ML19317G969

Forwards Request for Info Required to Complete Review of Util ECCS Reanalysis

ML19317G969
Person / Time
Site:	Rancho Seco
Issue date:	06/18/1975
From:	Purple R Office of Nuclear Reactor Regulation
To:	Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
750613, NUDOCS 8004020629
Download: ML19317G969 (21)

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Docket V

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NRC PDR

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Local PDR ORB-1 Reading Docket No.'50-312 J'n 131975 KRGoller g

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Sacramento Municipal Utility District SSheppard ATTN:

Mr. E. K. Davis JRBuchanan General !!anager TBAbernathy 6201 S Street, P. O. Box 15S30 t

o Sacramento, California 95813 ACRS (14)

Gentlemen:

DEisenhut

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The enclosure to this letter identifies infornation that we require in order for us to completo our review of the ECCS reanalysis you g ;;-

will subnit pursuant to our Order for Modification of License dated Decenber 27, 1974. So:ne of the information has previously been

- :EE identified (e.g., potential boron precipitation for PlfRs, single

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failure analysis), but has been repeated here for completeness.

Sincerely, 9

Robert A. Purple, Chief Operating Reacttis Branch #1 Division of Reactor Licensing

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Enclosure:

Required Information (ECCS) cc w/ enc 1:

3 See next page

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Post ' Office Box 15S30 EE Sacramento, California 95S13

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RI', QUIRED INFORMATION 1.

Break Spectrum and Partial Loop Operation,_

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

2.

Potential Boron Precipitation (PWR's Only)

The ECCS syctem in each plant shou'ld be evaluated by the applicant (or licensec) to show that significant changes in chemical concentrations' will not occur during the long term af ter a loss-of-coclant accident (LOCA) and these potential changes have been specifically addressed by appropriate operating, procedures. Accordingly, the applicant should 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 should consider all aspects of the specific in plant design. including component qualification in the LOCA environment addition to a detailed revicu of operating procedures.

The applicant should examine the vulnerability of the specific plant design to single failures that would result in any significant boron precipitation.

3.

Single Feilure Analysis A singic failure evaluation of the ECCS should be provided by the applicant (or licensee) for his specific plant design, as required by Appendix K to 10 CFR 50, Section 1.D.l.

In performing this evaluation, the effects of a single failure or operator error that causes any manually controlled, electrically-operated valve to move to a position that could adversely affect the ECCS must be considered.

Therefore, if this consid-eration has not been specifically reported in the past, the applicants upcoming submittal must address this consideration.

Include a list of all of the ECCS valves that are currently required by the plant Technical Specifications to have power disconnected, and any proposed plant modifications and changes to the Technical Specifications that might be required in order to protect against any loss of safety function caused by this type of failure.

A copy of Branch Technical Pocition EICSB 1S f rom the U.S. Nuclear Regulatory Commission's Standard Review Plan is attached to provide you with guidance.

The single failure cva)ortion should include the potentini for passive failures of fluid syctems during long term cooling following a LOCA as well as single failures of active components.

For PWR plants, the single failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.

4.

Submerged Valves 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 submerged, not only those in th:a safety inject ion Valve: in other systems may be needed to limit boric acid con-syste centr.. tion in the react or vessel during long term cooling or may be required for containment isolation.

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The applicant (or licensee) is to provide the following information, for j

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cach plant:

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(1)

Whether or not any valve motors will be submerged following a LOCA in i

1 the plant being reviewed.

(2)

If any valve motors will be flooded in their plant, the applicant (or l-

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(a)

Identify the valves that will be submerged.

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(b) Evaluate the potential consequences of flooding of the valves TZ ~

for both the short term and long term ECCS functions and containment isolation. The long term should consider the

,,J potential problem of excessive concentrations of boric acid in.

PWR's.

(c) Propose a interim solution while necessary modifications are qsg being designed and implemented.

(currently operating plants

-a:

only).

(d) Propose design changes to solve the potential flooding problem.

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5.

Containnent Pressure (PWR's Only)

The containnent pressure used to evaluate the performance capability of the ECCS shall be calculated in accordance with the provisions of Branch Technical Position CSB 6-1, which is enclosed.

6.

Low ECCS.Reflood Rate (Westinghouse NSSS 0nly)

Plants that have a Westinghouse nuclear steam supply shall perform their ECCS analyses utilizing the proper version of the evaluation model, as defined below:

(1)

The December 25, 1974 version of the Westinghouse evaluation

model, i.e., the version without the nodifications described in UCAP-6471 is acceptable for previously analyzed plants for'which the peak clad temperature turnaround was identified prior to the reflood rate decreasing below 1.1 inches per second or for which the reflood rate was identified to remain above 1.0 inch per second; conditions for which the December 25, 1974 and March 15, 1975 versions would be equivalent.

(2) The March 15, 1975 version of the Westinghouse evaluation m'odel is an acceptable codel to be used for all previously analyzed plante, for which the peak clad temperature turnaround was identi-fied to occur after the reflood rate decreased below 1.1 inches per eccond, and fo.r which steam. cooling conditions (reflood rate less than 1 inch per second) exist prior to the time of-peak clad

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temperature turnaround.

The March 15, 1975 version will be used

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for all future plant analyses.

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HINDRS! REQUIRDfENTS FOR ECCS BREAK SPECTRUM SUBMITTALS I.

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 reactor designs only and must be re-assessed when new reactor concepts are submitted.

The current ECCS Acceptance Criteria requires that ECCS cooling performance

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be calculated in accordance with en acceptable evaluation model and for a nu-ber of postulated loss-of-coolait accidents of different sizes, locations and other properties cufficient to provide assurance that the entire spectrum cf postulated loss-of-coolant accidents is covered.

In addition, the calculation is to be conducted wi'.h at least three values of a discharge coefficient (Cp) applied to the r.ostulated break area, these values spanning the range from 0.6 to 1.0.

Sections IIA and IIIA define the acceptable break spectrum for most operating plants which have received Safety Orders.

Sections IIB and IIIB define the break spectrum requirements for most CP and OL case work (exceptions noted later). Sections IIC and IIIC provide an outline of the minimum requirements for an acceptable ccaplete break spectrum.

Such a complete break spectr'um

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could be appropriately referenced by some plants.

Sections IIID and IIIE previde the exceptions to certain plant types noted above.

A plant, due to reload a portion of its core will have previously submitted all or part of a break spectrum analysis (either by reference or by specific calculations).

If it is the intention of the Licensee to replace expended fuel with new fuel of the same design (no ccchanical design differences which could affect thereal and hydraulic performance), and if the Licensec intends to operate the reloaded core in compliance with previously approved Technical Specifications, no additional calculations are required.

If the reload core design has changed, the Licensee shall adopt either of Sections IIA or IIC, or of Sections IIIA or IIIC of this document, as appropriate to the plant type (BUR or PWR). The criterion for establishing uhether paragraph A or C shall be satisfied will be determined on the basis of whether the Licensee can demonstrate that the shape of the PCT versus break size curve has not bc:n modified as a consequence of changes to the reload core design. When the reload is supplied by a source other than the NSSS supplicr, the break spectrum analyses specified by Sections IIC or IIIC shall be submitted as a cininum (as appropriate to the plant type, BWR or PWR).

Additional sensitivity studies raay be required to assens the sensitivity of fuel changes in such areas as singic failures and reactor coolant pump performance.

II.

TR".SSl'RI7.ED WATER REACTORS A.

Operating Reactor Reanalyses (Plants for which Safety Orderc were issued)

If calculational changes

• were made to the LBM** to make it wholly in

• CalculatJonal changes /Model changes--those revisions made to calculational techniques or fixed parameters used fur the referenced completc spectrum.

• • LBM--Large break Madel; SUM--Small Break bdel

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conformance with 1GCFR50', Appendix K, the following minimum number of break

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Each sensitivity study performed during the nir.cs should he reanalyzed.

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development of the ECCS evaluation model shall be individually verified as 4

remaining.applicabic, or shall be repeated.

A plant may reference a break

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and core design.

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Reanalyze the limiting break.

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b'. Rcanalyze two smaller breaks.1,n the large break region.

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2.

If the largest break size doet not result in the highest PCT:

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Reanalyze the limiting b.reak.

a.

Reanalyze a break larger and a break smaller than the limiting

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If the limiting break is outside the range of Moody b.

i:.fli less than 0.6), then the limiting break.

multipliers of 0.6 to 1.0 (i.e.,

gj break plus two larger breaks must be analyzed.

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small break conformance with 10CFR50, Aopendix K, the analysis of the worst (SBM) as previously determined from paragraph C below should be repeated.

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

If a new plant is of the came general design as the plant used as a basis for a referenced complete spectrum analysis, but operating 1.

parameters have changed which would increase PCT or m

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change in PCT have been made to the ECCS mod

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'M plus a minimum of three small breaks (SBM), one of which is th 2

transition break.*

analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS cvaluation model.

(configuration and core design) is applicable to all 2.

If a new plant is the same with respect to the generic

. generic studies because it plant design and paraucters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 20 F change in PCT were made to the ECCS model'used for

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then no neu spectrum analyses are

'the referenced conplete spectrum,The new plant.may instead reference the applica required.

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'* Transition Break (TB)--that break size which is analyr.ed with both the LBM and SBM.

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Minimum Requirements for a Complete Break Spectrum g.

Since it is expecte'd that applicants will prefer to reference an applicabic

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complete break spectrum previously conducted on another plant, this EEds -

EE paragraph defines the mininum number of breaks required for an a'eceptable complete break. spectrum analysis, assuming the cold leg pump discharge is 12.

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established as the worst break location.

The worst single failure and vorst-case reactor coolant pump status (running or tripped) shall be These studies g3g established, utilizing appropriate sensitivity studies.

should show that the worst singic failure has been justified as a function of break size.

Each sensitivity study published during the development

+E of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated. Also, a proposal for partial loop operation shall be supported by identifying and analyzing the vorst break

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size and location (i.e., -idle loop versus operating loop).

In addition, sufficient justification shall be provided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by the partial loop configuration.

Unless this information is provided, plant Technical Specifications shall not permit operation with one or more idle reactor coolant pumps.

It must be demonstrated that the containment design used for the break It spectrum analysis is appropriate for the specific plant analyzed.

should be noted that this analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.

1.

LBM--Cold Leg-Reactor Coolant Pump Discharge Three guillotine type breaks spanning at least the range of a.

Hoody multipliers between 0.6 and 1.0.

b.

One split type break equivalent in size to twice the pipe cross-sectional area.

c.

Two intermediate split type breaks.

d.

The large-breas/small-break transition split.

2.

LBM--Cold Leg-Reactor Coolant Pump Suction Analyze the largest break size from part 1 above.

If the analyses in l

part 1 above should indicate that the worst cold leg break is an intermediate break size, then the largest break in the pump suction should be analyzed with an explanation of why the same trend would not apply.

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LBM--hot Leg Piping

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Analyze the largest rupture in the hot leg piping.

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SBM--Splits Analyze five different small b'reak.aizes. One of these breaks must include the transition split break. The CFT line break must be analyzed for B&W plants.

This break may also be one of the five

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small breaks.

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III.

BOILING WATER REACTORS

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The generic model developed by General Electric for BWRs proposed that split

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and guillotine type breaks are equivalent,in determining blowdown phenomena.

The staff concluded this was acceptable and that the break area may be considered at the vessel nozzle with a zero loss coefficient using a two phase critical flow model.

Changes in the break area are equivalent to changes in the Moody multiplier.

The minimum number of breaks required f'or a complete break spectrum analysis, assuming a suction side recirculation line break is the design basis accident (D3A) and the worst singic failure has been established utilizing appropriate Also, a proposal for sensitivity studies, are shown in paragraph C below.

592 partial loop operation sbn11 be supported by identifying and analyzing the worst In addition, *

-29 break size and location (i.e., idle loop versus operating loop).

suf ficient justification shall be provided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by the partial Unless this information is provided, plant Technical loop configuration.

Specifications shall not permit operation with one or more idle reactor coolant pumps.

BWR2, BWR3, and BWR4 Reanalysis (Plants for which Safety Orders were issued)

A.

if the referenced lead plant analysis is in accordance with Section III, paragraph C below, the following minimum number of break sizes should be It is to be noted that the lead plant analysis is to bc reanalyzed.

performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.

A plant may reference a break g;e-spectruc analysis conducted en another plant if it is the same confinuration and core design.

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Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicabic, or shall be repeated.

1.

If the larcest break results in the' highest PCT:

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Reanalyze the limiting break with the appropriate referenced a.

single failure.

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b.

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

Meanalyze the transition break with the singic failure and model c.

that predicts the highest PCT.

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2.

If the largest break does not result in the highest PCT:

~ Reanalyz.e the limiting break, the largest break, and a smaller break.

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7f calculational changes have been made to the SBM to make it wholly in conformance with 10CFR50, Appendix K, reanalyze -the small break (SBM) in 3s; accordance with Section IIIC.

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New CP and OL Case Work

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"E 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

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=m basis for the lead plant analysis, but operating parameters have changed which would increase PCT or metal-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 0

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.

analysis should be justified as remaining applica* ale, in the lead plant including the sensitivity studies used as a basis for the ECCS evaluation model.

(configuration or core design) is applicable to all 2.

If a new plant to the generic generic studies because it is the same with respect plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 20 F change in PCT were made to the ECCS model used for the then no new spectrum analyses are required.

referenced complete spectrum, The new plant may instead reference the applicable analysis.

g Minimum Requirements for a Complete Break Spectrum C.

This paragraph defines the minimum number of breaks required for an This complete spectrum analysis is acceptable complete spectrum analysis.

required for each of the lead plants of a given class (BWR2, BWR3, BWR4, Each sensitivity study published during the development BWR5, and BWR6).

of the ECCS evaluation model shall be individually justified as remaining applicabic, or shall be repeated.

Four recirculation line breaks at the worst location (pump suction or discharge), using the LBM, covering the range from the transition D coefficients of from 0.6 to 1.0 break (TB) to the DBA, including C times the DBA.

Five recirculation line breaks, us'ing the SBM, covering the range 2.

from the smallest line break to the TB.

e The following break locations assuming the worst singic failure:

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largest steamline break a.

b.

largest feedwater line break

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largest core spray line break g

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largest. recirculation pump discharge or suction break (opposite

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side of worst. location)

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BWR4 with " Modified" ECCS

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Same as Section IIIC.

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BWRS 5:::

Same as Secticn IIIC.

F.

EUR6 Same as Section IIIC.

IV.

LOCA PARAMETERS OF INTEREST bhb A.

On cach plant and for cach break analyzed, the following parameters (versus time unless otherwise noted) should be provided ca engineering

...,;;g; graph paper of a quality to facilitate calculations.

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--Peak clad temperature (ruptured and unruptured node)

--Reactor vessel pressure

--Vessel and downcomer water level (PWR only)

--Water level fnside the shroud (BWR or.ly)

--Thermal power Mj

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--Containment pressure (PUR only) s.;

B.

For the worst break analy cd, the following additional r'rameters (versus time unicss otherwisc noted) should be provided i engineering graph paper of, quality to facilitate calculations.

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

--Flooding rate (PWR only)

--Core flow (inlet and outlet)

-Core inlet enthalpy (BWR only)

--IIca t transfer coefficients

--MAPLHCR versus Exposure (BWR only)

--Reactor coolant temperature (PWR only)

--Mass released to containment (PWR only)

. --Energy, released to containment (PWR only)-

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--PCT versus Exposure (BWR only)

--Containment condensing heat transfer coefficient (PWR only)

--Hot spot flow (PWR only)

--Quality (hottest assembly) (PWR only)'

--Hot pin internal pressure

--Hot pot pel'let average temperature

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

A tabulation of peak clad temperature and metal-water reaction (local C.

and core-wide) shall be provided across the break spectrum.

Safety Analysis Reports (SARs) filed with the NRC shall identify on

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D.

the run date, version number, and version date of the computer each plot Should differences exist in model utilized for the LOCA analysis.

version nu=ber or version date from the most current code listings made available to the NRC staff, then cach modification shall be identified with an assersment of impact upon PCT an'd metal-water reaction (local and core-wide).

A tabulation of times at which significant eveats occur shall be E.

The following provided on each plant and for each break analyzed.

events shall be included as a minimum:

--End-of-bypass (PWR only)

--Beginning of core recovery (PWR only)

--Time of rupture

--Jet pumps uncovered (BWR cnly)

--MCPR (BWR only)

--Time of rated spray (BWR only)

--Can quench (BWR only)

--End-of-blowd own

--Planc of intercot uncovery (BWR only) 6 M

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BABCOCK AND WILCOX -

CATEGGRY I:

177 FA w/ Lowered Loops Arrangement Re-analysis (Safety Order Plants):

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1, 2, 3

-- IIA \\

These plants must resubmit at EE Ocyggg l

1 ast 3 breaks.

(They will do

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break spectrum reanalysis sub-

f 2535.

Arkansas Power 1

-- IIA mitted generically by B&W.)

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Rancho Seco

-- IIA 2772 New Ots:

Three Mile Island 2 --IIB (2) 3 Since these plants are the same 2772 design as the above plant, they Crystal River 3

--IIB (2) may reference the same reanalysis

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2452 of the complete spectrum above.

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Midland 1, 2

--IIB (2)f

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New cps:

None CATEGORY II:

177 FA w/ Raised loop Arrangement New Ols:

Davis Besse 1

--IIB Complete spectrum required.

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New cps Davis Besse 2, 3

--IIB Complete spectrum required.

CATEGORY III:

205-FA Plants New OLs:

None I

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APPLICATI.0:10F THE 51!iOLE FAILURE CRITERIO!i 10 PAliUALLY-CO:11 ROLLED p

ELECTRICALLY-OPERATED YALVES A.

_B3Cr.CR0tmD Where a single failure in an cicctrical system Can result in loss of Capability to perfort.

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This is necessary regard-p a safety function, the effect on plant safety.must be evaluated.

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' less of whether the loss of safety function is caused by a co ponent failing to perform a requisite mechanical motion, or by a component perforning an undesirable p.cchtnical r'otion.

C This position establishes the acceptability of discennecting po..cr to electrical components of a fluid system as one means of designing against a single failure that might cause an un-desirable co ponent action. These provisions are based on the assurption thet the cov onent is then equivalent to a similar coepenent that is not designed for electrical operatien, e of the valve.

e.g., a valve that can be oper.ed or closed only by direct canual operattor They are also based on the assu ption that no single failure can both restore poder,to the electrical syste, and cause rechanical motion cf the corponents served by the electrical Eth The validity of these assur.ptions should be verified when Ppplying this pcsition.

system.

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BRUiCH TEC!nICAL POS1710'i Failures in both the " fall to function" scnse and the " undesirable functic,' sense of l.

cocponents in electrical systeis of valves and other fluid systed coctr.er.ts 5%id

'oe considered in designing against a single failure, even though the valve or ot'.er fluid systen co.ponent eay not be called upcn te fActic.n in a given safety creratien61 sequence.

Where it is deterrined that failure of an electrical systen cc.conent ca*. cause

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undesired re'chanical cotton of a valve or other fluid system cc penent and t'eis motion results in loss cf the syster. Safety function, it is acceptabic, in lieu of design changes that also rcay t'c acceptable, to disconnect pcwer to the electric systers The plant technical specifications should' of the valve or other fluid sysica componcnt.

include a list of all' electrically-operated valves, and 'the r N utred positions of these

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valves, to which the requirencnt for removal of electric power is applied in order to satisfy the single failure criterion.

Electrically operated valves that are classified as " active" valves i.e., are required 3.

to open or close in various safety syster everational sequences, but are ranually-h Such valves n.ay not be controlled. Should be o? crated fren the r.ain crotrol rocc.

..,, l included anon,J those valves from which,0..cr is rer.oved in order to ncet the single failure criterion unless: (i) clectrical power can be restored to the valves from the main control room.(b) valve operation is not necessary for at 1 cast tm minutes following occurrence of the event requiring such operation, and(c) it is demonstrated

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that there is reasonable ass rance that all necessary operator actions ull) be per-i / is a=;s;.

forced within the tire shown to be adequate by the analysis. The plant technical mE specifications should include a list of the required positions of manually controlled.

electrically-operated valves and should identify those valves to which the require.

rent for rc9 oval of electric power is applied in order to satisfy the single failure criterion.'

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k' hen the single failure criterion is satisfied by removal. of electrical power from

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valves described in(2) and (3), above, these valves should have redundant position

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indication in the r:ain control" room ar.d the position indication system should.itself.

reet the single failure criterion.

5.

The phrase " electrically-operated valves" includes both valves operated directly by an

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electrical device (e.g., a rotor-operated valve or a solenoid-operated valve) and those valves operated indirectly by an electrical device (e.g., an air-operated valve whose

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air supply is controlled by an electrical solenoid valve).

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REFERE!;CES 1.

Pemorardum to R. C. DeYoung and V. A. Moore from V. Stello. October 1,1973.

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MINIMLN C0fiTA!!.*4 Erit PRESSURE H0 DEL FOR PL'R ECCS PERFOR'%';CE EVALUATIO!1 A.

BACKCROU :D Paragraph I.D.2 of Appendix K to 10 CFR Part 50 (P.cf.1) ' requires that the containment ggg

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j pressure used to. evaluate the perfont.ance capability of a pressurized water reactor (PWR)

E emergency core cooling system (ECCS) not exceed a pressure calculated conservatively for It further requires that the calculation inclu'de the effects of operation of that purpose.

all installed pressure-reducing systems and processes. Therefore, the following br.anch technical position has been developed to provide guidance in the performance of minfrun containment pressure analysis. The approach described beloa applies only to the ECCS-related containr.cnt pressure evaluation and not to the containment f'unctional capability i

evaluation for postulated design basis accide.'.ts.

B.

BRA!lCH TECH':! CAL POSIT!03 g;

1.

Input Infor stien for Model E

Initial Contain~?nt Internal Conditions The minirum contain ent gas tem;crature, ninimur contain ent pressure, and naximut, humidity that rc.ay be encountered under limiting normal op: rating conditions shnuld t'c used.

b.

Init tal Outside_ Contain ent A-Men _t,,C_oeditions A reasenably los arbient temperature external to the contain ent shculd be uscd.

c.

Contain-'ent Yelu-c.

g_,.

The maximun net free co".tainr:ent volume should be wed. This raxi.un frec voluee should be deterrined frc, the gross containment volu c nirus the volu?cs of interrial structures 5;ch as v.3115 and floors, structural stect, ra'cr ecuir. ent, g

and piping. The individ,a1 volute calculations shculd reflect the uncertainty in the component volu.cs.

2,

&tiveHeatSinks_

a.

Spejy and Fan Coolins Systems The operation cf all engineered safety feature contain?. nt heat removal systems operating at maximum heat removal capacityt i.e., with all containment spray trains operating at maximum flow conditions and all c.crgency fan cooler units operating, should be assuccd. In addition, the minimum tenperature of the stored water for the spray cooling syste t and the cooling water supplied to the fan i'

coolers, based on technical spccification limits should be assumed.

6.2.1.5-3 s

+**===w.mo

. eem o e.e-em me a e e.

• c8 0. 0 4-0 2 0

=.

=

.~..

(

===u Deviations from the foregoing will be accepted if it can be shown that the worst conditions regarding a singic active failure. '.tored water torperature, and cooling water te:nperature have been selected from the standpoint of the overall ECCS model.

==

b.

Containment $ team Mixinj With Spilled FCCS Water The spillage of subcooled ECCS water into the containment provides an additional heat sink as the subcooled ECCS water mixes with the steam in the containment.

. gg The effect of the steam-water nixing should be considered in the contain.9ent 4EE pressure calculations.

g E=

c.

Containment Stean Mixing With Water fro.9 Ice Melt The water resulting from ice celting in an ice condenser contairc.cnt provides an additional heat sink as the sub, cooled water mixes with the steam while draining i

from the ice condenser into the lower containment volume. The effect of.the steam-eiater mixing should be considered in the containment pressure calculations.

=

3.

Rissive Heat Sinks a.

Identification The passive heat sinks that should be included in the contain ent evaluation model should be established by identifying those structures and components within the containecnt that could influence the precsure response. The kinds.of struc-tures and conponents that should be included are listed in Table 1.

Data on passive heat sinks have been compiled from previo,5 revie.s and have j

been use' as a basis for the sirplified nodel outlined tielos. This rodel is d

acceptable for minimum contain:cnt pressure analyses for ccnstruction pernit applications, and until such tir.e (i.e., at the operating license review) that a complete identification of available heat sir.%s can be ecde. This sirplified approach has also been folle,.ed for o; crating plants by liccr. secs cc plying with Section 50.46 (a)(2) of 10 CFR Part 53. For such cases, and for conste.:ction permit reviews, v.$.ere a detailed listing of Feat sinks within the contain ent of ten cannot be provided, the following procedure ray be used to codel the passive heat sinks within the containment:

(1) t!se the surface area and thickness of the primary containm nt steel shcIl or steel liner and associated anchors and Concreto, as apprceriate.

(2). Estinate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assu c an average thickness of 3/S inch.

(3) Model the internal concrete structures as a slab with a thickness of I foot and exposed surface of 160,C33 f t'.

The heat sink thereophysical properties that would be acceptable are shown in Table 2.

'N 6.2.1.5-4 j

C

, C'

• • E'.7

~ ~ ' " "

e T.l'.

~

L

...=.m,.

Ei u

At the operating Ilcense stage, arpilcants should provide a detailed If st of G;;

passive heat sinks, with appropriate dimensions and properties.-

b; Heat Transfer Coefficients The following conservative e ondensing heat transfer coefficients far heat transfer

. : ces i

to the exposed passive heat sinks during the blowdown and post-blowdown' phases of USE rn the loss-of-coolant accident should be used (See Figure 2):

=

(1) During the blowdown phase, assume a linear increase in the condensing heat E

2

/

transfer coefficient from hinitial=8 Btu /hr-ft

,.F. at t = 0, to a peak "value four times greater than the,maxinum calculated condensing heat trans-

.+ -

fer coefficient at.the end of blowdown, using the Tagami correlation (Ref.2)*,

0.62 h

= 72.5 1 eax Vt

~p 2

where h,, = naxinum heat transfer coefficient, Btu /hr-ft,.7 Q

= primary coolant energy. Stu

,6E 3

V

= net free contain:ent volune, ft t,

= time interval to end of blowdown, sec.

[

l (2) During the long-term post blowdown phase of the accident, characterized by low turbulence in the contaircent ateosphere, asstr e, condensing heat transfe'r coefficients 1.2 times greater than those predicted by the Uchida data

=.

(Ref. 3) and given in Tabic 3.

f (3) During the transitten phase of the accident, between the end o' blowdo n and the long-tern post-bloedcwn phase, a reasonably centervative es.:onsntial transition in the condensing heat transfer coefficient should be assu ed (See Figure 2).

.:;.g The calculated con:'ensing heat transfer cetfficients based es. the above reth'd

• s should be applied to all expcsed passive heat sinks. both retal and ccecrete. and for both. painted and unpainted surfaces.

Heat transfer between adjoining raterials in passive heat sinks should 'ce based on the assumption of no resistance to heat flow at the r:sterial interfaces. An example of this is the contain.ent liner to concrete' interface.

C.

REfERE!:C,ES, 1.

10 CFR 150.45. " Acceptance Criteria for Emergency Core Coel'ng Systems for Light k' ster

_=

fiuclear power Reactors, and 10 CFR Part 50. Appendix K. "'l'S Evaluation Models."

2.

T. Tagami, " Interim Report on Safety Assesscents and Facilitics Establish-ent Project in Japan for Period Ending June 1955 (No.1)." prepared for the fiational React ~ Testing Station, February 28, 1955 (unpublished wo'rk).

6.2.1.5-5 "1

L.-_.

~

1 3

=3

3.

H. Uchida. A. Oyama, and Y. Toga. " Evaluation of Post-Incident Cooling Systems of Light-Water Power P.eactors." Proc. Third International Conference on the Peaceful Uses of 9

Atomic Energy. Volvac 13. Session 3.9. United Nations. Geneva (1964).

........ Ii[

= = =

91..'Es t?=4::.::.:

.....i $$

= = =

=

- ;;g

=,:

=..

J.iEi--

iN hii2i-i i

\\.

e ni ir 15==

t

4.4 o

2E

=--

lbkh k;[;

6. 2.1. 5-6

l

=

...ir

.: v..

(.

TABLE 1

..^

O IDENTIFICAT10'l OF C0:1TAINMENT HEAT Sif KS

=_=

Containment Building (e.g., liner plate and external concrete walls, floor, and su:np. and 1.

lineranchors).

==

Containment Internal Structures (e.g., internal separation walls and floors, refueling 2.

Es pool and fuel transfer pit walls, and shielding walls).

Supports (e.g., reactor vessel, steen generator.~ pumps, tanks.. major components, pipe

,,.2:;

3.

~

supports, and storage racks).

Uninsulated Systens and Components (e.g., cold water s; % heating, ventilation, and 4.

air conditioning systems, pumps, motors, fan coolers, re.

'r.ers, and tani.s).

.==

=?:

Miscellaneous Equipnent (e.g., ladders, grathgs, ciectrical cable trays, and cranes).

5.

=

e 5

u..

a.

E:.9

3

j

}

l 1

l 6.2.1.5-7

.N

U tie 2 t...

,....... -..,...r.

HCAT SINK THERM 3 PHYSICAL' PROPERTIES-g,}. ""~ '

Specific 1hemal Densigy Heat Conductivity '

Material Ib/ft Stu/lb *F_

Stu/hr-ft_'F_

"e-E

,:.;;i

=

Concrete 145

~0.156 0.92 g;

..gr..;_

' = " ~

5 teel. *

• 490 0.12 27.0

=.

2 E

.f5 Tlll *. '

L:

E.1 i

4 e

t

i I"]:i

-.- J 6.2.1.5-8 f

4.

e

.a n: :.., sfr : :-

- :::::r; at:-

..l

..m...

.~.Z,..-

-,,J...

.. ~. -.......

e

.<.,w r

se me

3

-~

glg g -

UCH1DA HFAT TRAfl5FER COErr!CIE!1TS Mass' Heat Transfer Mass Heat Transfer

" ' ~

=... _.

Ratio,

Coefficignt Ratio Coefficigqt

[1b air /lb steam)

D tu/hr-ft 'F)-

(1b air /lb steay1 $ tu/hr-ft 'r1

.:2......_,

=..

.3_'.

['.~" R" 50 2

3 29 20 8-2.3 37 18' 9

1.E 46

~ l 1.3 63 14 10 g,.;.. ;,

10 14, 0.8 98

=5 7.

17 0.5 140 Er 5

21 0.1 280 4

24 av k

-:id

.'?.'**.~.'

r se.

e e

9 e

e e

e.

=

6.2.1.5-9 o

9 8

- ao e e.meaum o eegeme e em +

9-* *em e

' WMese ese e emme e.e ee e e m.gumamenem We Eume *'UN'S

1. 6-s g.... _. _..

.; 3.;

,.3.-

~,

t t

~.,

i l-Figure 2 Condensing Ilcat Transfer Coefficients for Static Heat Sinks 11; f.

3

- oe

.c e4

.oa w

w 8

h

= 4. x h Tagami

.m o

max 8

N linear o

7 I

.025(t-t )

stag) e P

c h=h

+ (h

-h m

I L

2-I stag max a

t-*

I o

c

')

l

=

I h,,g - 1.2 x hUchida l

I co I

c h =8.

g o

.j i

t Time p

1 I

blowdown i reflood I

I i

g 1

O e

9 1

.: 3

..--- ? ?7 F*

j; 3. ;

.. -.=<

.s g.p

.o

,..+;,; i + ;,.;i

,