ML18046B087

Forwards Final Evaluations of SEP Topics VI.2.D, Mass & Energy Release for Possible Pipe Break Inside Containment & VI-3, Containment Pressure & Heat Removal Capability. Design Mods to Accomodate Air MSIV Failure Necessary

ML18046B087
Person / Time
Site:	Palisades
Issue date:	11/16/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-81-11-030, LSO5-81-11-30, NUDOCS 8111240342
Download: ML18046B087 (85)

Text

Docekt No. 50-255 LS05-8l-11-030 Mr. David P. Hoffman November 16, 1981 Nuclear Licensing Administrator Consumers Power Company 1945 W. Parnall Road Jackson, Michigan 49201

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) FOR PALISADES NUCLEAR POWER PLANT, UNIT 1 - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 (DOCKET NO. 50-255)

REFERENCES:

(l) D. Hoffman from D. Crutchfield, Letter, Same Subject, dated June 17, 1981

{2)

D. Crutchfield From R. Vincent, Letter,

Subject:

Comments on June 17, 1981 Letter, Dated July 9, 1981 Enclosed is a copy of our final evaluation of SEP Topics VI-2. D, 11Mass and Energy Release for Possible Pipe Break Inside Containment 11

, and VI-3, 11 Containment Pressure and Heat Removal Capability 11 This evaluation compares your facility" as described in Doclt":et No. 50-255, with the criteria currently used by the regulatory staff for licensing new facilities.

Appendix A to our evaluation is a draft Technical Evaluation Report from our contractors Lawrence Livermore National Laboratory. Appendix B, con-tains the resolution of the comments provided (see Reference 2) while we agree with the validity of comments which point out conservatism in our analysis; it is our opinion that their overall effect would not signifi-cantly modify our conclusion that design modifications to accommodate air MSIV failure are necessary.

The staff does not plan to reanalyze the accidents.

This evaluation will be a basic input to the integrated safety assessment 4f.~C>c.f for your facility. This topic assessment may be changed in the future if..;J..s

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'DS'-' "~'- (o'I Ac\\)~

..,.. ""Q'*'b(A.C,h OFFICE~ ************************

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• ** e1 ****eo *ee ******** **

• • • • • • • • • • • • • • • • • • • • • • • • ****** ** *****1111e ***** elle SURNAME.................................................................................................................................................... ************************)

DATE} ************************ -...... *-***************............................. *********.................................. ************************ ************............

NRC FORM 318 (10*80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

J*

• your. facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.

Enclosure:

Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing Draft SEP Topics VI-2.D and VI-3 cc w/enclosure:

See next page SER OFFICE~...............

SBrow.

SURNAME~ **********:;;***********

DATE *

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SEP :S*

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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

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~* o Mr. David P. Hoffman cc M. I. Miller, Esquire Isham, Lincoln & Beale Suite 4200 One First National Plaza Chicago, Illinois 60670 Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Myron M. Cherry, Esquire Suite 4501 One IBM Plaza Chicago, Illinois 60611 Ms. Mary P.*s1nclair

. Great Lakes Energy Alliance 5711 Summerset Drive Midland, Michigan 48640 Kalamazoo Public Library 315 South Rose Street Kalamazoo, Michigan 49006 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan Office of the Governor (2)

Room 1 - Capitol Building Lansing, Michigan 48913 William J. Scanlon, Esquire 2034 Pauline Boulevard Ann Arbor, Michigan 48103 Palisades Plant ATTN:

Mr. Robert Montross Plant Manager Covert, Michigan 49043.

49043 PALISADES Docket No. 50-255 -

u. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:

EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604 Charles Bechhoefer, Esqv, Chairman Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission.

Washington, Dg C.

20555 Dr. George C. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 Dr. M. Stanley Livingston 1005 Calle Largo Santa Fe, New Mexico 87501 Resident Inspector c/o U. S. NRC Palisades Plant Route 2, P. o. Box 155 Covertp Michigan 49043

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~-*- *---

-SAFETY EVALUATION:REPORT

-~-"-:-:

ON CONTAINMENT PRESSURE AND

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HEAT REMOVAL CAPABILITY 7

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SEP TOPIC VI-3

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AND MASS AND ENERGY RELEASE

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FOR POSSIBLE PIPE BREAK SEP TOPIC VI-2.D FOR THE.

PALISADES ~lUCLEAR POWER PLANT DOCKET NO. 50-255

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.1 TABLE 9F CONTENTS.

I.

II.

III.

Introduction

'.~.-

. *-***~-------------***--

... Review Criteria.

IV.

Revi~w_Gujde}ines

• V.

- *Evaluation --'.* * -

VI. * '"*conclus1ons~~--: -

... *.. -.: :; ~

VII.

- *References 7"".=.--. :

Appendix A.

(Draft)*

Containment Analysis and Evaluation for the

.Palisades* Nuclear Power Plant, Unit 1

.. ~

v I

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j.

e.

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Introduction

-Th;i'alisades-Nuclear P.ower Plant, Unit 1 began corrmercia1-operations tn

_1~7L Since"then_ t~e_staff's.. safetY:.. :r~view. criteriac-have, changed.

As -

-~

part of ~he-systematic Evaluation: Pr,~ogram.(SEP), --*the contaiiiment*.. pressure

• and heat. remova1.capability *(Topic.YI-'3) and the mass and energy-release,_.

. for po~sible:pipe*break*inside~containment (T6pic:VI~2:D) -~av~'beift~~e~rJ

eva1 uated *.. ~ -...-. "-._*

...

-The purpose-.of this __ eva1uation;is to document the*devi:atfonscfrom*current

~s~f e_ty * ~ri_t~ri ~ ::as.. :oth.ey r~l ate ;to the containment_ pressure* and*:.heat ~ r:..;'"

mqval ~apabil:ity and.::the.. mass/anergy release for..opossil51e-p.ipe ~reak *fostde

-_containmen-;. *_ F-urthermo.re, *_i:.nd~pendent analyses :.in.accordarice* with curtent

. -.ortteri:a..:.were ~perfo.rmed :.:to -determi;rie the ~dequacy_:of :th*e *:conta'-i nme:nt "desi-gn.

. : ~bas_is ( e. g-.-.; :des.ign: pressu*re ~nd. temperature) and to *provi.de i:nput :fo_r Un-resolved Safety Issue (USI) A-24~.Qualification of Class *1E; Safety Related*

Equip~ent. The s~gnificance of the identified.deviations, and recommended_

corrective measures to improve safety~ will_ be the subject of a subsequent, integrated assessment of the Palisades plant.

II.

Review Crit~ria The review criteria used in the current evaluation of SEP Topics V!-2.D and ~I-3 for the Palisades plant ~re contained in the following documents:

1) 10 CFR Part 50, Appendix A, General Design Criteria for tfoclear Power*

2)

Plants:

a)

GOC 16 - Containmment design; b)

• GOC 38 - Containment heat removal; and c.). GDC 50 - Containment des~gn basis *.

10 CFR Section 50.46, "Acceptance Criteria for Emergency* Core Cooling System for Light Water Nu cl ear Power Reactors."

,,r--.

r::.. *.

~*.'

9.

2 - "

3)..~0 CFR Part 50, Appendix K, i'ECCS E~a1uat{on M~d~is 11 -.. *

4)

NUREG -75/os7*. -Standard. Review Plan._ for the* Revie~ of s~fety: An~lys_is- *

• _. - -.Rep~r~~*:*t~r N~cl ear: Power Pl a~t~ _( SRP--6 ~2.i, -Contai ~~e~~ '.; F~~~ti o*nai -~.: _, 2

~.. ": -:.-~...:.. -.-:-::

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Design).

-~.-*

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. The review: areas: i9entified be1ow.-ar~ n9t-addressed in_ th-is. ~p9rt,_ but. __

-~are*related:to the SEP~topics*:of mass arid* energy re1eas~- for:P9$Si'b1e-::-:-:

s pipe" br1=ak i nsj de:~on-1:a:inmeot;* and/9r":".cont~Jnrnen-t press~rt;:arid: he~~.,f'.'e- :: ;::- -

• -moval capability.. ~-:**--**.**-~:*_:-="""--_,,,._.... _..,

~-

"1. * -III-1, *Cl assi-fi cation* of. Structures, Components and Sys~ems-_-{Se~ smi c-

~nd Qua1ity)

. 2.

1-1!~78,"'oeiign c'odes~~b~si~~:crit~~i~-~\\oad Co~binations:"~niRea'c~-;

. tor::Cavjty*oes:igfl:Cr:it~ri-a::-

~---=--

-3.* _VI-7.8,- ::ESF. S~i-tchoyer frP.lli-,'Injection-to;RecircuJa:tion~Mpda {Auto-:..-

matic ECCS. Rea1 ignment)

4.

IX-3,* Station Service and Cooling Water Systems

5.

X, Auxiliary Feedwater System

6.

III-12, Environmenta1 Qualification of Safety Related Equipment*

IV.

Review Guide1-ines General Design Cri~erion (GDC} 16 of Appendix A to. 10 CFR Part 50 requires

. that a reactor containment an~ associated systems shall. be provided to es-tablish an essentially leak-tight barrier against the uncontro11ed release of radioacti.vity to the en_vironment and to assure that the containment de-sign conditions important to safety are not exceeded for as 1ong as the postulated accident conditions require.

GDC 38 requires a containment heat

-remova 1.sy~tem "be provided whose system~safety ~unction sha11 be to reduce the containment pressure and temperature following any loss-of-coolant acci-dent (LOCA) and maintain them at an acceptably low level; furthermore, the I'

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. C

v. system safety function shall be:.achievable assuming a single failure*.

Goe*

50 ;;qufres that the containment structure and the containment heat remova 1 system s_hal,~,be desi gne4 so. that th.~t!tr.ucture can accomnoda~e *. ~i.th. suffi~.

-..... - ~

cient margin,-. th~ calculated _pressur~_and temperature_.conditi9ns _resu1til19-

.~

'..,JI I.. t I~

f~o~ any _L~~A~.:.~hi.s. ~argin J~:mg~.ta~~~~;from ~he:_~~~:~:v~:1~~*-;c~l~~1-~t~,9n of mass/en~t:gy.release and the containm~o.t_model is_~iscussed fo_the Standard

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_Review Pl~n (SRP)~ection.-6.2.1, Containment_fuoctionaJ Oesjgn*~------ _. _

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Tne coiitainment design-basis includes:*:the* effects of=stored-and=generated~-. *:.:-..

~ri~~g/;k:it,~-accident.

• C-alculations of the energy.available for rele~se shoul d.. be:do~e.. in,_accort;lance. with~tbe_requirements. of_lO_ Cf& Par~ ~O_. :Sec-

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ti on 50.. 46-.~nd ~ppendi x K11 : paragraph I.A,. and the_ coriserya~ism_ ~s. specified

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• in_ SRP..:6.2.1.3. __ ~Ihe=m~ss.. and eg~.rgy"r~leas~ to th.e.cont.ainment_from_~LLQGA_.:

~

- ~ *-. -

. =.... -

=* : ' :.. -:: -.

-* should be-considered. in terms !)f blowd9wn, reflood, and post-ref1ood. _ The.

• mas*s and-energy rel ease for postul ate.d secondary system pipe ruptures sho1,.11 ~

be calculated in accordance with SRP 6.2.1.4. The review also includes the.

analysis of postulated single active failures of components in the se.condary system.

S_y reviewing* the licensee's ana1ysis, we identify deviations from the cur-rent criteria and we perform independent analyses, as required, to evaluate

~h* significance of these deviati9ns.

In our analyses, we use the best esti~

mate method; i.e *., by using actual pl ant design data, we obtain a best esti-mate, but still reasonably conservative, containment ana1ysis.

The evaluation is completed by comparing the results with the containmertt design _basis.

Eval uati.on Based on a review of the existing docket of Palisades, two areas -n*ere identi-fied as deviating from the current ~iteria. First, in the LOCA analysis as I *..

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• ' *4.

described in the Palisades Final Safety Analysis Report (FS.AR), the lice*n-

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see did not consider the ref1ood phase-*or energy released from the secondary system to the containmerit-through-the.steam generator~ If the refloo.d phas:e were. consi de-red, tlie Pali-sades ~pfant'woul d :-~eet =the:. p:tovi~ i eris: of SRP :6-:2 :1.3 and *10 CFR *p*art so, Appendix-*K. - :seca*ri-d, :-fo.th~ main s-tei~cli~e ;b~~ak-(MSLS(

. a:nalys*i s -as' presented-in= tfi{ i; censee' ~sUi:imi ttaf d~te(f Jatiuary -Zf,- i9sb::r,~~:;~

(:Reference *n, *tt,e:-sing1:e: fail-tire ~of 'a:*main stearii-iso1~tio~* valve (MS!V) was not considered.

We find that the MSIV failure can be shown to be t'he worst

.. : -.. _ -* -.. ~ -~ -~:: =-

~ing-l'e-fai1ure si_nce it WQ!l1_~ __ a11_o!".b~tt_h.§t_ea~ _gen~r.a_t_Qrs., __ to...blo~*.down.

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• cause the P~l ts_ade~ pJ ~_nt.M.S~V,s_ a.r~ oe_s_i_g_n_~~ as check_ xa.1 ve_s_, _tti,ey_. _cl--0se_.o,ri.ly.,,..

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in the for<< a rd flow direction.

In the e.vent of a MSLB upstream o{ the-MSIV-9

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• the fluid from the broken. loop st~am generator, and also from the. unbroken *

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loop *steam gener*ator cou1 d be released to.. the containment. *This blowdown of tWo steain'-ge*nerators-co'ui ((*prodlJ.ce-: peak-containment p'ressu-re; and. t'emp-~ra-tures that' are beyond desfgn: limits~" Therefor~, the 1icens~e's ~n~1ys*is -

which did not consider the MSIV single failure represents a deviation from the.provisions of SRP 6.2.1.4.

To assess the significance of these two deviations, our consultant, Law-rence Livermore National Laboratory ( LLNL). performed an independent ana1y-s is which is presented *in draft Appendix A of this repdrt.. Mass and energy

.. telease rates utilized in the analysis were calculated using both RELAP-4 MOD 6 and MOD 7 in accordance with current criteria. Calculation of the post-accident* containment pressure and temperature was done us\\ng CONTEMPT-LT/028.,

Several cases were analyzed for both the primary and secondary system analy-ses. For 'j:he p-rimary system pipe -breaks, the calculated transient reflects a post-accident peak containment pressure from 66 to 69.7 psia arid a temperature I

--:-* *.*~:

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_from 273°F to 325°F, dependi n9 on the particu.1 ar assumptions made.

The con-tainment design pressure and tempera:~re for Palisades are 55 psig (70. psia) and 283°F. There is, therefore, a r~~ge of oi to 5% margin between the cal-culated post-accident containment pressure for primary system breaks as

• identified in draft Appe~dix* A-a~((*the-design pressure. Fu.rtherrnore 9 because

-of the thennal capacity of the containment structure, the containment design temperature is not e*~ce-~d~d even:. tiiotigh:: the-containment atmosphere'"ternpera tu re exceeds 283°F.

The containment re_sponse to a s.e.~ondary system pipe break is also given in draft Appendix A. This. analysis. was for a MSLB which considered the blow-

• down of both steam g_enera.t.ors resu1ting from the postulated single failure of°

~ a MSIV.

The calculated transient reflects a. post-accident contajnment pres-*

.<.. _:.. /<

. sure of 97 to 107 psia and a temperature of 420°F to 465°F. The peak calcu-lated -containment pressure exceeds the design value by a substantial amount~

Analyses have a1so been performed assuming a design change which wou1d prevent the b1owdown of both steam generators.

The results show that the calculated*

peak pressure is 68.5 psi a, which is 1.2 psi below 'the design value.

The cal*

_culated peak containment temperature is 413°F. Because the thermal capacity of the. containment structure, the containment design temperature is not ex-ceeded.

VI.

Conclusions

• We have identified the deviations of the Palisades plant from current 1ice.ns-ing criteria in Section V, above.

From the independent containment analyses

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• e.

' A 6 -.

~

rep~d in draft Appendix A,, we conclu~e that the Palisades con:tainment de-sign pressure can meet our current cri.teria, if design changes were made so

~

that the sirig'fe failure of a MSIV can be accommodated.

The short duration cont_ai r:imen1:_'.a'b!losphere over-temperature conditions during* a MSLB (Fi gu*re 4.ST) and LOCA* (Figure 3.llT) should be considered as an input to the environmental qual i f-ieation of. Class lE safety related equipment, USI A-24.

v11*. Referehees ~- :. -:- :: -=~*

-~

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1. -. Letter-fr.om.R_og~r W. HJJston of Consumers Power Company to Dennis L. Zi e-

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~.... -=: ;. :::. :.::*

mann o.f NRR, NU:clear Regulatory Commission,

Subject:

Automatic* Initiation of-Auxiliary Feedwater System at Pali.sades Plant, Docket No. 50-255, Li-

.. _ce.nse. _DJ~P_-20, _dated J_anuary21, 1980.

~ -- ' ** :-...

2. ~* :_L-~t.~e~ :~rom o. P. Hoffman of...Consumers Po~:er Compar~y to D. C.rutchfiel d of NRC,

Subject:

Proposed T~chnical Specific~tions Chinge Relate~ to

. Containment Spray Initiation Time, Docket No*. 50-255, License DPR-20, dated November 24, 1980.

I

Appendix A I

SEP Containment Analysis and Evaluation for the Palisades Nuclear Po\\Ver Plant Contents

• 1.0 Introduction and Background 2.0 Containment Functional Design

_. 2.1 Review of Palisades Containment Design Analysis 2.2 Primary. Sys_tem_ Pipe. Break

• 2.3 Secondary System Pipe Break 2.4 Reanalysis of Pali$ades Containment De~ign

• 3.o*Primary System.P1pe Break 3.1 Initial and-* Boundar.y-Conditions 3.2 Slowdown Phase 3.3 Reflood Phase 3e4 Post-Ref lood and Containment Response

• 3;5 Containment Response Results 4.0 Secondary System Pipe Break* -

• 4.1 Assumptions 4.2. Containment Response Results Ca lcul at ion Page_

1 1

• 2

3.

3 4-4 4-6.

7.

9 14 16 16-17

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Table 3.1 3.2 3~3 3.4 4.-1 LIST OF TABLES Palisades Double-Ended Suction Leg Break Slowdown Energy Ba 1 ance.

Palisades Double-Ended Suction Leg Break Reflood Energy

• Balance.

Heat Structures in Palisades Containment Model Palisades Containment--Initial Conditions Main Steam L foe Break Mass/Energy Release Data

~age 23*

24 25 26 27

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Figure

3. 1

.. 3.2 3.3 3.4 3.5

~ 3,.6' 3.7 3.8 3.9P

_ 3-. 9T

.a LIST OF FIGURES

• Break Flow,- Steam* Generator Side of Break Break Enthalpy, Steam Generator Side of Break Break Flow~ Pump Side of Break Break Enthalpy, Pump Side of Break Break Flow - Steam Generator Side Break* Flow.Enthalpy_-* Steam Generator Side

. Break __ Flow -. Pump Side.

Break Flow Enthalpy - Pump Side

• C.o.nt'afoment Pressure* - S"uperheatec:i" Steam Only (Met~od-A) - :.. : -

• containment.T~mperature -: Superheated Steam Only (Method A).*

Page 29 30 31

. 32 33 34

. 35

36.

37' 38 3elOT Containment Pressure - Superheated Steam Only (Method B) 39

3. lOT Containment Tempe_rature ~ Superheated Steam Only (Met.hod B) 40

• 3Q llP *-~-Containment. Pres-sure.* Water -and Superheated Steam (Method A)..

• 41 3.llT Containment Tem~e'rature - Water and Superheated Steam", (Method A) 42

3. 12P 3.12T 3~ 13P

3. 13T

3. 14P

3. 14T

4. 1 4.2 4.3P 4.3T-A 4.3T-B 4.4P 4.4T 4.SP 4.ST 4.6P 4.6T

. 4.7P 4.7T 4.8P 4.9T Contain.ment Pressure - Water and Superheated Stearn (Method.B) 43 Containment Temperatur:e Water and Superheated Stearn (Method B) 44 Containment Pressure - Wafer and Saturated Steam (Method A) 45*

Containment Temperature - Water and Saturated Steam (Method A}

46 Containment Pressure - Water and Saturated Steam (Method B).

47 Cont a foment Temperature - Water and Saturated Stearn (Method B) 48 Main Steam Line Break Flow 36" Diameter 49 Main Steam Line Break Flow 24" Diameter 50 Comparison of MSLB Case 1 and Reference 1, Containment Pressure 51 Containment Temperatures - MSLB Case 1 with Spray. a~ 84 seconds 52 Containment Temperatures - MSLB Case l with Spray at 30 seconds 53 Containment Pressure - MSLB Case 2 54 Containment Temperature - MSLB Case 2 SS

• Containment Pressure - MSLB Case 3 56 Containment Temperature - MSLB Case 3 57 Containment Pressure - MSLB Case 4 58 Containm~mt Temp.erature - MSLB Case 4 59 Containment Pressure - MSLB.Case S 60 Containment Temperature - MSLB Case 5 61 Containment Pressure - MSLB Case 6 62 Cori ta i nm_ent Temperature - MSLB Case fr:

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1. 0 Int reduction and Background-On January 1, 1980 the* Office o'f Nuclear Reactor Regulation {NRR) initiated a two-year program with Lawrence Livermore National Laboratory

{LLNL) titled. Containment Analysis Support for the Systematic Evaluation P~ogram (SEP).

This program is directed toward resolution of SEP Safety Topic Vl-2.D, Mass and Energy Release for Possible Pipe Break Inside Containment, and Safety _Topic VI-3, Containment Pressure and Heat Removal Capability *. The CQntainment structure encloses.the reactor system and is the final barrier*

against the release of radioactive fission.products in the event of an*"

~ccident. The containmerit structure must,- therefore, be capable of

.. *. ~

withstanding, withqut loss of function, the pressure and temperature conditions resulting *from postulated LOCA and steam line break accidents.

Furthermore, e~uipment having a post-accident safety function must be environmentally qualified tor the resulting adverse pr~ssure and ~emperature cpnditions.

To accomplish* the o.bjectiv_es of_ this-program, first, the existing

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docket information was reviewed and evaluated and then additional analyses i

\\

were performed as requ*ired.

The purpose of this report is to document

• • original analyses performed by the LLNL on*the containment functional design capability of the Palisades Nuclear Power Plant and evaluate existing analyses-for conformance.with current NRC criteria.

2.0 Containment Functional Design Palisades is a Combustion Engineering PWR licensed to operate at 2250 MWt.

The primary coolant system is a two loop system consisting of two steam

• generators with two co_ld l~g loops per steam generator. The containment systems include the containment structure and associated systems. These systems include containment heat removal systems, containment* isolation systems a~d a*tombust~ble gas control system.

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The containment is a steel-lined, pre-stressed, post-tension concrete structure with a net free.volume of 1,640,000 cubic feet. The* containment structure houses the nuclear steam supply system, including the reactor, steam generators, reactor coo*lant pumps and pressurizer, as well as certain components of the engineered safety features sys.terns.

The co.ntainme.rit is designed for an i nterna 1 i:>>re,ssure of 55 psi g and a temperature of 283°F.

2.1. Review of Pali"sades' Contafnment Design Analysis There are two separate* calculations which make up the cont"ainment design analysis *. First is the mass and... energ/release' analysis for postulated LCX:A's.

This consists of.a blowdo~, reflood and post-reflood phasese The results are mass and energy release rates into the containment.

For PWR's there are two poss1b1e b_reak types which must be analyzed 9 a prima*ry system.

pipe. break and a secondary'system p1pe break.- A break on the primary side

• ~enerally results in the most severe pressure response in the containment while a break on the secondary side results in the most severe temperature conditions in the containment.

The second calculation which is performed in the containment design analysis is the containment response calculation. This results in the containment temperature and pressure response to the mass and energy release from the postulated breaks.

The acceptance criteria used to evaluate Palisades* Containment Design Analysis was based on the Standar~ Review Plan (SRP).

In order for the containment design analysis to be found acceptable both the mass and energy release and containment respo~se calculation must meet the acceptance criteria specified in the SRP.._.__-~~-. ----.-* -----....-'=""'

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'I 2.2 Primary*System' Pipe** Break The SRP* specifies several acceptance criteria applied. to the mass and energy release analysis for primary system pipe breaks.

Among these are break location. In. the Palisades FSAR the most severe mass and energy release rate.

calculated for containment design was done assuming a double-ended cold leg

_d.ischarge break with no accounting for the reflood phase or energy in the.

~econdary system.

Since this does not meet the acceptance criteria specified

.~n-~he SRP ~r previo~sly accepted methods by the NRC staff,:thi_s an~lys_is is

~nsuitable fc;>r containment design calculation. Since the mass and energy.

release.rate analysis is found unacceptable, so is the containment response calculation based on the mass and_ energy release rat~s.

f *3 Secondary System Pipe Break The mos~ recent secondary system pipe break analysis that was reviewed was submitted by Consumers P_ower co*. to the U.S. NRC on January 21, 1980.1 In this analysis a main steam line break (MSLB) analysis was'performed~ In this analysis the blowdown'of one steam generator with feedwater isolation and loss-of-offsite power was c*onsidered.

HoWever, the analysis did not address the possibility of a single failure of one of the main ste~~ isolation valves*

which could lead to the blowdown of both steam generators. Therefore, the analysis was considered incomplete and unacceptable. A more thorough discussfon of the MSLB analysis is given in Section.4.0, Secondary System Pipe Breaks.

1 Palisades Plant - Autmoatic Initiation of Auxiliary Feedwater Ssytem at Palisades Plant, Do"c:ket 50-255 - License DPR-20, January.21, 1980 letter from Roger W. Huston of Consumer~.Po~r. Co. to Dennis L. Zieman of NRR~

NRC. * * * *.

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2.4 Reanalysi.s of Palisades.' Containment Design As mentioned earlier. in Section 2.1, Review of Palisades' Containment Analysis, there are two separate calculations which make up the containment design analysis, the mass and energy release rate and the co~tainment

. ;:2

. response.

The mass and energy_release rate.c~l~ulation can be the result of

~~-~

.either. a primary or secondary pipe break.

The primary pipe break generally results in the.limiting con.d.ition fo.r calculating the peak. pressure inside the containment. Th_e.secondary pipe break analysis gener.ally is. the mos~_ limning case for temperature conditions inside the containment.

Both of thes~~

_analyses were performed.and are discussed be.low.

... -.. *: ~-

3.0 Primary System Pipe Break *-

For a primary system pipe break there.are three phases. i!l.. calculating mass.

and energy release rates. :rh.es.e are. the blowdown, reflood, and post-reflood computer codes used.

The primary limitation was the carry-over rate fraction which was discussed in some detail in the Methodology Report for the.Palisades Nuclear Power Plant. In.general, the analysis was done in a manner that conservatively ~stablishes the containment design pressure; i.e., maximizes I

the post-accident containment pressure.

The worst break location was determined to be*at the cold-leg pump suction side because-~f.the consideration of energy*input during the reflood phase and flow resistance.

3. 1 Initial and Boundary Conditions The initial and boundary conditions for this analysis were defined to satisfy the requirements of the Standa~d Review Plan.

The sin~le failure assumption for these analyses was a loss of one diesel generator.

The initi_al

~ *-. - -

power was specified to be 102% of s_afeguards design rating or 2690.76 MWt.

A

(

steady-state mass and energy distribution was provided in the primary and secondary coo.lant systems consistent with the conservative core power.

Th~

break flows were calculated using a discharge coefficient of 1.0, with the Henry-Fauske correlation for subcooled and the Moody correlation for saturated fluid. The safety injection flows were minimum, corresponding to the diesel generator failure.* The mass and enetgy rel_ease analysis was performed with-RELAP4 MOD6.

Steam quenching by the safety.injection water.occurred-due to *

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the homogeneous equilibrium (HEM) assumptions of the RELAP4 MOD6 code. All of

.

• o _.

the safety injection water temperatures were defined to be 9Q F.

Scram was assumed to occur with a low pressurizer-pressure of 1750 psia.

A l.0-second delay time was used in the *model for conservatism; however, the moder~tor 'reactivity-feedback caused core shutdown before the control 'rods were effective. The main coolant pump power was tripped off at the time of th~ break. Steam generator i so 1 at ion was. initiated one second after* the break and the valves were assumed to completely close in five seconds. A 15-psia constant containment backpressure was assumed to maximiz_e mass and energy release ~hroughout the blowdown.

The end of blowdown was defined as the time the primary system pressure reached the containment design pressure of 55 psig.

The RELAP4 input deck was obtained from NRC and. was* carefully reviewed for code options, initial.and boundary conditions. The plant physica*l description was.generally assumed to be correct. *Additional information required for* the analysis was obtained from the Palisades FSAR, and telephone conversations with C. Tinkler of NRC and D. Vandewalle of Consumers Power Company.

A thorough discussion of the.model can be found in the Methodology Rep~rt for the Palisades Nuclear Power Plant..

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4 3.2 Slowdown Phase the blowdown analysis* results are summarized in Table 3~1 and Figures 3.1 through 3.4. Table 3.1 itemizes the energy sources for the ~uration of the blowdown which endea at 20.4* seconds after the break.- The total energy *.

released during blowdown was. approximately 253.4 million Btu *.

• Figures 3.1.

through 3.4 provide break-flow and enthalpy out the break.

The accumulator flows start-after 16 seconds and do *not-reach maximum.flow rates -by the. end-of::.blowdown ~

• The pumps coast down* at different rates.* -The -

\\A pump nearest the.breakreathes.zero*rpm before tw0 secanEls*because of reverse flow through.the pump.: The pumps. were not- 0al1owed-£0* reverse~ providirig-a*:.-

conservatively h"igh resistance whic;h allows~more flow through the*steam'*-

generator side of the break.

The other pump in the broken loop coasts doWn to zero rpm at about 11.seconds. The pumps in the unbroken loop continue to have a positive rotation"through"o*ut tile blowclowns although it decreases -to 500 rpm in about:10 seconds. -Although the scram occurred at about eight seconds,.

moderator.reactivity feedback had already reduced the* power to less than 7-1/2% of the initia~ power.

The mass and energy release rates and energy sources were qualitatively compared to the CESSAR results for a double-ended suction leg slot break with the same area.

~he similarity of the results suggests the RELAP4 calculated blo\\vdown results are reasonable.

3.j:Reflood Phase The reflood analysis for the double-ended pump-suction break was assumed.

• to immediately follow the LOCA blowdown analysis. The analysis was perfonned using RELAP4 MOD7.

Within the limitations of RELAP4 MOD7, the analysis was performed in accordance with the requfrements of Section 6.2.1.3 of the Standard Review Plan (SRP).

6 -

I Initial conditions for the start of the reflood analysis were based on the

(

end-of-blowdown (EOB) re'sults.

EOB was defined. to occur when the primar-y system pressure fell.below the Palisades containment design pressure of 55 psig w~ich occurred at 20.4 seconds after the start of blowdown.

At that time, th~

core power level had dropped to 159 *.41 MWt or approximately 6% of the initial power.

The accumulator flows had been initiated on low cold leg pressure trips of 262.5.psia which occurred at about 16 seconds into the blowdown and had reached a total of 5900 lbm/sec at the start of reflood.

coolant pumps had coasted down and the rotors were locked *.

The reactor * * ~

For the reflood aria_lysis~ the primary system*was initialized.at.the**::

containment design pressure, 69.7 psia. The primary system junction flows were zero except for the accumulator and lower plenum inlet* and-outlet junCtions. Heat conductor temperature and primary system* state conditions were established based on the EOB conditions. Core power continued to***

r--

decrease according to the ANS decay heat curve.

l A natural circulation heat transfer model was use*d in the steam generator secondary to maximize.the energy transf.er rates to the break. The primary coolant pump rotors were assumed locked to conservatively.provide resistance to flow.

A closed valve was modeled in the intact cold leg of the broken loop to conservative1y increase the flow through the steam generator *.

For numerical stability 6f the RELAP4 computer code, the Emerge~cy Core Cooling System (ECCS) flow was modele9 as being injected directly into the downcomer at a temperature of 300°F.

Plant specific information was predominantly derived from a RELAP4 Reflood i.nput listing for the Pali*sades power plant which was obtained from the Nuclear Regulatory CollllT!ission (NRC), and from the Palisades Final Safety analysis Report lFSAR).

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Several sensitivity calc;ulations were performed to evaluate various.input

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model and code options. The results of the sensitivity studie~ are document~d in the methodology report. The-Palisades reflood transient results are presented in Table 3-.2 and F1gures 3.5 through 3.8. Table 3.2 is a summary of

• the energy balance at the beginning and end of reflood *. Figu~es 3~5 throug~

3.8 provide break flow* and enthalpy out the break.

The accumulator flow is initiated at 5900 lbm/sec and quickly rises to

.

• 6400-lbm/sec.

The flow remains' constant until 40 seconds, and then-is ramped down.too lbm}sec at.SO seconds when-the accumulate~ is e~pty. The HPri16~

~ome~- on"at 0.6 ~~c~~d~--~rid ~einains at-about 650 gai/min fo~- the auration cit the trarisient. Tne LPiflow comes.on at 7.6 seconds and varies in magnitude betwee~ 400 and 600 lbm/sec:for the duration of the transient, depending on

. the primary system pressure *..

The primary system.pressure starts at 69.7 psia, increases*to 160 psia at 20 seconds, and then slowly decreases to 100 psia. The pressure increase can.

be attributed to steam binding in the primary system._ As the ECCS water.*

enters the core, it boils away f,aster than the generated steam can escape through the break. After 20 seconds, the core is quenched and the steam generation rate reaches a new pseudo-steady-state with the break flow.

Normally, t~e end of reflood is defined as the time when the core recovers to within two feet from the top of the core.

In the case of Palisades, the maximum mixture level is less than seven feet at 50+ seconds into the transient which is still four feet below the top of the 11-foot core.

However, the core-stored energy was essentially removed at 30 seconds into the transient.

The reflood calculation was extended to 100 seconds to determine when and if the steam generator. side break flo.w *woul.p begin a rapid d-ecay expeCted after the accumulators emptied at 50 seconds. *Since the rapid flow decay did 8 -

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c not occur; the reflood calculation was continued beyond the time when the containmen~ calculation 'predicts the peak pressure and temperature at 84 seconds after break or 64 seconds after start of reflood. Because the safety injection water was assumed to be at 300°F, the extended duration of the reflood analysis is considered to provide a conservatively high energy transfer rate to the secondary.

3.4 Post-Reflood and Containment Res.ponse Ca lcul at ion.

• *.. The containment model-used was based on a CONTEMPT deck received from the NRC~ The mass and energy flows to the c:ontainment were replaced and-the remaining dat~ carefully *checked against the FSAR and other sources. The analysis was performed using CONTEMPT-LT/028 *

. The heat structures used are listed in Table 3.3. All the structures are repr~sented in rectangular geometry.

~T~e thermal conductivi~y_and the*

volumetric heat capacity wete checked for the four materials used:* steel, concrete, insulation, and* air (gap). The heat capacity was found t~'. be about two orders of magnitude low for fosulation and was changed. Tagami/Uchida

• boundary condition's were used for all h.eat structure surfaces except the base slab, which was assumed to be covered with water.

Th~ Tagami peak time used was 20 seconds, 0.5 second before the end of blowdown.

The basic assumption was that off-site power was lost and that one diesel generator failed to start. The cooler and spray pump start times are based on the generator loading sequence.

It was assumed that one fan cooler was operating and that it started at 23 seconds after the break.

The heat removal rate was variable, ranging from 97.5.MBtu/hr at a containment temperature at 3S0°F to 3.0 MBtu/hr at 104°F.

The one operating diesel generator was also assumed Capable. of poWering two spra_y pumps.

Both pump~_ tog~1her were capable of 1.34 Mlbm/h.r (2700 gpm) with spray efficiency of 90%.

The containment spray started at 84

- 9

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seconds ani::I used wa~er _fron:i. the R~f.ue ling Water Storage T~nk unt i 1 30 minutes when the tank was empty, at _which time the.water.source switched to the containment sump.

The heat removal rate for the shell and tube heat exchanger

• in the containment spray_system is computed in CONTEMPT.

The entered

, parameters ~ere: the.PrOdt.!.Ct of ~he heat exchange surface area.and the overall heat transf~r coef:f~cient was 2.28_MBtu/hr/F; the coolan~inlet tempera~ure* wa~ 1_~4°~; __ a~.d. ~h~.c~o~~nt flo~ _rate.~a~ _2o0 Ml~/hr~

~eneral initial cqndi~ions are given in Table 3.4~ Initial ~onditions for the primary system _refer.to.the enc! of blowdown.

"'lo water was introduoed to

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the. co_ntainment as an initi.a~.step. input. The evaporat.ion-.cond.ensation mo.de1

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in. the.containment.was bypassed.unti 1 _the end of blowdown. -lhe fraction of

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wa 1. l ~".'. coo1{ng CQ.il_.c~n_def!sat_e.t.ra~s.fer.re.d from.the superheat~d _cont~ir:unent

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analysis and three different assumpti'ons made about the mass and energy release during reflood, resulting in six cases.

The blowdown mass and energy release was the same ~or all cases.

Peak f.low was about 7-7,000 lbm/sec at 525 Btu/lbm, and the blowdown ended at 20.4 seconds.

The reflood data lasted 100 seconds, whi..Ch corresponds to 120.4 seconds after the break.

The core was not covered, or even two-thirds covered at this time, but it was substantially cooled. Therefore, the end of the RELAP4 reflood run was oefined to be the em!' of reflood even though the acc*umulator flow had been ramped down to zero bet"ween 60 and 70 seconds after the break.

During the post-_reflood period, decay heat, heat from the secondary system, and heat from the heat structures in the primary system are released to the containment.

The decay heat is released over the duration of the run based upon 'the ANS standard decay heat curve plus 20% and an ultimate reactor

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'l power of. 2638 MWt plus 2% for instrument error (excluding pump heat). The heat from the secondary *system ( 61 MBtu) and the primary heat structures (53 MBtu) was all released by one hour after the break. A linear ramp to zero was used.

The amount of heat.released to the containment by the secoDdary was detentiined by obtaining the stored energy in the water.in steam generators and in the SG tubes at. the end of the RE.LAP4 reflood calculation. Assuming that this.- was based on 32°F and that the entire steam generator would be* at

• A 212°F after one* hour, the amount of heat available to be released was * * -..

computed to be 61 MBtu.*: Th1s is conservative because the co~t~inment pressure will not decrease to* atmospheric' pressure in one hour, and so the secondary system will be hotter.than 212°F.

For the primary.heat structures, the*

energy stored' in all of the heat structures used in the reflood model, except the core (fuel rods) *and the-steam generator tubes, was used in the model.

It

.was conservatively assumed that all this metal *would be at 212°F aft'er one hour, with the difference (53 MBtu) being released to the containment.

During the post-reflood period, *two* different methods were used. which differ only in the manner in which the energy from the secondary system and primary heat structures is released after reflood.

In Method A, the energy is released direct1y to the containment without considering the primary coolant system.

The amount of mass accompanying this energy re*lease is required, and it is 6b.tained by assuming that the h~at is used in converting water-at saturation to steam. A typical value for the heat of vaporization at the pressures experience~ in the containment for the first hour is 925 Btu/lbm, and this has been used to calculate the mass release rate. This method is conservative, since after some time the water in the primary system will cool to below ~he boilJng point and most of the~decay heat will go into heating.the water to saturation and leaving very little to generate steam.

The systems (.,.,.~.,-""°r.'"'~.-:0~'"'.'~7"'<;,,. *. ~'::*'C"~.'*"'""'"=*--~--,-,:.....,...,,,.~,.,.,:,,..,..,...,~--,-:--*'":°.",-"C:-:*0':-~-_.;',, -~ -.~*:* -.-, --.-~*'*." * - *.., **.... '..'Ci' * * '.,:~~; :'.'.,--;i-'C,"";:;...,.......,,.. _:,.., *;'O*::~:- '-,",":'/,'.'~-C:"~ c-,*;;C :.*.-.~::-:-." -_"_." -._-.....-:..,_.,..,,'.'°'.:°;~

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(HPIS and LPlS) that inject. water into the:.primary are* not modeled in M~thod ;A t"

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since the mass and energy.flow:.from the primary is.already calculated as described above.

In Method B, the decay heat and heat from th~ secondary system and from the primary* heat structures is passed to the compartment throµgh the reactQr coolant system.. The model in.CONTEMPT then determines how much steam is proouced, how m1.,Jc.h., heat goes. jnto increasing water temperature~ and S!J one Thi.s _is much more realistic_ tha_n Method A ~ince it _allows the steam pro9uct.ion

• to decrease.wi~h time. __ lt_j~ still conservative s~rycethe hea~.-inpu~ *pa~.. ~een calculated to be conse.ryatively_ hig~ *. In Method _B, injection _into tb~_pi:-jf!lary

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• i...

is-explicitly mode1ed. _The.LPIS (?000 gpm) and t~~-~PIS.(450 gp~) bc;i~bta~e

. water from the. refueling w~ter storage tank until it emptie~. at.. 30 _mjnyt~s..

After_ ttiat, only the HPI.S continues, taking water.from.. the.containl'l!ent sump.

c....:. - :.

There is no heat exchanger'on either of these systems. The containment. sump

. - -. ~..

i~ gradually coo le~,. since the water that is recircu)ated through ~?*~~~::-* __

containment spray system does pass through a heat exchanger.

Three assumptions were considered for the mass and energy release to the containment during reflood.

In the first assumption, only the steam* flow from the SG side of the brea~ was used from the RELAP4 results.

To be conservative, it was then assumed that all of this dry steam was superheated to 1300 Btu/lbm (about soo°F), and the energy release rate was obtained by multiplying the steam flow rate ~y 1300 Btu/lbm.

The actual eff1uent enthalpy is.*about 1200 Btu/lbm for the first 20 seconds of reflood and gradually decreases to about 600 Btu/lbm after that. Thus, assuming that only dry superheated steam is released is conservative.

In the second assumption, the mass and energy flow rates from the SG side of the break were_- used, but the energy-flo~ was augmented to account for superheating.

At 70 psia, saturation is about 310°F, and the specific 12 -

... ~..

.e enthalpy of steam is 1185 Btu/lbm.

At this pressure, the specific entha,.py at soo°F is only 1282 Btu/Tb~. While the SG tubes are a little below 500°~

at the start of*reflood, their temperature is on the order of 3S0°F at the end of reflood. Therefore, adding 100 Btu/lbm for each pound of steam flow is conservative. The steam flow rate used to calculate this adde~ energy was the same as that used in the first assumption.

The additional energy was about 8%

of that computed by RELAP4 at the beginning of reflood and about.6% at the end of.reflood.

... -.. ~...

.. In the third assumption, the.. mass and energy 'released from the. SG 'side* of the _break *by RELAP4* were* used directly.*. The liquid phase falls to the,pool as released *.. Natura11y, this_case,re~ults in lower peak temperatures ancL *.--

pressures than the.superheated-steam-only case, but it is more l'.'eal_istic and it is ~conservative *.. -

For all three assumptions, the release of water from the pump side.of.the break is ignored. This releas~ is all liquid phase and go~s directly to the containment sump.

The amount of water in the sump has a negligible.effect on the temperature and pressure history of. the containment vapor region. The RELAP4 model had to use ECCS water at 30o0r in order. to avoid instabilities, 0

whereas the ECCS water would actually be about" 100 F.

Since the water coming out the pump side of the break wi 11 have had no.contact with the core and little with any of the metal enclosing the primary system, it should not be signi'ficantly warmer than when it left the accumu.lators.

In.view of the large difference between th actual and the model ECCS water temperatures, neglecting the liquid flow from the pump side of the break is more realistic than*including it.

13 -

(

\\ __.

l~...

3.5 Containment Results The results of the CONTEMPT runs are shown in figure 3.9 through 3.14.

Figures 3.9 and 3. iO.show the results.for the case where only dry, superheated steam flow from the SG side of the break*:was considered, and _the energy reiease rate during reflood was obtained by multiplying the steam flOw rate by 1300 Btu/lb~ {which is approximately the specific enthalpy at soo°F and*

70 psia).* *Figure 3.9 shows the.results for.Method A and.Figure:3.1o*shows the

~es.ults for :Method B.

• The twe *cases. are identical to 120.4 seconds since the differe.nce is in how the mas_s and energy releases are handled after reflood.

The figures.* show that the containment atmosphere reached almgst 70 psi.a and 32S°F at 84 seconds. just before.the containment spray began.** ~*Since only dry steam was released to the containment during.reflood~.the containment spray has an ~mmedi ate and dramatic-effect on the* containment vapor. regi.6n....

temperature and.pressure~ *The-peal<" pressure equals the containment design.

press~~e of 69.7 psi a.

In view of t~e c6hservative assumption pf -releasing only dry, superheated steam during reflood, this is riot considered*

significant. It is inconceivable that the superheated. steam could flow from the steam generator to the break with the saturated water and not mix to form a homogeneous, t we-phase fl ow.

Since the Method A assumptions are not suitable for a long-term model, the run shown in Figure 3.9 was terminated at two hours, while. the Method B run in Fi.g.ure 3.10 was continu~d to ten days.

The results of the two methods are quite close at two hours.

The dip in the atmosphere pressure and temperature at 30 minutes (1800 sec) in Method B is due to the shutdown of the LPIS at the time when the RWST runs dry.

This does not show up in Method A since the primary system is not modeled.

The change in slope at 30 minutes in the Method A result ;-s due to the fact that the source of water for the containment spray changes from the RWST to the warmer containment sump.

The

/

containment vapor region temperature reaches 13S°F at about 8.08 days (698~400 sec) in Method"B (see Figure 3.10).

• Figures 3. 11 and 3.12 show the results for the release of a two-phase mixture with the energy flow augmented to account for superhe-ating the.steam.*

fraction.

Peak. pressure is about two psi below.the design pressure, and the pe*ak containment atmosphere t.emperature is about 283°.

Behavfor a,:fter a few

hundred seconds is nearly identical for all the A cases and all the B cases *

.J.hi s is to be expected s i nee events are dominated by the absorptive* capacity of the heat structures* and the effect of :the sprays *. The. c.o.n~a_inme_nt* _

• 0 atmospheric temperature* reach~s.135 F at about 8 days.*

r*.

~

.In view* of.all the conservative assumptions made elsewhere';.the:. r.esul.ts shown in Figures* 3. ll and 3.12 are sufficiently conservative and meet the

Standard. Review Plan' requirement. for superheated steam.

S.ince ~.he _effluent to

_the containment will certainly.not be dry steam in view_ of the carryover rate fraction in the core; this a.ssumption of.wet.steam with the steam.fract.ion-

. \\__

arbitrarily superheated appears to be the maximum which can be justified as realistic.

Figures 3.13 and 3.14 show the results when _the RELAP4 calculated releases from the SG side of the break without adding any energy to account for supe_rheating steam release during the reflood period. *Peak pressure was over 4 psi below the design pressure, and the peak containment atmospheric temperature was more* less than 273°F.

The containment atmospheric 0

temperature dropped to 135 F after about 8 days.

So many conservatisms, including the.use of 300°F ECCS water, were made in arriving at the releases to the containment during *reflood and post-reflood that these results are definitely conservat iv~. However, they do not meet the SRP mandate for superheated steam.

15 -

(

-~*

4.0 Secondary System Pipe Break Analyses of the containmen~ response to a secondary system pipe break were also made~ For PWR's the most limiting break location is a_main steam line break with pure steam._ blowdown.

In the case of Palisades the results show

• that a sing.le failure assumption which allows both steam generators to blowdown:will produce peak pressures and temperaturs which* exceed.design valu~s~ The.model *and assumptions that were used in~analyiing the main *steam line break are given in the-following discussion.

4.1 Assumptions-A main-steam line-break (MSLB}* analysis was performed by.Consumers Power Company. 1 Results given in this 'reference are used for comparison pure.~~es. In particular, mass and energy release data for the full p~wer MSLB case ~iscussed in the_Palisades FSAR is provided in Table 1 of refererice.1.

The Palisades FSAR full-power analysis ~ssumed:

(1) A double-ended guillotine rupture of a main ~team line inside the containment.

(2) A reduction in feedwater flow from full flow to zero over the 60 seconds immediately following scram at less than 2 seconds on high containment pressure.

(3)

Both main steam isolation valves ~ould close on l~w steam generator pressure (500 psia) caus.i.ng the unruptured steam generator to isolate in eight seconds.

(4)

Off-site power was available.

1 Palisades Plant - Automatic Initiation *of Auxiliary Feedwater System at Palisades Plant, Docket 50-255-Lic~nse DPR-20, Janury 21, 1980 letter from Roger w. J'fuston of Consumers Powe-r Co~ *to Dennis L. Ziemann of NRR, Nuclear Regulatory Commission.

16 -

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

---

~~--~--:.~~-~~-~--~~:~~-- *,-.*--.-.-..--..,_--;-: ::::-~=*: ~: ::..~*~;c.,:~*~-~:~~7~~~::::;:-;r* ;.:~ -::~~~-;~*~;-:~-.-**~?-~'::7,7':?w::-::.~~T;"~~~-~;:-~:i_~:.~~~;;:~'~:-~.~*.-~*:; **~;~}~*;:.:::::~~2-T.f.}-:"f.\\?~:?~l\\.?:~~~,~~~;._~.~£-?':-.:_.

~ --~~~~ ~-=-:

-*.. --*-* ----- ---;----.. ~-*.

. e

--~

(S)

• Only two containment spray pumps were available; no air cooler~ were

(

available.

-.1

• i

~-~

.{.,\\

':}.

'~

. \\.__

-~

(6) Pure steam blowdown (no moisture carryover).

(7) A highly conservative containment heat transfer model.

In addition to the results reported in reference 1,* a number of analyses for Palisades containment response to the MSLB were made.

The analyses employed RELAP4 to obtain.mass and energy release rates and CONTEMPT-LT/028 to obtain containment*response.

The RELAP4 mass and energy release rates were obtained using a simplified

~odel based:on*one volume, one heat cond~ctor; one break.junction,-and bne: :

feedwater fill junction. Two break sizes and three feedwater f1ow ass~mptions were analyzed.

The resulting break flow rates are summarized and.compared to reference-1 results* in Figures 4.;l and A.2. For-the RELAP4 analyses~ the.:::.*

steam generators were assu~ed to be at 770 psia 9 with an average water enthalpy of 552.2 Btu/lbm and conta.in 128,456 lbm each.

The primary_ system was assumed to be held constant during the blowdown with 513.S3°F local temperature and a 952.Btu/hr/ft2J°F *heat transfer coefficient in th~ steam generator.

Fi_gures 4.1 and 4.2 show that the main effect of the feedwater is to prolong the time of blowdown period and increase the total mass and energy to the containment.

The containment responses for a number of MSLB cases have been compared.

The results are given in the following discussion.

• 4.2.Containment Response Results Case*l The first case sele~ted for analysis was intended to determine if the CONTEMPT-LTi028 model used would give.results similar to those in reference 1 if similar assumptions were employed.

Consequently, two CONTEMPT runs were..

. I

made, using the reference l-mass and energy release. rates. These CONTEMPT

(_

runs assumed two spray pumps and one fan cooler were available.and a TPEAK of eight seconds for* the CONTEMPT Tagami/Uchida heat transfer ~orrelation.

The one run assumed that off~site power was not available so the two spray pumps were *started at 84 seconds.

• The other CONTEMPT r*un assumed off-site.

po'wer was* avai'lable so-the spray pumps were started at *30 s*eco-nds*~'

'"The *containment' histories-for the two Case 1 runs* are*compated* to*the=

reference. l p.ressure"°lii:story* in** Figure 4.3P*.*. The c~mparison-.b.etwee~ the results with spray after 30 seconds and the reference 1 results is close, indicating_ an acceptable CONTEMPT model.

The containment t~perature hiStories for t*h~ Case-1 model: a*re* shown in Figu*res 4.3T-A-an.d 4.3T--s.-'

Case 2 The purpose of the Case *2 an*a lys is was to determine -if the one- *volume RELAP4 model was adequate for* obtaining mass and energy release rates to the*

containment.

The blowdown mass and energy release rates for various feedwater and break area combinations have been.noted in Figures 4.1 and 4.2 For Case 2, the ruptured steam generator blowdown was simulated by the 36-inch break (area = 6.12 tt2) with m*ain feedwater only. Th1s feedwater flow was initially 1650 ib/sec and ramped down to zero flow at 60 seconds.

The unruptured steam generator was assumed to isolaie (MSIV closure, not failure) so _the mass and energy rel ease ra.tes were obtained from ref re nee l

• The mass an~ energy release rates for the two steam generators were added for input to the CONTEMPT model.

The CONTEMPT assumptions for Case 2 were similar to the assumptions for the Case l run with spray. after a 30 second_ de 1 ay.

The containment pressure

-and temperature response from Case 2 *is s*hawn in Figures 4.4P and 4.4T, respectively. The peak pressure is ~bout 65 psia, which is sJightly less than

.e for the-comparable Case 1 run and the reference 1 value. Therefore, the one

(

volume RELAP4 model was Judged to be adequate for obtaining mass and energy

\\,

I.

release rates. It is noted that complete phase separation is mode,led in the R,ELAP4 analyses so that. pure steam blowdown occurs.

Case 3

,*,c-*. Cases 1 and 2 established th,at the C,ONTEMPT and RELAP4 models _were:

-adeqiiate for: obtaining cont'a'i,nmen}:* 'response to a MSLB.

Cases 3 and 4 were desi-gned to determine the response of the Palisades containment to the !'1$~8 for blowdown, of bpth steam generators, with off-site power available. Case 3 ass4111_ed that each steam generator would blow down through a 24-inch break

• ( 3.06 f1:2) *. Case 4 ass.urned that the ruptured steam generator would blow.

down through_ ~he :~6::-i.nch_.* ~(6.n ft 2) break and the unruptured steam generator would blow *down through a 24-inch br:eak.

The assumptions used for Case 3 include:

(1) If off-site power is available, the spray pumps will be activated at 30 seconds after high containment pressure (5 psig).

High_

contai nrilent pressure occurs in* about l. 7, seconds.

A conservative value of 33 seconds was used in the analysis.

• (2)

• Singleafailure is the Main Steam Isolation Valve (MSIV) failure causing both steam generators to blow down.

(3}

Ruptured steam generator blo~s down through one~half the maximum 2

areas, or 3.06 ft, as areas larger than.this would not give a pure stean:i blowdown.

(4)

Isolated steam generator blows down through one-half the steam-line area becuase of MSIV flow a~ea restrictions, thus through 3.06 ft2*

(5)

All.three spray pumps and a1J foyi: fan coolers will be available.

The single failure is assumed in the *MSIV.

19 -

--- ---:*-:*.... ~,.-~~:-**7'00*- *.* -.~:-_**~.:--::****~-~*::~-:..-;_"'-:.:.~~{;',-r-::'i~"'.1.. ~ ~~~-~c':r-~:::*-z-:-::7:":>..:.. r ***~~:::-:z;:::~Q-:*~*~:;.:;-;-;,_:,r.-.~::.!1::_;; ::::c"':T-.~~~=:7:,

-::::*~:-;:--~-::!.'*-:-::~~-( r:i:-~.~=-~:-::'-'""'~-.'t°~<<r~--:-... ;r!'-.-.... :_...::-:~-c~~:T:'!_.~:--*~-~~-~ ~i..

-e 4

(6)

The CONTEMPT time TPEAK for the Tagami/Uchida heat transfer

(

correlation was changed to 99 seconds to correspond with the end of I,

blowdown-.

• p) Maii'l feedwater is available to each steam generator at 1650 lb/sec initially and ramps down to zero flow at 60" sec.

The* re.sulting* containment pressure and temperature history is shown *;n figures 4.SP and 4.ST, respectively~ The peak pressure* is abOut 107 psia,

.whJch.is sub.st-anti ally. greater than the 55 psi g design-press-ure:. * * -

• .The assumptions used for Case 4 were identical to those used in Case 3

.except that the ruptured steam generator was allowed to blowdown through the maximum area of. 6~J2. ft~ *. *The. resultjng_ containment pressure.-and.*.*.. ;~::

temperature_ predictions are"shown in Figures 4.6P and 4.6T, respectively. The peak pressure. is about 106 psia.

Case 5

• Cases 3 and 4 assumed that off-site power was available. Case 5 was run to investigate the containment response for the loss of off-site power assumption.

Case 5 is similar to Cases 3 and 4 except for two assumptions.

First, because off-s~te power is lost, the spray pumps are not available until

84. seconds.

In addition, the loss of off-site power result"s in a complete and immediate loss of feedwater.

Case 5 was based on each steam generator blowing down through a 24-inch.

diameter brf?ak.

The pressure and temperature response is*shown in Figures 4.7P and 4.7T, respectively. The peak pressure is about 98 psia and a

~eak containment ijtmospheric temperature af ~ 465°F at 70 seconds. ~...... _._.. _____ _

-~:~~.. ~-~;~: ~-.-: -: *-;....'.... :\\.-.::-*:~~--~~:!:~~

Case 6

• Analyses have also been performed assuming a fix which would prevent the blawdown of both steam generators.

In this case the single failure assumption is loss-of-offsite power wittt {failure.of one diesel generator. The mass and energy release data used in the analysis is for the full power MSLB with one steam generator blowdown.

This is discussed by Consumer Power Cpmpany in.

Ref-~ *l and_"prov*ided in*Tab*le-4~*t;- The *assumptions made in*the mass and energy re:lease analysis are the following:'

1. - A double-ended guillotine rupture o*f a main steam line inside the

2.

containment.

A reduction in feedwater flow from--ful-1 flow to zero* over the 60 seconds immediately following scram at less than 2 seconds on high containment pressure.

3.

Both mainsteam isolation valves would close on low steam generator

~.

pressure (~00 psia) causing the unrl.ipture steam generator.. t?. isol_ate

• in eight seconds.

4.

Off-site power was available to maximize the rate of energy. transfer from the primary to secondary.

S.

Pure steam blowdown (no moisture carryover).

For the containm*ent response calculation the following assumptions were made:

l.

Loss-of-offsite pqwer and failure of one diesel generator.

2.

2 of 3 containment spray pump.s available.

3.

Containment spray initiation at 36.7 seconds (200 gpm) and full flow

~t 52.5 seconds (2680 gpm).

(Reference 2)

2. Palisades Plant--Proposed Technical Specifications Change Related to Contarnment Spray Initiation Times..

Do.ckt 50-255 License DPR-20, Nov *. 24, 1980.

Letter from 0. P. Hoffman of Consumer Power Co. to 0. Crutchfield of NRC.

(

\\

4.

1 of 4 air coolers* availabie *at 23 -seconds.

So Tagami/Uchida*, heat tr.ansfer correlation with Tagami peak time at er:id of b 1 owdown { 68 seconds).*

The results of this analysis are the pressure and temperature responses shown in Figures 4.8P and 4.SL: The calculated pea~ pressure*<

• js ~~~~ P~1a ~each~d ~t.6! se~o~ds. This is ~.2 psi below design.

The~ ~

~~i~~iat~d.*p~akte~~~ra~ure i~ 413°F reached at 37 seconds*~- -Therefore11-

• bas~d: on""this. analysis, a fix which would prevent the blowdo~n of both-= -

~t~~-*generatol'.'s ~uld limit.the calculated peak pressure.. to'_ 1.2 psia,,.- --

be lo~ desig~. *

• *:-... ~-

~ -.

--~ !

~ -

conciusi6n

.. B~sed or:i the. results.of Case l and 2, the CONTEMPT *and RELAP 4* models are* adequate* for.. obtaining 'containment respo~se :to a MSLB. *fhe results...

• of Cases 3, 4, and* 5 showed that the blowdown of both steam genera~or. -.-

will result in containment pressure exceed design values. This is regardless of whether off-site power is available.

The results of Case 6 showed that a design change which would prevent the blowdown of both steam generators would keep the containment peak pressure within the design limit.

22 -

1' I *...

~* -.

\\.

Table -3. l

. Palisades Double-Ended Suction Leg.Break-L.:;i ::..

Slowdown Energy Balance :-.;;.:.. '*-_-_

(Million Btu)

2.

~:-

.. ~.. :..-

~:-.

\\..:-*

-- -* _*:.;~~:;

I

~-..... _

Seconds

..=

-~

.

;.,, i I~*

~-~._*.. -*.

,.. sr.e=;..

Primary System Coolant Inventory Steam Generator Coolant Inventory S-econ.dary Flow to T~rbi~~1Tl-

• Accumulator System lnventory(2)

• Core Stored Heat Conductor Stored Heat( 3)

Decay and Fission Heat Note:

253.7 140.0 19.5 18.9 111.0 543. l

~- ' -

I

~'*,

.... ~: *. *.. ~

'... _,.. \\~*

Decrease

.. i..

.::.::.. -~*.... :

246.6 2.0

-9.2 0.7 4.6 3.7 5.0 253.4

~-"....

r

• *.o *.J I.J ::

I I

20.4 7.=

7. 1

. 138.0 18.8 14.3 107.3 285.5

/.

(1) ~low continues until valuve is fully closed at six seconds after the break. Energy value. is net loss for steam and feedwater flows.

(2) Accumulators and lioes at 9QOF.

(3) Conductors include all metal tran5ferri.ng heat to the primary coolant systein *ex*cept for the fuel rods.

23 -

f \\

~*-

(.

Table 3.2 Pa 1 i sa_des.. Doub 1 e-Ended Suet ion Leg Break Ref lood Energy Balance (Million Btu)

Reactor Coolant System Inventory Safety Inje~tion Tank Water(l)

Safety Injection Pump Flow(l)

Cor~ Stored Heat ( 2 ~

Decay and Fission Heat Primary Vessel Walls Primary Vessel Internals I

Primary Loop Metal Steam Generator Inventory (I.L. 69.8 -,59.1 = 10_.7)(3)

Steam Generator_ Tube Metal**

(I.L. 7.6 - 6.3 = 1.1)(3)

Approximate Brea~ Flow Energy S.G. Side Pump Side Total 72 ( 106) Btu 81 (106) Btu 153 ( 106) Btu Reference Temperature is 32°F.

Notes:

20 Seconds 22.4 16.2 59.3 12.4 28.4 137*.9 14.3 TOTALS 290.9 Decrease

-3.1 77.6 13.5 11.0 12.3 3o3.

3.5 1.cr 33.7 120 Seconds 25.5

....5.2.

56.0 8.9 27.4 104.2 l0.7

-237.9

{l).°The S.I. water temperature was 3QQOF to prevent numerical instabilties.

Actual value should be llOoF.

(2) Based on ANS + 20% decay heat curve.

(3) Energy from intact loop steam generator.

- 24

Table 3.3 Heat Structures in Palisades Containment Model Structure Area ft2 Thickness ft

1. Tanks and piping (.453 inch) 19,332

.0378

2.

Ducts (.10 inch) 20,072

.0083

3.

Reacto~ crane (2.35 inch) 6,973

• 1958

4.

Internal concrete (33 inch) 9,401

2. 75*.

5.

• Gratings and trusses 20,996

.0144 6

• Containment dome 7,270 3:0217

7.. containment dome base 11,000 7.75

8. Containment side wall 50,600 3.5217

9. Storage pool floor and shielded walls 4,456 4.35

10. Containment base slab 8,229 12.44

11. Biological shield wall 2,340 7.8672

12. Structur.a 1 support steel 26,320

.45 Deck received had 2.25, which was in error.

25 -

"'.*'.< *.. -~-.-*---* ---- -.*.,

-~~.. -..., :**-. ':'"--*1..:*.-~~...:...,~

' \\..

e.

Table 3.4 Palisades Containment - Initial Conditions Outside air temperature Outside air pressure Relative humidity of outside air Volume of primary capable of holding liquid Temperature of primary system vapor region Temperature of primary system liquid region Volume of containment Volume of liquid pool in co~tainment sump Temperature of containment vapor region Temperature of coniainment liqui"d region, Pressure in containment Relative humidity in containment Horizontal cross-sectional area of containment

.. 9S°F 14.7psia Oo60 -

3 3050.3 ft.

2S0°F 2S0°F.*

1.6E6 ft 3

~

10 ft3 -

120°F-*--

o. --*

120 F - - -

14.7-psia 1.0 8,229~ft~-

.. ~**.

. *-* *--....... -~--=--.... *-.

e k

TABLE 4. l

/

Main Steam Line Break

\\

Mass/Energy Release Data Ruetured Steam Generator Time (Sec)

Lbm/hr Btu/Lbm o.o 3.266E07 1200.7

o. 1 3.266E07 1200~8 0.2 3.186E07 1201.2 0.3 3.106E07 1201.5 0.4 3.037E07 1202~0 o.s 2.957E07 1202.3 0.7 2.826E07 1W2.8 1.0 2.637E07 1203.5 1.3 2.501E07 1204.0 1.5

• 2.409E07 1204.3 1.8 2.283E07 1204.5 2.0 2.215E07 1204.6 2.5 2.067E07 1204.8.

\\.

3.0 l.942E07 1204.8 4.0

1. 759E07

• 1204.5 5.0 h621E07 1204.1 7.2

l. 370E07 1203.2 8.0 1.336E07 1202.9 10.0 1.233E07 1202.3

. 15.0 l.062E07 1200. 9 20.0 9.592E07 1199. 9 30.0 8.221E06 1197.6 40.0

• 7. 193E06 1196.0 45~0 6.852E06 1195. l so.a 6.566E06 1194. l 54.0 6.280E06 1193.5 60.0 5.938E06 1191. 1 68.0.

5.481E06 1191. 1 68.0 0.0 --- -.w:-'";"

... - ---*-::--=*:*--:-..

~":-.::**~-~-,.,.

..... *-- ---~.. '..

,[

e TABLE 4. l (cont'd)

Main Steam Line Break

\\,

Mass/Energy Release Data Isolated Steam Generator Time

{Sec~

Lbm/hr Btu/Lbm OoO 1.656E07 1200.4

o. 1 l.656E07 1200.4 0.2 l.656E07 1200~6 0.6 l.587E07 1201.3 1.0 1.530E07 1201.8 1.3 l.496E07 1202.2 1.5 1:473E07 1202.5 2.0 1.416E07 1202.9 3.0.

1.279E07 1203.6 3.5 1.233E07 1203.8 4.0 1.153E07 1204.0 s.o 9.930E06 1204.2 5.4

9. 135E06 1204.2 6.0 7.650E06 1204.3

' \\

6.4 6.622E06

- 1204.2 6.8 '

5.}09E06 1204.2 7.2 4.339E06 1204. 1 7.8 1.484E06 1203.9

8. 1 o.o 1203.8

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PAL.ISl\\DES ULOHDOWN, RUN 30 12 16.*

TIME (SEC) *

. f1gur~ 3.1 *.

BREAK FLOW, STEAM GENERATOR~ SIDE OF BREAK

• e I.

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325, P~LISAD[!I COIHAlttH[Nl HSLD.. llHHOUT orrsn1;., POUER, NO J0111R. 24~Dom 02f03181 F1gure 4. 7T

~...

CONTAINMENT TEMPERATURE.;: ATMOSPHERE & POOL, CASE.*~

A Canta i nment Atmosphere B

Canta i nment Sump-.

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APPENDIX s:* LICENSEE COMMENTS

'AND STAFF RESPONSES This appendi~.co.ntai!ls. the c.o.mmen~s received from ~he,Consumers_ Power Company

~

• cg~_ference Bl) regarding Appendix A, and the responses to those* comments" The

• I ".;,! :..:

licensee's comments are presented in the Attachment to Appendix B.

The re-(.......

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~

• ':""..;.... _*,t*

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...... ~,,;.

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sponses were prepared by the Containment*systems Bra.nch*and its contracti:>r 9

• the

'Lawrence Livermore Nat fona l Labor.atory **.

~ -.

~

-.* -* ;_.. :~ ~ -~::..:

• .. *-.41!>

• 7* *.-

The following _present~_.a bri~f summary of th~ containment analysis performedw a Z:.~-.:. :.. ~ -.: :.::..:..... :... _ ~ :*,. --: ~.-*.:*2 ; :*.-

.-..::*. ;...:~-.:..:~: -

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~~ ~-*..:.. ;....;~: :-:

-.~*!.=

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~-

-.*.:. *: "':'. *.- =- !". r-c:

~ '

The containment analysis in Reference 82 presented both primary system pipe

,....:.. -- : : --~; - ;:'; ~:

i5Y.-eak*s and* secondary system pipe breaks. The primary system pipe break *_was a do~b)e~~nded cold.:,leg* suctiOn' break~. )"he_ ~econdary pip~ -bre~k-- analysi~- i~ciuded six diffe.rent calculations based-on different ii'liti~l a_~sur:nptions'.. Cases l _.

through 5 were 102% power MSLBs which showed that a MSIV failure resulting in the blowdown of both steam generators would exceed design conditions by a sub-stantial amount.

Case 6 was a 102% power MSLB analysis based on the mass and en-ergy release rate data provided in Reference 83.

In this case a MSIV failure was not assumed.

The major conclusions presented were the following:

.lo The resulting contain~ent pressure and temperature from a double-ended cold leg suction brea~ are less than des~gn donditions.

2.

The resu~ting containment pressure from the.blowdown of both steam genera-tors due to a MSIV failure exceeds design by a substantial margin.

3.

If a fix* were imposed to prevent the blowdown of both steam generators>>

then the resulting containment pressure and temperature would be less than ENCLOSURE

2- -.

~

.the de.sign.

This analysis was ba.sed on the 102% power case presented

i l'l Reference*.. 83.* * *

.... *: : *-~

4.

.*.;. ~

=~::c*,,..:

.~

The most severe pressure and temperature conditions resultfog from a

pipe break*ins_ide tne:contacinmen.t would* be from a**MSLB.:,~

.)..,,,.*"::

Respo.nse to. Comments

.Q9J11me11.t_ l.is.d_i rec.ted_ at_::t~-- LOC/:\\: _an~lys i_s.. : t T_!le contai.nm~nt-analysis.. for::~

LOCA.~as. _perf~ed: a~. 102_% of: ~~a-feguard?: -rati_ng or 2690.76 !'1\\.1.E ~s. s~te( on-~

page ::5_., Se_ctj_on- -~.*l: *.

• The 2200 MWt power level mentioned in Section 2.0 is in

.. error and should be corrected*to read 2530 MWt.

Comment? 2 throu-gh 5 are_ d_irected -tow~.:.rd*-the-MSLB analysis cases,l thro!Jgh* -S __

The..-i-naccuracies. ment-ioned :in ~hese.comments* are a result of a techrii~al speci-fication.change occurring after the analysis was performed.

However, they do not change.the conclusion drawn from the analysis. This was that a MSIV fail-

~re would result in the blowdown Of both steam generators and the resulting containment pressure would exceed design.

Comments 6. and 7 are directed at the MSLB case 6 *. Case 6 was the most 1 imiting pipe break analysis with a fix imposed on the blowdown of both steam generate-rs resulting in a peak pressure 1.2 psi below design.

The mass and energy re-lease rates were provi9ed by Consumers Power Company in Reference 83. This analysis assumed 102% power and offsite power was available.

However, the single failure assumption for the containment response calculation assumed loss of off-site power and failure of one diesel generator. The available heat removal sys..;

terns and spray.initiation times were taken *from Reference 84.

Since the result_ing

-J.-

4.

• e.

containment.. pressure*is*-only.1.2'psi below de$ign,_.*any c_hange iri the heat

~.

removal capability or spray initiation times could have an impact on the con-clusions*~..... :o ___,c,*;.. ~.-~~_.:::::*:_;.~_: _ _.;;_ *-

.*.?.

.. :::*:'~.'

.:~ :.*.. -.,._..-

~-

-~*--:.,..-1_-:,;;:

• _~-... : *'

Camme~t.~_ i-S:. di recte_d.. at ttie heat* siri~s*:used *~in "the containment*:*analysi_S:-.-_-~ The c....:*-=:: ;:>:. *.*

.. ::::..:(.i -. :_:

.:.":.:.:...~

....;-...; *~~.. -~-;* -=~:-::

-~ -.-.:

-....... -:"' *** i-heat sinks used were* provided -by th*e Ticensee along with the CONTEMPT input

':.:;*::- ~: *1* :~ '._.:.2.:..

~_.. :;.-..-_;(..:

• deck for the Palisades plant.

The differences between the heat sink data used

_: ~ --:...

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and tha(p~~v-id~d-_*i*~: -Ff.gur~ 1° ~o( Refererl"ce Bl< appe~rs :to *be* enough *to *have some

~~ :.~-~;:.~..:;.:..::.~~a::.:~ ~-.:..:;:..?-.-;_v.. ::

~:l::i~ ~i~~al..

impact on the* -containm_ent response calculations._ T~e heat sink data used in the 2~:..::-

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analysH~ar~~mc)rEr cons~!/iE~iv_~}:h_~_n;,1~-i11'rov!~ed, in~ Refer~nee~1: :::__::;'*

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concl.~sion - -----

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Based on the abov_e,.discussion~ *comme~ts.*1 through 5-do not alter'--the conclu-.

..:~.... *- -->.. **

sions* r~ached in the Palisades report. ::.liowevet, comments 6 through 8 pofnt to some discrepancies in the assumptions used in the containment analysis which could have an impact.

In addition, the information provided by Consumers Power Company in Reference 83 regarding the mass and energy release rate data for MSLBs is not adequate to determine if all power levels and single failure assumptions were done in accordance with the SRP.

However, Consumers Power Company does.state that power levels from 0%*to 102% were consi~ered, and that the 1.02% power case was the most limiting.

In summary, we do not plan to reanalyze the accidents even though results cou~d be more favorable.

Our conclusion on the need for design modif?cations to accom-modate the single failure of a MSIV has not changed.

• ---:--*-:--:--*:~.--~- _., ***_:::_. ~ * * -*- ~:~ -.:.:~~,--.~,::*::=:~~~-***,-* ~

• • • .-~~r ""~-~;**;:;:--~*::****-:-'"~-~~~:.~*_.TI:~;-~:;.~;:::;*;_~~~~~~:.:*.:*;_:~:~::~*~*~*:~-~-:-.-,,*:::~--~--~~~~~~:.* ~;:~.~~~.~::::-:-::~~';:?"°1;:7~;:;~.::::~~*:::~:f

~

References-:*. :._'.:-

• Bl. Letter -~~?!ll ___ R.~- _yi_~_~en~_ of Co~~~:_:.s _ P~_wer_ ~om~_any to.D. Crutchfield of

-:*~**-- -----.---~-~--*-

NRC, July*9,_*19a1~

Subject:

SEP Topics VI-2.Dand VI;,.3~

82; Letter from:~NRC to Consumers Powe.r Company, June 17, 1981,

Subject:

• • ~-:-**:--*---***--~.*---*-~* -'--*-*-*

~. ;......

Dr~ft-E~-al*u-atio'n-o~f--sf:p Topics-*vr.;;2-~*o* arid VI-3* for the Palisades-Plant.

--~;.

~ *-.

83.

Letter from R. Hu.ston of.ConsumersP'ower-Company to--o~*-z;-emann"i)f NRC,

~*.: -. -.,

January:21.",::.J980,L~~-Au.t9~~1ic_ Initiation of AuX'iliary Feedwater System at

---

~--~

-~-

._j ':

Palisades Plant.

11

.:...:_~: _*_ -

~

84.

Letter from D. Hoffma~ -.of Consumers Power Company tC? D. Crutchfield of "NRC,-* Nov~~b:e?.24;~-~l980~,~~-iPalisades Plant -- -Proposedleclfnical--specifica-

.: tion:.Changes Related to Containmen:t _Spray Initiation Times."

'/

µ

~

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2.

• a PALISADES PLABT 4

A.HMENT ro APPENQir.. e Page *1 of 4 :

CONSUMERS POWER COMPAHY COMME?rl'S 011 STAFF EVALUATION SEP 'l'OPICS VI-2.D AliD VI-3

~--.. --:~

. : :*-. ~ ~.:::..

PBP:e 2. Section 2.0 - Pal.isades is licensed to operate at-2530 Mw:t Tice 2200 MW't. * ".' '* ** :, 'o Page 17~~ ~e *ma:;.s.. ~d--~~~* release aata ~ ret'erence *1 ot. the report *

• does not correctly account.ror f'eedwater addition to the*--mtact steam generator," resultllig in about a 5% det'icit in mass addition. This was

~ported in LER-80-003*~ '; :*..

°'.'

- j;:

• .=. -

._J.** -::... : ;.

Page 17,_ Section 3 - _As a. res.ul. t ot LER...80-003. tbe MSIV' s also close on CEP.. Analysis done *by' MPR -.CMPR-654) determined the val:ve* closure time to be within~l;JJ_ec.*~o~ ~~~pt.o~ trip signal *

. ~ ~.:. 2 c : 4:

=

4. *pa.ge 20 2 Section l

~* Wi~- ottsite power a'V8ilable, the sprq pumps will start on ~.

• * ~

new-is~ establi:shed--in.. 18. 3 sec. :rollcndng _cte.._

This includes 4 seconds tor the pump to come up to speed mid 14.3 seconds to fill the *spray *headerse ~Bote that less *than full :t'lov will initiate before 'this _tµie*.. ~~ v~te:r~ !".'* -1rpray-trcm the rings 'as they~*.

-~ f; !_*: ~ : 2 *

~ -

- G

.: -::

• 5o Page 21., Case 5 ~ With.. 1o*ss"'o:t' of:t'site power mid no diesel/generator

6.

7.

'8.

.: failure, full_ conta.:illlnent-* SP.ray nov--begins at 36 seconds n~t. 84 seconds.

This is a*.n!;!u;i. t of'. cont&.imDent spray pump resequencing and pre:filling.

the spray headers as' :reported in u:a-80-003.

Page 23, Case 6 - The case vi th l contaimnent spray pump. and 3 air coolers is more* limiting than 2 pumps and l air cooler.

This is due to the rela-tive inefficiency of the coolers in the superheated ~tmosphere and the longer time for spray initiation.

Page 23 7 Case 6 = With-2 containment spray pumps, full spray is established in 45 seconds. Partial flow begins in 31 seconds. With one pump, full now is established in 56 seconds vi th partial flow in 3l seconds~

Page 27. Table 3.3 - For LER-80-003, a more representative model of the containment heat sinks was made.

These are listed in t'igure 1.

The valves were developed by Bechtel in support of Exxon LOCA analysis "CXN-75-64A).

For contaimnent response, the LOCA conservatisms were removed.

I~ -

{. -*

.e Page 2 of 4

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