ML20148H213

Forwards Info Required to Complete Review of ECCS Reanalysis to Be Submitted Per 741227 Order for Mod of License

ML20148H213
Person / Time
Site:	Yankee Rowe
Issue date:	06/13/1975
From:	Purple R Office of Nuclear Reactor Regulation
To:	Andognini G YANKEE ATOMIC ELECTRIC CO.
References
NUDOCS 8011170402
Download: ML20148H213 (23)

Text

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-ISTRIBUTION:

Docket -

NRC PDR Local PDR ORB-1 Redding

, KRGoller OELD Docket No. 50-29 OI6E (3) l TJCarter Jurt 131975 RAPurple ABurger SSheppard VStello Ynnkee Atomic Electric Company TNovak ATT!1: Mr. G. Carl Andognini JRBuchanan Assistant to the Vice President TBAbernathy 20 Turnpike Road ACRS (14)

Hestboro, Massachusetts 01581 DEisenhut Gentlemen:

The enclosure to this letter identifies information that we require in order for us to complete our review of the ECCS reanalysis you m will submit pursuant to our Order for Modification of License dated C December 27, 1974. Some of the information has previously been b identified'(e.g., potential boron precipitation for PWRs, single h failure analysis), but has been repeated here for completeness. p Sincerely, S .

Robert A. Purplc, Chief Operating Reactors liranch #1 Division of Reactor Licensing

Enclosure:

Required Information (ECCS)

-1 cc w/ enc 1' Sec next page r- -

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cc: lir. Donald G. Allen, President Yanhec Atomic Elcetric Co=pany 20 Turnpike, Road .,.

Ucstboro, !!assachusetts 01581 Greenfield public Library ..

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I Attachment 1 REQUIRED INFORMATION

1. Break Spectrum and Partial Loop Operation ,_

The information provided for cach plant shall comply with the provisions of the attached memorandum entitled, " Minimum Requirements y for ECCS Break Spectrum Submittals."

2. Potential Boron Prec'pitation (PUR's Only)

The ECCS system in each plant should be evaluated by the applicant

- (or licensec) to show that significant changes in chemical concentrations will not occur during the long term after a loss-of-coolant accident (LOCA) and these potential changes have been specifically addressed by appropriate operating procedures. Accordingly, the applicant should .

L review the system capabilities and operating procedures to assure that boron precipitation would not compromise long-term core cooling capability [~

following a LOCA. This review shou]d consider all aspects of the specific plant design, including compenent qualification in the LOCA environment in .

addition to a detailed review of operating procedures. The applicant [

should examine the vulacrability of the specific plant design to single h failures that vould result in any significant boron, precipitation. p q

3. Single Failure Analysis A singic failure evaluation of the ECCS should be provided by the i.

applicant (or licensce) for his specific plant design, as required by l' Appendix K to 10 CTR 50,Section I.D.l. In performing this evaluation, the effects of a single failure or operator error that causes any manually ,

controlled, c)cetrically-operated valve to move to a position that could adversely affect the ECCS rust be consiGered. Therefore, if this consid-cration has not been speci ically reported in the past, the applicanto upcoming submittal must address this consideration. Include a Idst of all of the ECCS valves that are currently required by t':e plant Technica]

Specificatiens to have po ct disconnected, and any proposed plant nodifications and changes to the Technical Specifications that night be required in order to protect against any loss of safety function caused by this type cf failure. A ccpy of Dranch Technical Positien EICSD IS from the U.S. Nuclear Regulatory Conctission's Standard Review Plan is attached to provide you with guidance.

The singic f ailuro evolut tion should include the potential f or passive failurcs of fluid systems during long term cooling following a LOCA as well as singic failures of active components. For PUR plants, the singic failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.

4. Submerged Va]ves The applicant should review the specific equipment arrangement with-in his plant to determine if any valve motors within containment will become submerged following a LOCA. The review should include all valve motors that may become subnerged, not only those in the safety injection system. Yalves in other syst ems may be needed t o limit boric acid con-centration in the reactor vensel during long term cooling or may be required for containment isolation.

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. ThE applicant (or licensec) is to provide the following infornation, for each plant:

(1) - Whether or not any valve notors will be submerged following a LOCA in the plant be'ing reviewed.

(2). If any valve r.otors will be flooded in their plant, the applicant (or licensce) is to:

(a) Identi.fy the valves that will be subocrged.

(b) Evaluat'c tho potential consequences of' flooding.of the valves

- for both the short term and long t'crm ECCS functions and '

containe.cnt isolation. .The long term should consider the potential prob 1cm of excessive concentrations of boric acid in PWR's. .

(c) Propose a interim solution while necessary modifications are  :

being designed and implemented. (currently operating plants .

only) . i (d) Propose design changes to solve the potential flooding prob 1cm.

5. Containment Pressure (P..'R's only) .

The containnent pressure used to evaluate the performance capability of N the ECCS shall'bc calculated in accordance uith the provisions of 4

Branch Technical Position CS3 6-1, which is enclosed. ..

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Low ECCS Reflood Rate (Westinghouse SSSS Only) L  :

j o, - - - p Plants,that have a Uestinchouse nuclear steam supp3y shall perforn +

their ECC3 analyses utilizing'the proper version of the evaluation model,  ;

as defined belot ,.

(1) The December 25, 1974 version of the Ucstinghousa evaluatica model, i.e., the version without the modificaticas described in PCAF-8471 is ac. '*abic for previously analy:cd plants for hich ,

the peah clad temperature turn reund was identifice prior to the reflood rate decreasing belew 3.1 inches per second or f er rhich '

the reflood rate was identified to remain above 2.0 inch ner second; conditiens for which the December 25, 197' and March 15, 1975 versions would be ceuivaient.

(2) The March 15,-1975 version of the 5'estinghouse evaluation n'odel is an acceptable c.edel to be used for all nreviously analy:cd '

planto for which the peak clad temperature turnarouad was identi-fied to occur after the reflood rate decreased below 3.1 . inches per accond, and for which steam cooline, conditions (reflood rate ,

1 css than 1 inch per second) exist prior to the tino of peak clad I temperature turnaround. The March 15, 1975 version will be used for all future plant analyses.

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' JUBMTTTALS KINUPJM REOLn .EMENTS FOR ECCS BREAR SPECTRU ,

a b 10 INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectrum submittals. These guidelines have been formulated for contemporary recctor designs only and cust be re-assessed when new reactor concepts are submitted.

ECCS cooling performance The current ECCS Acceptance Criteria requires that be calculated in accordance with an acceptabic cvaluation model andlocations sizes, for a number of postulated loss-of-coolant accidents of dif ferent and other properties sufficient to nrovide assurance that the entire spectrum In addition, the i

of postulated less-of-coolant accidents is covered. -

calculation is to be concucted with at least three values of avalues these discharge spanning l coefficient (Cp) cpplied to the postulated break area, .l the range from 0.6 to 1.0. '

f operating Sections IIA and IIIA define the acceptable break spectrum for most Sections IIB and IIIB define the plants which have received Safety Orders.CP and OL case work (exceptions noted g break spectrum requirements for cost

later). Sections IIC and IIIC provide an outline of the minitum requircaents '

Such a conplete break spectrum for an acceptabic cc plete break spectrum.

could be appropriately referenced by sore plants. Sections IIID and IIIE i

provide the exceptions to certain'pl' ant types noted above, h.

A plant,due to reload a portion of its core will have previcusly submitted all f7 or part of a break spectrun analysis (cither by reference or by specific calculations). fuel If it is the intention of the Licensee to replace expended of the sste design (no ecchanical design differences which fuel with ne'.. thermal and hydraulic perferr.:nce), and if the Licensee intends could affect to operate the relcaded core in compliance with previouslyIfapproved the reloadTechnical core Specifications, no additional calculations are required.either of Sections 11A or IIC, design has changed, the Licensee shall adopt or of Sections IIIA or IIIC of this document, as appropriate to the plant' type (B'.:R er TWR) .

The criterien for establishinn whethe: paragraph A or C shall be satisfied willthe beshape cetermined on the of the PCT basis versus of whether break sire curvethe hasLicensee not can denonstrate that When been todified as a consequence of changes to the reload core design. the break the reload is supplied by a rource other than the NSSS supplicr, spectrum analyses specified by Sections IIC or IIIC shall Additional be submitted as a sensitivity minicum (as appropriate to the plant type, F>WR or PRR).

studies may be required to assess the sensitivity of fuel changes in such areas as singic failurcs and reactor coolant pump performance.

II. PRESSURIZED UATER REACTORS Operatinn Reactor K.aanalvses (Plants for which Safety Orders were issued)

A.

to make it wholly in If calculational changes

• were raade to the LBMA*

• Calculational changes /Medcl changes--those revir. ions made to calculational techniques or fixed parancters used for the referenced completc spectrum.

• • LBM--Large Dreak Model; SUM--Small Break Model

- _ . . -. ~_ .- -- . - - . . - - .. . ._

L10CFR50, Appendix R. the fol't 'ing minimum number of break conformance.wi j slaeu should he reanalysed. Each sensitivity acudy.perf ormed during the z -

individus11y verified as development of'the 1:CCS-cvaluation model sha.11 be j remaindng applicabic, or shall-be repeated. A plantismay

~ reference a break the same configuration d if it

. spectrum analysis conducted on'another plant -

j and core des,ign.

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1. 'If the largest break size results in the highest ' PCT: .3.

l a.. Reanalyze the limiting break. J 3

.b.. Reanalyze two smaller. breaks.in the large break region.

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in the hichest PCT: q

2. If the larcest break size does not result

a. Reenalyze the' limiting break. .j j

b .- Reanalyze a break'1arger.and a break smaller than the limiting q break. If the limiting break is outside the range of Moody nj multiplicts of 0.6 to'1.0 (i.e., less than 0.6), then the limiting-

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9 break plus two larger breaks must be analyzed. C. . *l j

If calculational changes have been made,to the SBM to nahe it wholly in y conformance with 10CFn50, Appendix R, the analysis of the worst small break j7 ,_]

(SBM) as previously determined from paragraph C belcw should be repeated. pp i

. B. New CP and OL Case Work e y

A complete break spectrum should be provided in accordance with paragraph C

• 9 below, except for the.following:  :

1. If a new plant is of the sane general design as the plant used as a

' basis f or a referenced cc plete spectrum analysis, but operating j parameters have ch::,ned which would increase PCT or r.cLa3-water q reaction, or approved calculational changer resulting in acre than 20 F 1 change in PCT have been made to the FCCS =oJc1 uced for the ref.crenced l completc spectrun, the analyses of paragraph one A above shou:d be provided of which is the plus a ninimum of three small breaks (SBM).The shape of the brech spe q transition break.* including the 4 analysis should be justified as remaining applicable, sensitivity studies used as a basis for the ECCS cvaluation cm'e1.

2. If a now plant (configuration and core design) is applicabic to all generic studies because it is the same with respect to the generic plant design and para *.aeters used as a basis f or a ref erenced complete spectrum defined In paragraph C, and no calculational chaarcs rcsulting in more than 20 F change in PCT were made to the ECCS model used-for the refercuced complecc spectrun, then no new spectrum an,1yscs are required. The new plant may instead reference the applicchie analysis.

• Transition Break (TB)--that break si.~.e which ir, analyzed with both the LBM and SBM.

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C. MknimunRecuirenents'forn'CompleteBreakSpectrum .

Since-it is expected that applicants will. prefer t'o reference an applicabic completcl break spectrum previously conducted on another plant, this r

paragraph defines the _nini-un number of breaks required'for an acceptabic complete break spectrun analyris,' assuming the' cold leg pump discharge is- 1 estab'lished as the worst break location. The worr,t single failure and worst-case reactor coolant pump status (running or tripped) shall be y established utilizing appropriate sensitivity studies. These studies should show that : the worst singic failure has been justified as a function of break size. Each sensitivity study published during the development

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of the ECCS evaluati'on nod;1 shall be individually- justified as reaaining applicable,- or . shall be. repeated. Also, a proposal for partial loop break

. operation shall be supported by identifying and analyzing theInworst addition, size and location (i.e., idle loop versus operating loop).  ;

sufficient justification shall be provided to conclude that the shape of j the PCT versus 3reak Size curve would not be significantly altered by the partial loep configuration. Unless this information is provided, plant F U

l Technical Specifications shall not permit operation with one or core idic reactor coolant pumps.

It must be demonstrated that the contain=ent design used for the break  ;

spectrua analysis is appropriate for the specific plant analyzed. It jj should be noted that this analysis is to be performed uith an approved - l cvaluation model wholly in conformance with the current ECCS Acceptance ,

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Criteria.

i' LBM--Cold Leg-Reactor Coolant Punp Discharge q

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a. Three guillotine type. breaks spanning at 1 cast the range of l Moody cultipliers between 0.6 and 1.0. .

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One split type break equivalent in size to twice the pipe '!

b. -l crcss-sectional area. I

c. Two intermediate split type breaks. .

d. The larga-break /small-break transition split. -

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2. LBM--Cold Leg-Reactor Coolant Pu=p Suction  ;

Analy:c the largest break size from part 1 above. If the analyses in i

.part 1 above should indicate that the worst cold leg break is an l intermediate break size, then the largest break in the pump suction  !

should be analyzed with an explanation of why the same trend would f not apply.

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3. LBM--liot Leg Piping I Analyze the largest rupture in the hot leg piping.  ;

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SBM--Splits-4 ,

j Analyze five different small break'eizes.' One of these breaks must ,

include the transition split break. The CFT line break must bc .l This break may also be one of the five

l analyzed for B6W plants.

small breaks.  :]

q III. BOILING WATER REACTORS The generic model. developed by Cencral Electric for BRRs proposed that split and ' guillotine type breaks are equivalent .in determining blowdown phencmena.  ;

- The staff' concluded this was acceptable and that the break area may bc considered at the vessc1 nozzle with a zero loss coefficient using a two to phase critical flow model. Changes ini the break area arc equivalent

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! changes in thc Moody multiplier. ,

The minimun. number of . breaks required f'or a complete break' spectrum analysis, 'C assuming a suction side recirculation line breah is the design basis accident

'(D3A) and the worst single failure has been established utilizing appropriate ,

sensitivity studies, are shown in paragraph C below. Also, a proposal for partial loop cperation shall be supported by identifying and analy:ing the worst In addition, break size and location (i.e., idle loop versus operating loop). .

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sufficient justification shall be provided to conclude that the shape of the r PCT versus Break Size curve would not be significantly altered by the partici p loop configuration. Unicss this information is provided, plant Technical Specifications shall not permit operation with one or more idic reactor coolant pumps.

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I' A, BWR2, LWR 3 and BC4 Reanalvsis (Plants f or which Saf ety Orders were issuod)

If the referenced lead plant analysis is in accordance with Section III, '

paragraph C below, the following minimur number of break sinos should be reanaly cd. It is to be noted that the lead plant analysis is to bc ,

performed with an approved evaluation model wholly in cenformance with the current ECCS Acceptance Critcria. A plant may reference a break spectrum analysis conducted rn another plant if it is the s2mc confiauratic:.

and core design. .

Each sensitivity study published during the development of the ECCS '

cvaluation model shall be individually justified as remaining applicable, .

or shall be repeated. 4

1. If the Jarrest break results in the highest PCT:

a. Reanalyac the limiting break with the appropriate referenced single failure.

b. Reanalyze the worst small break with the appropriate referenced single failure,

c. Reanalyze the transition break with the single failure and model that predicts the highest PCT.  ;

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,2.. If the._larne break does not resuli in ti e ' inhest PCT:

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a. Rcanalyze the limiting break, the larc, cst break, and a smaller break. g .

If ~ calculational changes have been made to.the SUM to make it wholly in [

conformanec' with 10CFR50, Appendix K, reanalyze the small break (SBM) in accordance with Section IIIC.

B. New Cp and OL Case Work A complete break spectrum should be provided in accordance with Section III, paragraph C below, except for the following:

If a new plant is of the same general design as the plant used as a 1.

basis for the Icad plant analysis, but operating parameters have changed which would increase PCT or cetal-water reaction, or approved calculational changes have been,made to the ECCS model resulting in more than 20 F change in PCT, the analyses of Section III, paragraph A  !

above should be provided plus a minimum of three small breaks (SBM),

The shape of the break spectrum one of which is the transition break.

in the lead plant analysis should be justified as remaining applicabic, including the sensitivity studies used as a basis for the ECCS ,

evaluation codel. .

If a new plant (configuration or core design) is applicabic to all

2. to the generic p generic studies because it is the same with respect  ;

plant design and parameters' used as a basis for a referenced completc spectrum defined in paragraph C, and no calculational changes resulting  ;

in more than 200F change in PCT werc made to the ECCS model used for the referenced complete spectrum, then no new spectrum analyses are required.

The new plant may-instead reference the applicabic analysis.

C. Minimum Renuirerents fer a Complete Dreak Spectru- 'l' This paragraph defines the minimum number of breaks required f or an

' acceptable complete spectrum analysis. This complete spectrua analysis i is required for each of the 1 cad plantr of a given class (LWR 2, LWR. , LWE4, BWRS, and BWR6). Each sensitivity study published durinr, the develep:wnt of the ECCS cvaluation model shall be individucily justified as remaining ,

l applicabic, or shall be repeated.

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1. Four' recirculation line breaks at the worst location (pump suction or j discharge), ocing the LBM, covering the range from the transition break (TD) .o the DBA, including Cp coeffic5ents of from 0.6 to 1.0 times the DBA.

2. Five recirculation line breaks, us'ing the SBM, covering the range from the smallest line break to the TB.

3. The following break locations assuming the worst singic failure:

a. largest steamline break

b. largest feedwater line break WW

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') c, largest usrc spray.line bryak

d. . largest'.recirculationipump discharge or suction break'(opposite

.sid.e of worst location)

D. BWR4 with " Modified" ECCS Same as Section IIIC.

E. BURS ,. .

Same as Sectio'n IIIC. ,

F. BRR6 .

Same as Section IIIC. ,

t IV. LOCA PAUJ272RS OF INTEREST A. On'cach plant and for cach break analyzed, the following paraccters (versus time unicss otherwise noted) should be provided on engineering ,

graph pcper of a quality to f acilitare ceiroletions. {' 5 -

--Peak clad temperature (ruptured and unruptured node) ,_

--Reactor vosoci pressure l5 W 3 I

--Vessel and downcomer water level (PUR only)

--Water Icvc1 inside the shroud (BWR only)

--Therr.a1 power

--Containment presourc (PUR only) a D. For the worst brcak analy:cd, the f cl3owing additional parancters (versus tire unicss otnerwise noted) should be providedThe an worst engineering singic graph paper of a quality to facilitate calculations.

failure and worst-case reactor coolant pump status will have been established utilizing appropriate sensitivity studies. l

--Flooding rate (PRR only)

--Core flow (inlet and outlet)

--Core inlet enthalpy (DUR only)

--lica t transfer coefficients ,

--}!APLllCR versus C):posure (DWR only)

--Reactor coolant tcmperature (PWR enly) ,

--)! ass releasel to containment (PUR only) ,

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--Energy released to containment (pWR only) .. - .

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--PCT versus Exposurc (BWR only) . .I

--Containment condensing heat transfer coefficient (PWR only)

--Ilot spot flow (PWR only)

--Quality (hottest assembly) (PWR only)

--Ilot pin internal pressure

--Ilot spot pellet average temperature

--Fluid temperature (hottest assembly) (PWR only)

C. A tabulation of peak clad temperature and metal-water reaction (local and core-wide) shall be provided across the break spectrun. .

D. Safety Analysis Reports (SARs) filed with the NRC shall identify on cach plot the run date, version number, and version date of the computer model utilized for the LCCA analysis. Should differences exist in current code listings made version number or version date from the cost F availabic to the NRC staff, then each nodification shall be identified ,

with an assessment of impact upon PCT an'd actal-water reaction (local ,,_;

and core-wide). - .

E. A tabulation of times at wh'ich' significent events occur shall be The following l provided on cach plant and for cach break analyzed. 'I cvents shall be included as a mininun:

--End-of-bypass (P:3 enly)

--Beginning of core recovery (PWR only)

--Tine of rupture

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--Jct pt=ps uncovered (BWR only)

--MCPR (BWR only)

--Time of rated spray (BWR only)

--Can quench (BWR only)

-End-of-blowdown ,

--Planc of interest uncovery (BWR only) i j

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Westinghouse Operatina Reactors (Safety Order Plants)_* l 3-loon I 4-loon ._

2-1 con i Surry 1/2 Yankee Rowe

.Ginna Turkey Pt. 3/4 IP2 1 Aewaunee H. Bi Robinson 2 D. C. Cook 1 Pt. Beach 1/2 Zion 1/2 i i Prairie Isicnd 1/2 1 ,

Operatir.c License ** i: i

! 3-loon 4-loon .

2-loon Beave Ytiley 1 - Trejan' .

Farley 1/2 - Saler. 1/2+

-m.. -9 l; orth A:nn 1/2 - Diabl0 Ce ver. .J IP-3 D. C. C00!. 2 l

McGuicc 1/2 Scquoych '/2

• 3 breaks required (11A). One plant may reference another l if applicable. .

• • Cc:::pict e spect rum required. One ;>iant nay ref erence another if applicable (see paragra;'h 116).

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. BRANCH TECHNICAL POSITION CICSB 18 APPLICAT!.03 0F THE SIMOLE FAllt .E CP.llERIC110 MANUALLY-COMIRCLLCD ELEClr.!CALLY OFIRAILD VALVE 5 A. BACf;Go.0"30-k'here a single failure in an electrical sysice can result in loss of capability to perf ort, lhis is necessary regard-a safety function, the ef fect on' plant' safety cust Le evaluated.

• less of whether the loss of safety function is caustd by a corponent f ailing tc reefer, a requisite rechanical rotion, or by a co ponent perforning'an undesirable e.echanicai rotie This position establistes the acceptability of discenrecting rewer to electrical c:. poner.ts of a fluid system as one reans of desi5ning 6;ainst a single failure that ei;5t cause an un-n desirable cc p:ner.t actien. These provisier.s are based on the assur, tion thet the c:-: r.cr.1 is then equivalent te a sinilar component that is n:t desicr.ed for electrical es:ra ti:n, e.g., a valve that can be onened or closed enly by cirect canual cperation of tnc valve. "

They are also b std on the assv stien that no sir.gle f ailure can both restore r: er to the '

electrical systra and cause rect nical rotic'. cf. te.c corponents served by the electrical systco. The validity of these assu .oti:1s sh:.ild be verified when trplyir.g this r:sitien.

( B. BRANCH liC"7'C.L 005'T:01 i ,

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1. Failures in botn the "f ali to functir.9" stnse ced the "urdesirable fur.:ti;n- serse c' 5 o.id conp:r.ents in elcctrical syste s cf sehes eric ciner fluid s,ste- co . re.: .ts  !

e t ".c r be consicered ie ncsi;nin: a;tirat e sir :le ~ ilure, even t'Ta;*. tr.e va've ce fluid syste cc ;;rer.: ay r.ot Le callc .p:n te fs.:tkr n in a ;i.cn sa'ct c.cretier.il seque':c.

2. L'here it is deter-* ired th:1 failure of 49 ele:trical sysic~ c: . ncnt c/- :s;st und:sirf.d rd: hen):rl :: ten of J v21.c er etter '!uid syste c ::re" t-d

.0 s totion rcsults in less cf ite syste safety fun: tic", it is e::cetable, i' 'ic; :(

0 the elcatric systc 5 design chan;es '. hat also r e) be a:tertab? ', to dist ".ne:t pic.r lec plar.1 te hnical s;c:{f':!!iens should of the valve er oth:r fluid systen : p nent.

include a list of cil cicttricall,.-::eratec v hes, and 'tre reevired cesitir~s of th.*se valves, to which the rer,uircrcnt fer rcroval of electric p .;cr is applied in creer to

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satisfy the single failure criterion.

3. Electrically-oterated valves that are classified as " active" valves, i.e., are recuired f

to open or Close in va-icus safety syste ercratienal seauences tut are r enu.,11 -

Su:h valscs r;< not be controlled, shesid 5' escrated frem the ain centrol reoc. I i lc incleded a~.,r,p these valves fren wnich ,c er is re oved in order to ncet t'.e s n:

f failure criterion unless: (t) electrical re cr can be restored to the valves f reni the '

riinute;

• • nain control re:m,(b) valve operatien 's not necessery for at least 0,)

following occurrence of the event requirinn such c;cratien, and (r) it is derenstrated

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that there is reasonabic assurance that all necessary operator acticas will te per- i ,.(

forred within the tire show.n to be adequate 'by the analysis. The plant tect.nical specifications should include a list of the required positions of nanually centrolled. l clectrically operated valves arid should identify those valves to which tFe rt;uire. -l rent for renovel of electric poece is applied in order to satisfy the sin;1e failure J criterion. 1 ii'I

4. L.' hen the single f ailure criterion is satisfied by removal of ele:trical :e,.er f rom valves described in(2) and (3), above, tnese valves should have redundant puit f en 1 indication in the r.ain control room and the position indication syste.'sPould. itself.

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• rect the singic failure criterion. *

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l S. 1he phrese " electrically-operated velves* includes both valves crerated directly by en  !

5 electrical device (e.g., a rotor-operated valve cr a solenoid operated selve) end those ,

valves opereted ir.directly by an electrical device (e.g., an air creretet valve .. nose  ;

air supply is contro11cd by an electrical solenoid vel'.c). l I

C. P.t F E P.!*.C E S il

1. I'.crorandur.i to R. C. DeYoung and V. A. Moore fr.om V. Stelle, 0:teber 1, 1?73.

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• BRANCH TECHNICAL P051710N C58'6-1  :

Hlfm*tN CL.';TAl'." INT r f 5' RC# MO.,EL

. FOR PR

. [CCS PI*f 0, T%';Cf [ VALUATION A. B AC rC?O'";3

. Paragraph 1.D.2 of Appendix t to 10 CFR Part 50 (Ref.1) requires that the contair ;cnt I

pressure used to evaluate the perforran:e casacility of a pressurized water reactor (P.;R) ercrgercy corc cooling system (ECCS) rot excccd a pressure calculated conservate.cly for

- that purpose. It further recuires that the criculation in:lude the effects of epcraticn of all installed pressure reducing systems and processes. Therefore, the folle.-ting trar.ch technicci position has been develc cd to provide guidance in the perferrance of mir.1 ur. ,

]i k ,

i con laircent pressure analysis. The approach d: Scribed belc.: ar. plies only to i.c [CCS- t i related contair.r.cnt pressure evaluation cad not to t*.e c:ntainrent functienal capadility evaluation for postulated design tesis accidents.

B. P.MN:H TECP';;CD POSIT!r.jj

1. .I_n.pu t Ir.fc r-s tien f: "u'el

.I -

a. Initial Contain ar.1 b ternal Cteditions The rtinir.un ccniair. cr.; gas tcr:.oretare, r.infrur. conta tr.~ent pres sure , p L

and r;axirut. humidity that t.ay Le encountered under li-itir.; r.orral c::ratir.) h conditions shnuld te used.

b. Inititi Outtic'e Ccntair"+nt L :' ret :cditic.rs A reas:nably Ic., a- bient t: ?cret. re c. sternal to the co .tcir.~cnt s';;1d be us cJ.

c. Cn9tei ent Ve.l_u t.

T he raxir:r. r.c t f rr c :: tit iv ent ce tu-c st;.,1d .-e cs g. .i! raxi .- feci volure sh0uld te : .i: fr - 1 . ;r:si co.tri~ cet 501. e :r,4 t c .:L es of internal str;:t.rcs s.:S as ..alis :-d fic: s, str;:t;ral sic:1, ,e:r ::_i.  :

.nt, and piping. The trdisid 31 vol re calcuistiens sr.:vi: <cn e:t . c u ccrt?ir.t. in the component volu cs. .

2. Active Heat Sirls .

a. Sf,ra,v e n d f a n C :l i r - P.ste-s The operation cf all en;inacred safet. feature contain-ent heat rc.? oval systers operating at raxirum heat re oval capacity; i.e.. uith all contair. rent s; ray trains operating at rcaxi.um m flow cor.ditiets and all crergency fan coolcr units operating, shesid te assar:d. In 3:ditien, the tr.ini ;, te .perature cf the stored water for the sprey ccoling systen ar.d the cooling ter supplied to the f an t.

coolers, based en tect.nical specification linits, sheuld.!c assu..cd.

6.2.1.5-3 I

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, conditien! ;arding a singic active failure, stor water terperature, and cooling waier temperature have been selected f rom inc standpoint cf the overall LLCS model. *  ?

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b. Contain ent Sica-dif rino Mit5 Spilled FCCS Mater The spfilage of sulccoled ICCS water into the containment provides an additional heat sink as the subccoled (CCS water mixes with the stcaN in the containrent.

The ef fect of the steam' water enixing should te considered in the contairrent pressure calculations.

c. Contain ent Steen l'iriro i:ith Ua ter f r: 9 lee Pelt The water resultir.g f r:a ice r elting in an ice condenser contair. ment ;revides an additional heat sink as the sub, cooled ,.ater r.ixes with the stca 1 s. nile decinir.g i

f rom the ice cor.dcr.scr ir.to the le.wer contain .cnt volur.c. tlc effcc cf the steam-water nixing sh:uld be censidercd in the contain ent pressure cciculatiens.

3. Ibssive Me.a t Sir.ks

e. Identification ,

The passive heat sir.Ls that snould 50 in:ludcd in the centair. ent evaluatien model should be estcblished by ider.:ifying those structures and cc-,e:.ents v.ithin p the contair?cnt that cosld influence the pressure response. Tne tir. s c.f struc- f-ll turcs and cen?onents that sh uld te included are listed in Table 1. (}

e Data en passive helt sinks have L:en cez. piled fro 1 previews r cic-5 erd have I. '

.t been used as a tesis fer :ne si clific: r.:d:.1 ostlir ed ! ele.:. Tr.is r:::1 is i accepitbic for .inir-.9 contrin.cnt prcssure analyses f: censtrs: tic territ l applicc tions , 2n: cntil 5,:5 ; ira (i.e., at the c;.cretir.g liccrse c.dc.3 - *a F cor.plete identif t:::i:n cf availabic ' e: sir.:s can te rede. inis 5 - lif'*d appreacn has aise teen (0110-:: fer c;cratir; elar.ts :9 licc H cs cc .lf i ; .119 Sectlen 50.*5 (1?l2) :f 10 CT; pcrt !?. For surn ecses, 2.d fc- :r u;- ::*e-perr.:i t revie..s ,

ere e :c'.2iled listin; of *ca: si.is ut hin ' e c: :3:~ cc:

of ten canret to ;revi:c:. t' e f '.lc..ir.; precccare a r: sse: ;> ::c! :-0 passi.c heat sinks witnin ta.e c: .: in cnt; (1) Use the surf a:e arca and thickr. css of the prir.ary ceitain en: s:rci shell er steel liner an: asseciated arcters.end ceno e:c, as appr:eriate. ,

(2) Estir. ate the estesed surface area of other steel heat sinis in acccrd:nce with figure I and assucc an average chictness of 3/S inch.

H (3) HodeltheinternalcencretestrucjuresasaslabalthathictnessofIfeet and expesed surft:c of 100.000 ft' ,

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lhe heat sint therre;Sysical preperties that would t'e acceptable are shown in 1 Table 2. 1 6.2.1.5-4

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passive * % sinks, with appropriate dienensions - properties.

b. Heat Transfer Coefficients The following conservative condensing heat transfer coefficients for heat transfer

• to the esposed passive heat sinks during the blowdosen and po,t blo.,de..n phases of the . loss of coolant accident should be used (See figure 2):

a (1) During the blo',do,en phase, assume a linear increase in the condcnsing heat j transfer cecfficient frci hg g g*C Clu/hr-f t *f, at t e 0, to a gesk

'value four times greater than th? raxiru , calculated condensir.g M.at trans-fer coef ficient at.the cr.d of bloed:c..n, usirig the Tagant corrcletien (Ref.2), 0.62 .

h = 72.5

.0 rax n 2

' ~

'T where bre .ax = riaxi'r' urn b. at transfer ccef ficient, Btu /hr-f t -

= prircary coolant (nergy, Stu Q

Y = net free contair-ent volure, f t 3 t = tirr.e interval to end of blowdeen, scc.

p (2) During the icng tern ;ost bicwdo..n p"ase of the accider.t characterind by low turtivien:e in the contair.r ent at osptere, assu e condensing r est trensf er "

coef ficients 1,2 tices greater than t'.ose predict (d by the !Ichida data (Ref. 3) and given in lable 3. ,

7 (3) During the trer.sitten , .ase of the accident, tet.:cen the e d of tic..>:..n and ,

the long-ter post-tic..de<.n enase, a reasenably cen'ervative cc W1 transitien in the cc .densin; heat trer.sfer cccf ficicnt s'.culd W ase.-cd ,

(Sce figure 2).

ite calculated centensic.; neat trar.tf er ce?f ficients t:sc cr. tre er . C. . :

- sh;.ald t e errlied tv all ee:std ;assive rest si t.5, !:tn -etel ind : :s c'.e , and for Loth.pcinted and nr.ainted ss faces. r Heat transfer betecen ad,icining r2terials in pssive rat sint.s s*a h . :35:d en the essa rtion of 90 resistan:e to heat fic. at the nertal intcrfner.. An example of this is the contair..ent liner to concrete' interface.

C. R[fERE!CCS.

1, 10 Cf R 150.45, "Accertar.ce Criteria f or Er.ergency Core Ccoling Systens fer Light '.:ater Nuclear reecr Recctors," and 10 CFR part 50, Arpendix K, "CCC5 Evaluation 'edels."

2. T. Teg.mi, "Interin Rerert on sarcty Assessr.ents and f acilities (sta91is> >:nt reaject )

in Japn fer Peried [nding June IMS (!'o 1),' rreparcJ for the !Dticnal Rea:t" lesting

$ tat ten, icdroary 28,1956 (unpublishcd wort).

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. Uc 6M .

, n! ster I'vaer Rear' s." Proc. Third Internatienal Conferer cn the Peaceful Uses of Atomic Energy. N ute 13. Session 3.9. United Nations. Gi..'.eea (1954).

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TABtf 1 10titT1rltt71C?l or C0' ital?PEft? HEtT Sit;r.5 i

1. Con *ainient Building (e.g., liner plate and external concrete walls, floor, and sump, and lineranchors),

2. Containrent Jr.ternal Structuras (e.g., intert.s1 separation walls and floors, refueling pool and fuel transfer pit walls, and shielding *:4115).

I

3. Supports (e.g., reactor vessel, steam generator, psmps, tants.,rajor components, pipe supports, and storage racks).

4, Uninsulated Syste..5 and Components (e.g., cold water systers, heating, ventilation, and air conditionin; sysicrs, pu eps, r.ctors, f an coolers, recorbiners, and tan;.5).

5.

Miscellaneous Equi;nent (e.g., ladders, gratir.gs, cicctricci cable trays, and ciancs).

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TABLE 2, i P.[ AT $1!D: TitiR:'.3PHYSICf L N'OPCRT!ts Specific Therra)

Denst}y Heat Conductivitv Htterial ib/ft Btu /lb *F Btu /r.r-fi-j Concrete 145 0.156 0.92 Steel 490 0.12 27.0 O e e

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98 e

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www~w

• rnMons UCHIM HttT TPA*:$r(R C0(Fr1CittiV5 Hass , Heat Vransfer Kass Heat fransfer Ratio Coefficicnt Ratio Coefficiq6t (1b air /lb stem) Mu/htftr)

(Ib air /1ti stNd !Ctu/hrf'..'fl i

50 2 3 29 20 8 2.3 37 9 1.8 40 18 14 . 10 1.3 6

'U 0.8 98 10 14 .

7 ,

17 0.5 140 .

2) 0.1 280 5

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6 8

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Fit,ure 1

, Area of Steel IIcat Sicks Inside Containment 5I .

n o

u 4 -

C t/3 AJ N

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> w .

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n s a

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' 1 2 3 4 F 3 Con t a in men t Free Volume, x 10 ' f t

' Revised 12/74 O ,

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Figure 2 Condensing IIcat Transfer Coefficients for Static Itcat Sinks - .

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O 0 h = 4. x h Tagami . -

3 u max '

linear N )J w.

e  ! .025(t-tp)

r:

o I h=h stag

+ (h max -h stag) e m

y -

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s  ; h - 1.2 x h Uchida c steg

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