ML18078A687

Forwards Fuel Grid Impact Loads for Salem Unit No 2 (Proprietary),Synopsis of WCAP-9283(nonpropietary).W/encl Applications for Withholding AW-79-04 & AW-77-27

ML18078A687
Person / Time
Site:	Salem
Issue date:	01/12/1979
From:	Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Parr O Office of Nuclear Reactor Regulation
Shared Package
ML18078A688	List: ML18078A687
References
NUDOCS 7901230174
Download: ML18078A687 (14)

Text

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I I - Westlngi10us~ Water Reactor PWR Systems Division Electric Corporation Divisions Box 355 I Pittsburgh Pennsylvania 15230 NS-TMA-2023

.January 12, 1979 Mr. Olan Parr, Chief Ref a: NS-TMA-1996 Light Water Reactcir Branch No. 3 dated 12/1 /78 Division of Project Manage~ent Office of _Nuclear Reactor Regulation Ref b: NS-CE-1737 U.S. Nuclear Regulatory Commission dated 3/21/78 7920 Norfolk Avenue Bethesda, Maryland 20014 r rriDV tj 1, '. *:*. _/ ~ ~ J

Dear Mr. Parr:

Please find enclosed:

.One (1) copy of "Fuel Grid Impact Loads fo.r Salem Unit No. 2" (Proprietary) as Attachment A.

One (1) copy of "Fuel Grid Impact Loads for Salem Unit No. 2" (Non-Proprietary) as Attachment B.

~~

~3. One (1) copy of a synopsis of WCAP 9283 (Non-Proprietary) as Attachment C.

Also enclosed are:

1. One (1) copy of Application for Withholding, AW-79-04.

-~2. One (1) copy of Application for Withholding, AW-77-27.

The purpose of this letter is to update and expand upon the information previously provided in the report, "Dynamic Analysis of the Reactor Coolant System for Loss-of-Coolant-Accidents: Salem Nuclear Generating Station I and II which was trans-11 mitted by Reference (a). The information herein was discussed with representatives of Westinghouse and the NRC Core Performance Branch in a tel econ on December 21, 1978.

Attachment A (proprietary) and Attachment B (non-proprietary) contain fuel grid load impact data which supplements the information previously provided in Reference (a). It should be recognized that grid impact loads are calculated using conservative analytical techniques and modeling assumptions. These con-servatisms, which have been discussed in other rlockets such as the North Anna FSAR (Appendix 5A, response to Question 5.71), include: (1) break opening time ~ ... "3Pr}Y-for postulated pip*e ruptures is typ*ically much longer (approximately 20 milli-seconds) than the assumed 1 millisecond break opening time; (2) internal hy-

. draul ic loads inside the reactor *vessel' which have a significant effect upon o"/

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Page 2 NS-TMA-2023 1/12/79 core grid impact loads are calculated using conservative hydro-elastic interaction methods; (3) break opening areas*assumed in the analysis are much larger than the actual areas which the structure's rigidity.will permit to develop; (4) structural models are based upon conservative modeling assumption; (5) the allowable grid impact value is the lower bound 95 x 95 value; etc. Considering these and other conservatisms, the calculated grid impacL loads are higher than can be expected to occur. It is, therefore, unnecessary to apply arbitra*ry factors of safety to the loads resulting from the core evaluations.

The information provided in Table*A-1 represents the grid impact loads for both seismic and blowdown forces calculated by the conservative methods described above. As demonstrated by analyses in \JICAP-9283, "Integrity of the Primary Piping Systems of Westinghouse Nuclear Pow.er Plants during Postulated Seismic Events," transmitted to the NRC by Reference (b), a seismic event will not cause a primary coolant system pipe rupture~ Therefore, the impact loads presented in Table A-1 do not require combination to assure adequate conservatism in the plant design. For your information, a synopsis of WCAP-9283 is provided in Attachment C.

This submittal contains proprietary information of*v1esti11ghouse Electric Corporation. In conformance with the requirements of 10CFR2.790, as amended, of the Commission's regulations, we are enclosing with this submittal, an application for withholding from public disclosure and an affidavit. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Committee. *

• Correspondence with respect to the appli*cation for withholding should reference A\JJ-79-04 and should be addressed to R. A. \iJeisemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh Pennsylvania, 15230.

~ T. M. Anderson, Manager Nuclear Safety Department J. J. Mcinerney/W. T. Bogard

/keg Attachments cc: R. J. Mattson, Chief NRC Division of Systems Safety

. *~ .

. *westinghouse E_lectric _Corporation Power Systems PWR Systems Division Box 355 Pittsburgh Penns1*ivania 15230 AW-79-04 Mr. Olan Parr January 12, 1979 Light Water Reactot~ Branch No. 3 Division of Project Management Bffice 6f Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20014 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

SUBJECT:

11 Fuel Grid Impact Loads. for Salem Unit No. 2 11 REF: Westinghouse Letter No. NS-TMA-2023, Anderson to Parr, dated January 12, 1979

Dear Mr. Parr:

The proprietary material transmitted by the referenced letter is of the same technical type as the proprietary material previously submitted con-cerning the analysis of the reactor coolant system for postulated loss-of-coolant accident for the Indian Point 3 Nuclear Power Plant. Further, the affidavit submitted to justjfy the material previously submitted, AW-77-27, is equally applicable to this material.

Accordingly, withhold~ng the subject information from public disclosure is requested in accordance with the previously submitted non-proprietary affi-davit and application for withholding, AW-77-27, dated June 15, 1977, which was approved by the Commission on June 14, 1978, a copy of which is attached.

The previous submittal was further supported by a proprietary affidavit, not attached, which was also sent to the Commission on June 15, 1977.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference AW-79-04, and should be addressed to the undersigned.

Very truly yours,

/bek Robert A. Wiesemann, Manager Attachment Regulato~y & Legislative Affairs cc: J. A. Cooke, Esq. (NRC)

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.. Box 355 f'itlstiurgt1 PU-!11Sjl'12'::~ 15230 June 15, 1977 AH-77-27

• Mr. \Victor Stello, Director Divi~ion of Operating P.eactors Offi~e of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cc;;;mission 7920 Norfolk Avenue Bethesda, Maryland 20014 APPLICATION FOR W!THHO~DING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

SUBJECT:

WCAP-9117, 11 Analysis of Reactor Coolant System for Postulated Loss .. of~Coolant Accident: Indian Point 3 Nuclear Po\':er Plant 11 (Proprietary)

REF: Westinghouse Letter No. NS-CE-1460 Eicheldir.ger to Stello

. Dated June 15~ 1977

Dear Mr. Stello:

This applicatfon for withholding is submitted by Hestfoghouse Electric Corporation pursuant to t~e provisions of paragraph (b)(l) of 10 CFR

• Section 2.790 of the Corrmission's regulations and is accompanied by a proprietary and a non-proprietary affidavit.

The proprietary affidavit sets forth the basis on v1hich the information may be vlithl:eld from public disclosure by the CoITTi1ission and address2s with specificity the considerations listed in paragraph (b)(4) of*

Section 2.790 of the Corr.mission's regulations. Because this affidavit contains Westinghouse p1~oprietary information, it is being submitted in confidence and is marked \*!estinghouse Proprietary Class 2. .Accordingly~

pursuant to the provisions of Section 2. 790(b)(i )(ii), \*1e request that the proprietary affidavit be withheld from public disclosure .

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The'unc!crsigncd ' lws revie'i!ed the inforn1ation souqht to be \*1ithhcld nnd i*s autLorizod t0 arr:ily for its withf**oldin'.] on behalf of ~~1~sti11i:::liouse,

~mo, no'::H'ication* of v1hic!11*rns sent to the Sccrr.-tC\ry of the Coii1:;*,issio1i on April 19,.1976.

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I I I I \.* I i .I U I. l\J I I hp~ng L' = I transmitted by ou1* 1etter (referenced above) be vii th he 1d fror.1 puu 1i c disclosure in accordance with the provisions of 10 CFR Section 2.790 of the Commission's regulations.

Correspondence with respect to the proprietary aspects of this applica-tion for withholding or the accompanying affidavits should reference AW-77-27 and should be addressed to the undersigned.

~~

Robert A. ~iesemann, Manaaer Licensing Prograns -

Enclosure cc: J. A. Cooke, Esq.

Office of the Executive legal Director, NRC a

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, . ... e _" /'M-77-27 AFFIDAVIT COMl'.Or::*lEAL TH OF PENilSYL VAr~ IA:

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COUNTY OF ftLLEGHENY:

Before me," the undersigned authority; personally appeared Robei*t P** Wiesemann, \*1ho, be*1ng by , me du1y s*,.;-orn u.ccording to 1m*r, de-

_poses and says that he is authorized to execute this Affidavit on behu17 of Westinghouse Electric Corporation ("Westinghouse") ~nd that tht aver-merits of fact set forth in this Affidavit are true and correct to the best of his knowledge, information; and belief:

• .Robert f... Hi ese~ann, Manager

.. . Licensing Prog ra:-ns -.

Sworn to and subscribed .:  !'.,

before *me this /:~ day of Y,t.t'_ 1977. *

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• (1)* I am. Manager, Licens.ing Progri:r.:S, in the Pressurized ~'utcr f!euctor

. **. . .systems Oivi.s ion. of Hes tinshcuse Electric Corporation and as such,

. I have been specifically delegated the functi6n of reviewing the

• p_roprietary informution sought _to ~c \*rithheld from.pt:bl.ic dis-

... closu~c in . coJncction with nuclear power plant 11censing . or rule-making proceeqings, and am autherized to apply foF its withholding ori behalf ,of the ~estinghouse Water ~eactor Divisions.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2. 790 of the Commissi*on 's reg!llatio'ns and in con-

. junction with the .Westinghouse a~plication for withholding accom-

. I pan.Ying this Af ficlavi t.

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(3) I have personal knowledge of the criteria and procedure utilized by Westinghouse Nuclear Energy Systems in designating information as a trade secret, .priv.ileged or as con.fidential comnercial or

• financial information.

• • . ~ * ... ~ . *(4) Pursuant to the provisions of paragraph (B)(4) of Section 2.790*;*:;

• • - . . ... ' of the Commi'ssion's regulations, the follci*!ing iS furnished for ... ' '

consideration by the :~orrunission in determining \*thether the in-

.formation sought to be withheld from *public disclosure should be

.. .: . . - withheld .

. (i) The information sought to be wi_thhe1d' from public dis,closure

. ... is owned and has been he 1 d ir. confide nee by Westinghouse .

(ii) *The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the p_ubl ic ..

• wcst"inghouse has a rational basis for detennining the types of information. customarily held .in confidence by it and, in that connection, utilizes a system to cetermi ne '.':hen and

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*t~hcthc~" to .hold certain types of .infonr.ation in confidence.

The application of that system and the substunce Qf thu.t

.. *system constitutE:s Westinghouse policy and provides the

• .:rational basis required.

• Under that" systtm, information is held in c.onfid::?nce if it

• • falls in one or more of several types, the release of \'1hich

...* : : mig~t resu1 t in the. lcs$ of an .existing or potential com-

• petitive a~vantage, as follows~

(a) The information reveals the distinguishing aspects of a process (or component, ~tructure, tool, method, ~tc.}

where prevention.of its use by any of Hestinghouse's

.competitors Nithout 1icense. from Westinghouse consti-

. *tutes a {:Cm~etitive eccr.c:nic advantage* over other

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• companies *

{b} It consists of supporting data, including test data, relative to a process (qr component, structure, tool,

• . * *. -1 method, ett. ), tt:e application of which data secures

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. ' I .a c~mpetitive economic advantage, e.g., by optimization

.. or i"mproved ma.rketabi 1i ty

• I (c) Its use by a* competitor would reduce his expenditure

. .* of resources or improve ;,is competit"ive position in

~he design, manufacture, shipment, installation, assur-

...ance of quality, or licensing a similar product.

(d) It reveals cost or price information, produc~icn cap-acities, b!.ldget levels, or. corr..1.ercial strategies of Westinghouse, its customers or suppliers *

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.*-*... Ce). It."revcals aspects of_pust, prc~cnt, or future West-

~ .i~ghouse or custou.cr funded dev~ lop:nent pl uns end pro-

_,.. grams of ptjtentia~ commer.cial value of Westinghouse *

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. (f)"* It contains patentable ideas, for \'thich p.atEnt pro-tection may be desirable *

. \.- (g)

• It is not the property of Westinghouse,* but must be

.*. - \.-\ trca ted as proprietary by Westinghouse accordi r.g to agreements. with the owner

• There are sound po?icy reasons be~ind the Westinghous~

syste_m \'1hich include the follovdng: *

(a) *The use of such information by Westinghouse gives

- Wes~inghousc . a comp*etitive advantage over. . its co;r:-

petitors. It is, therefore, v1ithheld from disclosure to .protect the Westinghouse* competitive positior. .

. ..... ii-.

It is information which is marketable in many ways *.

The ex~tent: to \*:hi ch ?Uch inforr.ation is available to competitors diminishes the Westinghouse abi1ity to

~e11. products and services involving the use of the .*-

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• I (c) Use by our competitor would put Westinghouse at a

compe.titive *disadvantage by reducing his expencliture*-

.*Of resources at our expense .

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( d) r.I c*ompone11t . of ri e la ry i . .rm i.on pc rt in en *t to a particular: cor1petitive advantage is potentially

. as' ~a1uable as the total c;ompetitive advantage. If

• . competitors *acquire components of propri ct~ry in for-

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fIK!-tion, ~ny one component m~Y be the key to the entire puzzle, thereby depriving \*!estinghouse of a competitive 1

aqvantagc.

• \ (e) Unrestricted disclosure ~ould jeorardize ~he ~osition I

of pro:nine~ce of Hestingho~se in the .\;!orld mnrI~et, I

and thereby give a r.a*rket tidvantuge to the cc:npet i ti on

.i in those countries *

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• . (f) The ~*:es_tinghouse capacity to 5nves~ corporrite assets i~ research and developmen~ depends upon the success fn obtaining and maintaining a competitive advantage.

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(iii) The information is being trans~itted to-the tommissirrn in

• • *confidence and, under the provisions of 10 .CFR Section 2. 790,

-:... *-it is to be received in confidence by the Comnission *..

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• (iv) The informatio~ is not available in pub1i.c sources to.lhe .* *

.. best of our knowledge and be.1 i ef *

. (v} The proprietary information sought to" be withheld ;*n this s~bmitta1 i.s that which is attached to Westinghouse Letter Number HS-CE-1460, Eicheldinger to Ste11o, dated June 15,

• 1977. The.letter and attach~ent are b~ing submitted in

• support of tne Corr.mission 1 s review of the reactor pressure

• vessel supports analysis for Indian Point 3

• Pub1 ic di sc1 osure of the information sought to be wi thhe 1d.

• is likely to cause substantial ~arm to the competitive

• position of Westinghouse, taking into account the value of

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.. the information to ~!cstinghouse, tlie u.~ourit of effort and

• ~oney*c~pcndcd by Westinghouse in developing the information,

• and . considering the \':ays .in \*:hich . the information could be acquired or duplicated by others.

Further the deponent s~yeth not.

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ATTACHMENT C Synopsis of VI-CAP 9283 - "Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events" WCAP 9283 discusses the simultaneous occurence of seismic and LOCA event and provides a rationale demonstrating that a seismic event will not cause a loss of *coolant accident. The rationale presented by Westinghouse is based upon a detenninistic evaluation of the design margin and quality* of the reactor coolant system while hypothesizing large piping surface flaws. In addition, this report describes a probabilistic structural reliability analysis v1hich supports the above position.

The approach used to develop a rationale for the reactor coolant system integrity incorporates a line of defense methodology at the component level.

The six tiers of protection employed in this approach encompass such tech-niques as non destructive examination during fabrication, inservice inspec-tion, and leak detection. The underlying basis for the integrity of the reactor coolant system is the good engineering practices employed throughout design and construction foll owed by equal diligence during the operating phase.

Deterministic Approach:

The deterministic evaluation of the reactor coolant system is initiated through an examination of the material behavior and stress analysis of components.

Tensile and toughness properties of stainless steel are examined along with fatigue crack growth behavior. In addition, stresses in the reactor coolant system are evaluated based on-;a typical ~/estinghouse four loop plant in a severe seismic zone. Based upon this evaluation, only the reactor vessel outlet nozzle safe end experienced stresses which significantly exceeded the

_yield stresses under combined loadings. This being the case, the elastic-

-plastic behavior of the reactor vessel outlet nozzle safe end is taken into

• account when evaluating fatigue, crack growth, and the potential for localized failure.

Upon completing the review of the material properties, a discussion of quality assurance practice is presented to demonstrate the measures used during con-struction to assure the integrity of reactor coolant piping. Current quality assurance techniques can detect flaws .75 inches in lenqth and .l inches in depth. However, for the purposes of this study, a flaw .25 the thickness in depth and 1.5 the thickness in length at the worst orientation and loading condition is assumed. It should also be recognized that pipe cracks can be identified by leak detection systems sensitive to identify leakage before* such leakage could interfere with the safe operation of the plant .

With the ~ssumption of the piping flaw described above, fatigue crack growth during a seismic event is studied. An elastic analysis is performed

• except for the.reactor vessel outlet nozzle safe end (highest stress location).

In this case a detailed study of elastic-plastic behavior is conducted and two methods (J-integral and equivalent energy fracture methodol-ogies) of plastic.crack.growth evaluation are used. Based upon these evaluations, the reactor vessel outlet nozzle safe end is the only area exhibiting even modest crack grqwth. As a result of these anlayses, it is concluded that a seismic event wil 1 not be the caus.e of a violation of the primary pressure boundary due to cyclic crack growth.

The next phase of t.his study determined the critical flaw size at the worst location (e.g., reactor vessel outlet nozzle safe end). Based upon this evaluation, the critical flaw size at this location is determined to be through wall flaw 25.5 inches in circumference.

Probabilistic Aporoach:

To support the deterministic evaluation described above, probabilistic structural" reliability analyses are performed. Three separate analyses a re described in the. WCAP as follows:

l. Probability of a crack initiating due to se.ismic induced cyclic fatigue in the ~*1orst location but with no flaw present.

2. Probability of a part through crack initiating (ductile tearing) and possibly penet~ating the pipe wall.

3. *Probability of~:: seismic event being t'he initiator of a loss of coolant accident assuming a through wall flaw is present.

A*summary of the conservative probabilistic results is provided in the attached Table. These probabilities represent the upper bound and although the input used in these analyses do not support precise prediction of failure probabilities, the calculated numbers can be interpreted as evidence of the adequacy of the design against the specific failure mode in question.

The items in the attached Table are arranged in order of occurrences and may be used directly to estimate the probability of failure for any desired con-ditional assumptions._ 1,he probability of a LOCA caused by an SSE during plant 11fe is 2 x 10 (i.e., the product of the probabilities in Items 1, 3, 4, and 5). Assuming the earthquake occurs and the quarter thickness flaw is present in the sense of the analyses performed, the probability of a LOCA caused by* the SSE is 4 x lo-6 (i.e., l x 10-4 x 0.04). *Considering the prob-ability of the flaw being present, the probability of the LOCA caused by the SSE is 4* x lQ-9,

TABLE Summary* of Worst Cas~*Probab,listic Structural Reliability Evaluation LIMITING LOCATION: Reactor Vessel Outlet Nozzle Safe End Item . Estimated Condi ti or. Dcscri ptf on Probabi 1i ty

• -1 Ho.*

SSE Probability of Occurrence During Plant life 5x1o- 3

~robability of.Crack Initiating During SSE 6x10- 9 2 No Flaw Present e.

. Probability of quarter thickness circumferential 1x10-3

• 3 C1rcumferent1a1 flaw Present Flaw being in worst location 4 Quarter. Thickness Probability *of Flaw leading to Leakage

• /

Circumferential Flaw Growing During SSE 5 Leakage, Through Wall Probab1l i ty of Through Wa 11 Flaw Leading to Circumferential flaw Six Rupture Times Thickness in Length (15 in.)