ML25196A331

03-11-76 Report on Donald C. Cook Nuclear Plant Unit No. 1

ML25196A331
Person / Time
Site:	Cook
Issue date:	03/11/1976
From:	Moeller D Advisory Committee on Reactor Safeguards
To:	Anders W NRC/Chairman
References
Download: ML25196A331 (1)

Text

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 11, 1976 Honorable William A. Anders Chairman

u. s. Nuclear Regulatory Conmission washington, OC 20555 SUBJECI':

REPORT 00 DOOALD C. COCl{ NOCLFAR PIANT UNIT H>. 1

Dear Mr. Anders:

During its 190th meeting, February 5-7, 1976, and its 191st meeting, March 4-6, 1976, the.Advisory Committee on Reactor Safeguards canpleted a review of the proposal to increase the maximum operating power level of the D:>nald c.

Cook Nuclear Plant Unit l'-k>. 1 from 81% of rated power to the rated power of 3250 MWt.

'lhe Cornmittee has previously discussed this project in its reports of ~cember 13, 1968, and Cctober 17, 1973. A Subcamtittee meeting on the current proposal was held in Washington, D. C., on February 4, 1976.

During its review, the Committee had the benefit of discussions with repre-sentatives of the Indiana and Michigan :R>wer Company, American Electric Fower Service Corporation, t-estinghouse Electric Corporation1 and the Nuclear Regulatory COmmission (NRC) Staff. 'lhe Conmittee also had the benefit of the documents listed.

In its report of Cctober 17, 1973, the ACRS reconmended that a continuing series of monitoring and evaluation measures be applied to this first-of-a-kind ice condenser S}'Stem employed as part of the containment for Ibnald C. Cook Nuclear Plant Unit N:>. 1. 'lhe Applicant has conducted frequent periodic surveillance, including sample weighing of ice baskets.

'!he results of this surveillance program indicate a larger rate of ice loss than previously anticipated and a nontmiform rate of ice loss across ice bays and among ice bays. Baskets adjacent to walls appear to be subject to increased rates of ice loss~ however, such baskets are relatively inaccessible for weighing, and efforts to date have not led to an accurate method for measuring their weights.

'1he Conmittee believes that the current inventory of ice is acceptable.

However, the COO""ittee believes that further analysis and evaluation are required to asce1.tain the minimum acceptable basket weights, particularly in wall baskets, and to establish sampling and measurement methods *which are adequate to assure that these limits have been met. '1he ACRS believes that for the next year an appropriate sampling program, performed every three or four months as is currently planned by the Applicant, should provide suitable assurance of an adequate inventory while an improved basis for setting the inventory limits and sampling procedures is devel-oped. 'Ihe Cornmi ttee wishes to be kept informed with regard to this matter.

287

Honorable William March 11, 1976 In its re:PC)rt of O::tober 17, 1973, the ACRS also recamnended that the Regulatory Staff independently evaluate ice condenser behavior under

POStulated accident conditions. '.Ihe 1-U~ Staff reported that such confir-mation has not yet been accomplished but that they expect to have such a capability within the year.

'!he Applicant reported the results of a program of verification of the adequacy of the axial power distribution monitoring system (APil-tS) for steady state operation and for several transient power conditions. '.Ihe experience has been favorable.

'!he Applicant reported that the LOCA-ECCS analysis, on the basis of the Westinghouse evaluation model approved in March, 1975, would limit the maximum allowable nuclear peaking factor (Fq) to 1.84 without allowance for rod bowing at rated power. Westinghouse has recently p:roposed several changes in the evaluation model and, in a modified form., these have been tentatively accepted by the NRC Staff. ~en the D:>nald C. Cook Nuclear Plant Ulit 1-b. 1 is analyzed for operation covering only the remainder of the first fuel cycle, and when as-built parameters are used together with the modified changes in the evaluation model, a maximum Fq of 1.98 (plus an additional factor of about 0.08 for bowing) is calculated to correspond to the limits of Appendix K'of 10 CFR 50 at rated power.

In view of the favorable experience with APU.tS and the expected operation of IX>nald c. Cook Nuclear Plant Ulit 1-b. 1 in a base loaded fashion for the renainder of the first fuel cycle, the Conmittee believes it acceptable for this plant to be operated at powers up to rated power for the first fuel cycle. '.Ibis matter should be reviewed by the NRC Staff and the ACRS prior to operation at rated power with the next fuel load.

'!he ACRS reiterates its recommendations of O::tober 17, 1973, that considera-tion be given to the possibility of improvements in ECCS effectiveness to provide additional margin.

'!he NRC Staff is evaluating, on a generic basis, the adequacy of reactor vessel support sys tens.

Some aspects of reactor vessel cavity pressure loadings during postulated accidents may also warrant reexamination in connection with the application of the results of the generic study to IX>nald c. Cook Nuclear Plant Ulit No. 1. '.Ibis matter should be resolved in a manner satisfactory to the NRC Staff.

'!he Ccxrmittee recamnends that consideration of cumulative downtime of safety-related canponents and systems be included in the continuing developnent of technical specifications for D:>nald C. Cook Nuclear Plant Ulit 1-b. 1.

288

lbnorable William March 11, 1976 1he camdttee recamnends that continuing evaluation be made of the possi-bility of further enhancing the provisions for industrial security.

1he NRC staff has not yet canpleted its generic evaluation of possible nooifications in requirements for protection against fires in light water power reactors. 1he Conmittee wishes to be kept informed of the results of application of this stooy to Ik>nald c. Cook Nuclear Plant *01it l'i>. 1.

Generic problems relating to large water reactors are discussed in the Conmittee's report dated March 12, 1975. 1hese problems should be dealt with appropriately by the NRC Staff and the Applicant.

1he ACRS believes that, subject to the foregoing and to matters discussed in its report of O.::tober 16, 1973, the Ik>nald C. Cook Nuclear Plant Unit l'i>. I can be operated at powers up to rated power for the first fuel cycle.

'!his matter should be reviewed by the NOC Staff and the ACRS prior to opera-tion at rated power with the next fuel load.

Mditional comments by Ors. David Okrent and Milton Plesset, Dr. Herbert Isbin, and Mr. Myer Bender are presented on the following pages.

Sincerely yours,

~w,~~

Dade w. K>eller Olairman 289

Honorable William March 11, 1976 Additional Comments by Drs. D:lvid Okrent and Milton Plesset For several reasons, we do not concur with Jle recarmendation to penni t IDnald C. Cook Nuclear Plant tmit No. 1 to opPrate at rated power at this time.

First, while there may be merit in the proposed changes in the westirig-house evaluation model, we believe further examination is warranted of several factors, including the scaling of experiments, the scatter in data, and the possible influence of super-plasticity on clad behavior during postulated loss-of-coolant accidents. our reluctance to endorse these changes is also due, in large part, to signs of a continued process of cutting into the conservatisms built into the original evaluation models, without a concanitant buildup in our basic understanding or predictive ability for the overall I.OCA-ECCS process.

In this situation there are limits beyond which the use of best estimate heat tran.-3fer coefficients, etc., is no longer appropriate.

Second, even with application of the revised westinghouse evaluation model which has been judged acceptable by the NRC Staff, I'.bnald C. Cook Nuclear Plant Unit No. 1 requires a LOCA - limited maximum peaking factor (Fo) of l.98

{plus the margin for bowing) at rated power. 'Nlile this is somewhat nigher than the FQ which can be expected at steady operation for the rest of the first fuel cycle for n:mald c. Cook Nuclear Plant tmit No. 1, it still repre-sents a very large reduction in the margin that has been available for most plants between LOCA - limited FQ and that value which would be present most of the time. ~is margin has been eroded until it is a small fraction of its earlier values. Furthermore, if we accept this low FQvalue for D:>nald c.

Cook Nuclear Plant tmit No. 1, a precedent will be set by means of which all ~*swill be able to reduce what was a substantial safety margin only a few years ago. ~is previously available substantial safety margin could cover many of the existing uncertainties in the analysis of I.OCA-ECCS. ~e uncer*tainty aspect is highlighted by the less than perfect record obtained by the experts in their pre-prediction of various separate effects experiments, by the recognized difficulties in a calculation fran first principles, by the current tmavailabil;ity of experiments to test all relevant effects, and by the lack of a meaningful test of 'rkstinghouse predictive capability with experiment.

'lhird, the ACRS has in the past been reluctant to accept proposed operation of reactors with Fo's less than 2.2.

In part, such caution arose fran the knowledge that, wifh a more flattened power distribution, a much larger fraction of the fuel elements would be at or near peak temperatures, given a U:x::A, and therefore potentially vulnerable to an "anomaly" in ECCS function

.;uch as some three-dimensional flow effect or excessive steam generator leakage).

290

Honorable William March 11, 1976 Fourth, we find an absence of effort on the part of the Applicant to develop possible improvements in ECCS capability and reliability and too small an effort by the nuclear industry or the me safety research program itself in this regard.

We believe that, while experience with the API:r1S is favorable, it would be prudent to permit operation of JXmald c. Cook Nuclear Plant Unit :t-b. 1 at

• power levels only up to 90% or 92% of rated power during the remainder of the first fuel cycle. During this period, further infonna.tion should become available on IOCA-ECCS, on the response of more highly flattened cores to anomalous ECCS ftmction, and also on ice-condenser behavior. a;rually impor-tant, a precedent for operation with such low Fq values would not be set, and the industry might be encouraged to develop improvements in ECCS, partic-ularly with regard to reflooding rates and an insensitivity to steam binding, which the ACRS in many reports and the AOC Commissioners, themselves, in their decision on the ECCS Acceptance Criteria, have urged.

291

Honorable William March 11, 1976 Additional Comments by Dr. Herbert S. Isbin Although I concur with my colleagues that the Applicant can operate the D:mald C. Cook Nuclear Plant Unit r<<>. 1 at fUll power and meet the require-ments set forth by the NRC Staff's evaluation of the I.OCA-ECCS analyses, I suggest restricting the operations to a limit less than 100% power. Fur-thetmore., I believe that though this action may place an unusual burden upon this Applicant, the contributions to be made will, in the long range, advance safety in the nuclear industry.

My reconmendation is to arbitrarily restrict the power level to about 90% and is based upon the following factors:

I. '!be current operating limit is 81% and the proposed increase in power is limited to the present fuel which will extend the operations to about the end of this year. A new analysis will be needed for the second fuel loading which is to be* supplied by a new fuel supplier. '!be ACRS has recommended that fuel vendors provide independent overall analyses for reload cores, and it would be expected that the Applicant will suanit new eval-uations for the continuing operations. Additionally, the ACRS recamnended in its O::tober 17, 1973 report on r:x:>nald C. Cook Nuclear Plant unit r<<>. 1 that the NRC Staff independently evaluate the ice condenser behavior under postulated accident.

conditions. 'Ibis capability is expected to be achieved by the end of this year. '!bus, by the end of the year, two additional efforts for the evaluation of the I.OCA-ECCS phenanena will have been canpleted and the expected thoroughness of these stooies in determining the sensitivity of the various parameters.should 100re firmly establish the confidence to be placed in the analyses.

2. '!be Ice Condenser is a new concept and the I:X>nald C. Cook Nuclear Plant unit No. 1 is the first plant of its kind in operation. '!be Applicant has demonstrated competent management in resolving prob-leIDS that have arisen. Continued operation for the remainder of the first cycle, at a limit less*than full power, provides sane additional safety margin while th~ Applicant, through its proposed monitoring of the ice inventory, develops a suitable long-term management program.

3. '!be ACRS also suggested in its O::tober 17, 1973 report that con-sideration be given to the possibility of ing;>rovements in ECCS effectiveness to provide additional margin. '!be Applica~t effec-tively responded to all other ACRS recamnendations, but not to this item.

I believe that this request is reasonable and should be carried out.

292

Honorable William March 11, 1976

4. Generic problems identified by the NR: Staff and by b'le ACRS have received attention for the review of this.application and, for justifiable reasons, action and resolutions to be applied can be deferred1 however, in my opinion, an application that seeks to go fran a limi1:ed power to full power should contribute in sane meas-urable way to the canpletion of. sane of these generic problems that are dependent upon analyses to be suanitted by the Applicant or to definitions of programs that must.be completed for the resolutions.

s.

As noted in the ACRS letter, the Applicant's experiences with the APa-tS for steady state operation and for several transient power conditions have been favorable. 'lhe proposed maximum nuclear peaking factor at full power would be the lowest value for a pressurized water reactor. In this respect, too, IX>nald. c. Cook Nuclear Plant unit No. 1 becomes the lead plant in the united States to formalize such control and 100nitoring procedures.

In my judgment, it would be prudent to per-mit a relaxation of the 81% power limit and to acquire more experience for the remainder of this first core life, but at sane power limit intermediate between 81% and 100%.

293

lk>norable William March 11, 1976 Mditional Conments by Mr. Myer Bender Normal design practice for any process design is to use bounding values for sizing equipnent. It is, therefore, conventional as required by Appendix K of 10 CFR 50 to base ECCS design on the system requirements for the worst IOCA combined with extreme core performance conditions, including the lim-iting peaking factors as established by nuclear reactivity control and fuel element performance variables. However, the likelihood of a IOCA involving the bounding conditions related to the *warst pipe break during the period when the fuel and reactivity control conditions are at their design limits is probably orders of magnittrle less than the likelihood of a IOCA which requires a reliable actuation of the emergency core cooling system and is undoubtedly well below the probability of 10-6 events per r.eactor year that would exceed the limits of 10 CFR 100. Consequently, showing the ability to meet all the requirements of Appendix K of 10- CFR 50, while a useful analytical exercise, should not be a governing consideration in determining whether the few early design versions of nuclear power stations that were granted a construction license prior to the pranulgation of Appendix K of 10 CFR 50 should be operated at their rated power.

'Ihe interests of public safety are better served if those responsible for the operation of the pre-Appendix K of 10 CFR 50 designed power plants demonstrate an attentive interest in assuring that early vintage plants are not operated in a mode that encroaches on the limiting operating con-ditions that challenge the ECCS capacity for any significant portion of the plant's operating life. 'Ihe base load operating plans for IX>nald C.

Cook Nuclear Plant Unit l'b. 1 at rated power are thus consistent with the public safety interest for this pre-Appendix K of 10 CFR 50 vintage design.

'Ihe D:>nald C. Cook Nuclear Plant Unit l'b. 1 should be permitted to operate in this mode at rated power so that the using public can have maximum value fran this energy resource.

'Ihe emphasis on ECCS improvements for pre-Appendix K of 10 CFR 50 design, such as IX>nald. C. Cook NUclear Plant Unit No. 1, as well as those of cur-rent vintage, should be directed to assuring the adequacy of the emergency core cooling system hardware by verifying its reliability in terms of timely response to smaller IOCA's that will probably occur, assuring continuity of operation during ECCS operating modes, and where appropriate adding diverse and redundant features that will enhance the ECCS reliability under these less extreme but more likely events.

Fundamental understanding of reactor core performance during a IOCA is a matter warranting continuing sttrly through confinnatory research. 'Ihe thermal-hydraulics evaluation of ECCS performance, while founded on thEo basic principles of fluid mechanics, involves, as with other fluid mechan-ics canplexities, mainly empirical correlation of heat transfer and fluid 294

B:>norable William March 11, 1976 flow parameters that can be verified experimentally only by ncut-and-tryn test programs. '!he experimental and analytical work required to assure an adequate, though not necessarily perfect, understanding of emergency core cooling performance under the full spectrum of accident conditions is there-fore an important requirement to be separately satisfied for the purpose of minimizing the risk to public safety arising fran erroneous interpretation of the performance analysis. '!his should be pursued vigorously by the nuclear power industry and in so doing it should address principal attention to the behavior of the emergency core cooling system under less extreme, but more likely, LOCA conditions. 'Ibis last mentioned evaluation work is of more inmediate interest than that associated with the bounding conditions used for design purposes.

295

B>norable William

• March 11, 1976

References:

1. Donald C. Cook Nuclear Plant Ulit No. 1 Startup Test Report, dated O::tober 15, 1975

2. SUpplement N:>. 5 to the safety Evaluation Report on Donald C. Cook Nuclear Plant Ulit J:.b. 1, dated January 9, 1976

3.

0 Iong Term Evaluation of '.Ibe Ice condenser System - Results of the December 1974 Initial Ice Weighing Program,n by J. G. Feinstein, dated March 1975 {Revised December 7, 1975)

4. "Iong Term Evaluation of '.Ibe Ice condenser System - Results of the March 1975 Ice Weighing Program, 11 by J. G. Feinstein, dated June 1975

s. nrong Term Evaluation of '.Ibe Ice condenser System - Ice I.Dss Calculations Resulting fran the July 1975 and O::tober 1975 Ice Weighing Program, 11 by J. G. Feinstein, dated J:.bvernber 1975 (Revised December 9, 1975)

6. nAxial :EOwer Distribution l-bnitoring System Experience and Peaking Factor Determination at the D. C. Cook Nuclear Plant Ulit 'M:>. 1, 11 dated O::tober 15, 1975

7.

0 NRC Staff Evaluation of westinghouse ECCS Evaluation M:>del Changes n:>cumented in WCAP-8622 {TAR-3065), dated January 9, 1976

8. nWestinghouse Evaluation ~el, O::tober 1975 version," WCAP-8622, dated N:>vember 1975

9. Donald C. Cook Nuclear Plant Ulit N:>. 1, Technical Specifications, dated January 15, 1976 {Proposed Revision)

10. 'lWX, BPI to the ACRS, concerning potential fuel failure due to pellet-clad interaction, dated February 4, 1976 296