressure vessel will maintain adequate toughness and safety margins for the remaining life of the unit.

We have also reviewed, as requested, your draft response to Mr. Anderson's letter of February 2,1981 which you enclosed with your letter. We should point out that the San Onfre sleeving operation is not going as well as expected.

Southern California Edison is having difficulty ektaining a leak-tight joint between the top end of the sleeve and the t, team generator tube in certain areas. The problen is believed to be due to sludge. The Point Beach Unit 1 steam generators do not have a sludge problem of this magnitude, and therefore it may well be that this would not pose a problem for Wisconsin Electric Power Capany. Edison is now installing leak-limiting sleeves in areas where leak-tight sleeves have proved difficult.

Your proposed response A.3 (second paragraph) warrants minor comment.

In Appendix A to our Safety Evaluation Report of November 30,19/9, the NRC staff made two in-leakage calculations:

the first assumed that a crack existed in a steam generator tube in the mid-depth of the tube sheet (about 10 inches); the second, more conservative calculation assumed that a steam generator tube had a complete circumferential (" guillotine")

break 0.5" below the top cf the tubesheet.

Sinc rely, o_L

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Robert A. Clark, Chief Operating Reactors Branch #3 Divison of Licensing

Enclosures:

1.

NUREG-0686 2.

NRC Annual Report, pp. 75, 76

P00R ORIGINAL U.S. NUCLEAR REGULATORY COMMISSION 1979 1

Annual Report

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NRC staff's program for the resolution of generic issues

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under the DOE contract referred tu ab.ne. The staff l

The fundamental goal of Task A-Il is to provide an will carry out this additional nrk by contract with improved engineering meihud by which to assess the the Oak Ridge National Laboratory. Because of this i

safety margin in nuclear reactor pressure vessels and to problem, the schedule for completing Tack A !! has i

develop appropriate criteria for the evaluation of nor-slipped about one year, to December 1950.

4 mal, transient, or postu!.!cd accident conditions under the improved method. This method could then be used to provide such an assessment for those older Fracture Toughness and Potential for reactor pussure vessels that wdl eventually have Lamellar Tearing of PWR Stearn Generator marginal toughness according to the current method.

And Reactor Coolant Pump Supports Because relatively large amounts of prefracture plastic deformation on be expected at high temperatures During :he course of licensing review for a specific i

even in pnssure vessel steels of low toughness, the new Pressunzea water reactor (PWR) a number of ques-evahntion method will employ " elastic plastic" frac-tions were raised as to (1) the adequacy of the fracture ture rnechanics concepts. The basis for this improved toughness properties of the mateel used to fabricate methodolm ;s described in NUBEC-0311. "A Treat-the reactor coolant pump sepports and steam ment of tb ubject of Tearing bsMbility," developed generator supports, and (2) the potential for failure under an NRC-sponsored program at Washington due to lamellar tearing of these same supports. The Uriiversity. Additional Washington University work safety concern is that, although these supports are extending the methodology to reactor pressure vessels designed for worst case accident conditions.,%w frac-was funded bv the Decartment of Enertv. The ture toughness or lamcIlar teanng could caux the sup.

encineering m'ethod de(eloped will acco'u'nt for port to fail during such accidents. Support failure radiation.in' duced material degradation.

could conceivably impair the effectivenen of systems Task A-ll also includes or relies on piograms spon.

designed to mitigate the consequences of the accident, sored by the NRC Office of Nuclear Regulatory Re-An example of a postulated event sequence of potential search to provide: (1) an improved evaluation of concern would be a large pipe break in the reactor

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material degradation mechanisms resulting from coolant system which would severely load the sup.

neutron irradiation, and (b the development of im-ports, followed by a support failure of sufficient proved testing methods for use in determining the magnitude that a major wmponent such as a steam elastic-plastic properties of materiah.

generator would be displaced resultingin failure of the emergency e re e 1 ne system Since last } ear's report, the following has been ac-vide cooling water to the core, pipmg ne ded to pro-complished:

Because materials and designs similar to those of the

• Although delaved, an elastic plastic fracture test PWR originally revieped have been used in other method for r'outine determination of fracture plants, review of this issue was mcluded m the NRC P o ramjor Resolution of Cennic Issues as Cemne toughness was developed. Verification of the test f

x method is underway.

• The elastic-plastic iracture mechanics methods of A consultant was engaged to reassess the fracture NUREC-0311 were confirmed by work supported toughness of the steam generator and reactor coolant by an Electric Power Research Institute program, pump support materials for all operating PWR plants

" Methodology for Plastic Fracture."

and those in the later stages of operation license review. This reassessment included review of the

• The methods developed in these programs were materials utilized in the support of SS puentially af.

successfully used by NRC contractors to analyze fected PWRs. Based on the consultant's evaluation, it two pressure vessel burst tests reported in the was determined that there are 21 plants whose sup-Heavy Section Steel Technology Program, spon-ports are of questionable toughness and, accordingly, sored by the NRC Office of Nuclear Regulatory further detailed plant specific review is required. This

Research, decision concluded the generic study of this subject

• The potential for restoring by thermal annealing under Task A-12. During the plant. specific reviews the pressure vessel toughness lost by neutron that will follow, either the structural integrity of the radiation was shown to be impractical.

supports must be demonstrated, or measures to assure Significant delays have developed over the past year their structural integrity will be rerpiired.

A report describing the NRC staff's safety evalua-as a result of difficulties encountered in extending the tion and conclusions and describing its plans for imple-new engineering methodology to reactor pressure mentation (i.e., the more detailed plant-specific vessels. There is agreement among experts that the reviews referred to above) was issued for comment in methodology can be extended, but it will require a November 1979. It is entitled, " Potential for Low significantly greatu effort than that accomplished Fracture Toughness for Lamellar Tearing in PWR y,

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