ML20212M778

Topical Rept Evaluation of WCAP-10325, Westinghouse LOCA Mass & Energy Release Model for Containment Design - Mar 1979 Version. Rept Acceptable for Ref in Licensing Actions

ML20212M778
Person / Time
Issue date:	02/17/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19292G766	List: ML20211C456 ML20212M778
References
NUDOCS 8703120182
Download: ML20212M778 (10)

Text

ENCLOSURE SAFETY EVALUATION REPORT WCAP-10325 (Proprietary) 1.0 Introduction Westinghouse Topical Report WCAP-10325 (Reference 1), entitled " Westingh LOCA Mass and Energy Release Model for Containment Design - March 1979 is a revision of Westinghouse Topical Report WCAP-8264(Pi/WCAP-8312(NP)

(Reference 2) which was previously found acceptable by the staff in a letter dated March 12, 1975 (Reference 3), and which has been used by Westinahouse over the years. Westinghouse has stated that the purpose of the revision is to make the loss of coolant accident (LOCA) mass and energy release model consistent with its Appendix K ECCS computer codes and to incorporate improv The staff's review focused analytical models based on more recent test data.

on these changes.

7.0 Model Review and Evaluation The new'model includes such modifications as a two phase pump model, a drift flux model, a momentum flux model, a core heat release model, a wall heat transfer model, a steam-water mixing model during reflood, and a decay Most of the modifications were previously used by heat generation model.

Westinghouse for the ECCS Evaluation Podel (February 1978 version), presen in WCAD-830?, WCAP-8170, WCAP-9220, and WCAP-9?Pl.

blowdown, refill, The LOCA analysis is divided into four phases, namely:

reflood, and post-reflood.

The revisions affect all phases but the refill 8703120182 670217 PDR TOPRP EMVWEST C PDR

f 7 phase. Blowdown computations are now based on the computer code SATAN VI, rather than the SATAN V version. The reflood phase is now based on a new version of the WREFLOOD computer code. The changes to the SATAN and WREFLOOD codes were made to more realistically describe the physical phenomena, and have been found acceptable for the purpose of ECCS analysis. The present review of these modifications is with a view towards their applicability to containment analysis. The focus of containment analysis is the pressure and temperature response of containment, while the focus of ECCS analysis is the In addition thermal hydraulic response of the reactor core and primary loop.

to the above changes, the decay heat model and saturated steam release model at the steam generator outlet (for a pump suction LOCA) were modified in v! CAP-10325 for containment analysis considerations, not for ECCS evaluation.

t 2.1 Blowdown

' Table 1 of WCAP-10325 presents a comparison of the significant differences between SATAN V and SATAN VI. SATAN VI includes a drift flux model, a momentum flux model, a two phase pump model, a core heat release model, a nucleate boiling wall heat transfer correlation, and a thin metal heat release in model. These chanoes were found acceptable for ECCS evaluation by the sta##

f a letter from D. Vassallo (NRC) to C. Eicheldinger (Westinghouse), dated Fay 30, 1975 (Re#erence 4).

I The drift flux model was incorporated into the SATAN VI code to more f

i l

accurately calculate the water inventory in the vessel at the end of l

i l

N v blowdown. This model has no impact on the resultant mass and energy release rates since the reactor vessel lower plenum is conservatively assumed to be full at the end of blowdown for the purpose of containment analysis.

'The SATAN VI code includes a momentum flux term in the conservation of momentum equation. The complete momentum equation should have this momentum flux term. Including this term results in an initial reduction of break flow, slower depressurization and higher fluid enthalpy at the break. The results

-of a Vestinghouse sensitivity study show that including the monentum flux term increases the peak containment pressure.

The reactor coolant punp (RCPI model in SATAN VI is based on two-phase punp test data, whereas generalized pump characteristics are used in SATAf' V. The new model accounts for reverse two-phase flow through the pump and an equivalent density head. The resulting break flow, therefore, is higher on the steam generator side of a cold leg break and lower on the punp suction side of The the break early in the transient, when comparing SATAN VI with SATAN V.

trend reversec late in the blowdown transient so that the overall effect diminishes. Furthermore, system depressurization is somewhat slower, and the integrated break flow is slightly lower (by approximately 0.2%). The overall impact of including the two phase pump model is a slight increase in the predicted containment pressure.

s

. - - - . . - - - . . , , ,-- , , . . , ,- ..,. ,.- - - ,-.,..~ ,-

4 a-A refined core thermal rodel is incorporated into the SATAN VI code to more accurately model the behavior of the hottest fuel rod and assembly in the core during a LOCA. The model does not affect the average core behavior, which is what the blowdown mass and energy release rates are based on. Further-more, the calculational procedure for the stored energy in the fuel is automated in SATAN VI, whereas SATAM V required input data from a separate computer code (LOCTA-RP). These changes do not affect the blowdown mass and energy release rates, and, therefore, the containment response.

The Thom correlation is used in SATAN VI, in lieu of the i ens-Lottes l i correlation (SATAN V), to calculate the film heat transfer coefficient during nucleate boiling. Both correlations are emperical, but the Thom correlation bas shown better agreement with experimental data. The results of a sensitivity study show a negligible difference in the peak containment pressure with the two heat transfer correlations.

2.? Reflood The reflood phase is modeled by the WREFLOOD computer code, which was modified in the new model . The most important modification, which has been found acceptable for ECCS evaluation, is the steam / water mixing model.

Additional options added include (1) a transient core heat release model and I?) a secondary to primary heat transfer modification.

In the new version of the WREFLOOD code, complete steam / water mixing is assumed to occur in the intact loops to account for the interaction of loop s

, , - - . , - - . - . , - - . - - . - - - - - ~ - - -

' steam with ECCS injection water. Westinghouse justified the assumption by a series of 1/3 scale tests, which included a range of flow parameters corresponding to all phases of mixing conditions during the reflood phase.

The results of these tests showed that complete condensation occurred immediately, and that thermal equilibrium was obtained at a very short distance downstream of the injection nozzle. The steam condensation part of the nixing process has a more pronounced effect on the containment pressure and temperature response than the other changes discussed. Based on the test results presented ,

in WCAP-10325 and the staff's previous acceptance of this phenomenon for ECCS evaluation, the steam / water mixing model is viewed as a more realistic character-ization. Therefore, the staff finds it acceptable for blowdown calculations in containment analysis.

A transient core heat release model was developed and incorporated into the new model for long term mass and energy release calculations. In the previous model, a constant estimated value for the average core heat release rate was used; an itterative calculation was then used to assure an appropriate mass and energy balance. In the new model, decay heat is released into the fluid below the quench front as it moves up the core. A comparison of energy release was made; the new model calculated higher core heat release early in the transient and lower heat release later. A Westinghouse sensitivity study showed a negligible net effect on peak containment pressure.

A change in the secondary to primary heat transfer calculation was also made in the model. In the old model, the fluid leaving the steam generator during reflood was assured to be superheated to the temperature of the

6 secondary side. Westinghouse found this assumption was not consistent with some of the FLECHT experimental data. Moreover, a Westinghouse sensitivity study showed that assuming saturated steam at the er.it of the steam generator resulted in a slightly higher peak containment pressure. The reason for the higher pressure is that non-superheated steam has a higher density which results in a higher mass release rate. Even though the specific enthalpy is lower, the total energy release rate, which is the product of mass release rate and specific enthalpy, is higher. Based on the results of the Westinghouse sensitivity study, the staff concludes that the assumption of saturated steam at the exist o# the steam generator is acceptable.

2.3 Post-reflood Two modifications have been made in modeling the post-reflood phase, namely: (1) steam conditions are assumed to be at saturation at the outlet of the steam generator and (2) containment depressurization and equilibration with the primary coolant system are directly included in the calculation. .

The first modification has been discussed in the previous section. The second modification accounts for thermal equilibration of the primary coolant system with the containment and the continued energy release from the prinary coolant system as the containment depressurizes. The change, in essence, automates the calculational procedure. The staff finds this change acceptable.

e 7_

2.4 Decay Heat Vodel.

A new decay heat model, based on the American Nuclear Society (ANS)

Standard ANS-5.1, November 1979 revision, is used for containment analysis.

The new model results in reduced decay heat generation until about 50,000 seconds following a LOCA. This same decay heat model was incorporated into the McGuire FSAR for containment analysis. The staff's review of the McGuire FSAR concluded that for application to containment analysis, the model was acceptable. Since the new model is basically the same as the one that has been reviewed and found acceptable by the staff, no additional review is needed.

3.0 Staff Position The staff has reviewed the methodologies and assumptions described in WCAP-10325. The review focused on these aspects of the calculation that differ from WCAP-8264(P)/WCAP-8312(NP). The staff has concluded that the methodologies are acceptable for the calculation of mass and energy releases to the containment following a loss-of-coolant accident. Therefore, WCAP-10325 is an acceptable topical report for referencino when calculating containment design parameters, a.0. References

1. WCAP-10325, Letter from T. M. Anderson (Westinghouse) to J. Stolz (NRC),

April 25, 1979,

Subject:

Westinghouse LOCA Mass and Energy Release Vodel for Containment Design - March 1979 version.

~

Y li -

~s .g.

?. WCAP-8264(P)/WCAP-8312(NP), Topical Report - Westinghouse Mass and Energy Release Data for containment Design, August, 1975.

3. ' Letter from D. B. Vassallo (NRC) to C. Eiche1dinger (Westinghousel, dated Ma'rch 17, 1975.

3 s

4. Letter from D. B. Vassallo (NRC) to.C. Eiche1dinger (Westinghouse 1, dated 1

May 30, 1975.-

I i

h 4

i l

l T

+

1' i

l l

t- -

m . .,

William J. Johnson Should our criteria or regulations change such that our' conclusions as to the acceptability of the report are invalidated, Westinghouse and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report with revision of their respective documentation.

Sincerely, Orisica13;&nedby, Charles E. Rossi, Assistant Director Division of PWR Licensing . A

Enclosure:

As stated  ;

Distribution:

(w/o enclosure) (w/ enclosure)

PDR LHulman CBerlinger RBosnak DCS CERossi RBallard WPinners BClayton CMcCracken EDoolittle TNovak REmch JWShapaker BBuckley AD R/F JDosa CLi PTam ECestral]FileE RDiggs JCalvo P0'Connor PSB Read File JFunches EShomaker

'(

y

• SEE PREVIOUS CONCURRENCE g /

OFC :P5B:PWR-A* :PSB:PWR-A* :PSB:PWR-A* :0G '// :P :AD:PWR-A  : :

_____; ___________:____________: ___________:_ ,r_ _____: ________:7g_ _ _:___________

l NAME :Cli: art :JWShapaker :CMcCracken :E homaker  :.Funches .TERO  :

l DATE :1/21 /87 :I/21 /87 :1/21 /87 b 6 /87 :1/d/87 :1/Ag/87  :

i

( OFFICIAL RECORD COPY I

l i

t

. 3 .- v:,, :

N

)' Mr.~E. P. Rahe Jr., Manager' TNuclear Safety-Department Westinghouse Electric' Corporation x Box 335 -

Pittsburgh, Pennsylvania 15230"

Subject:

Acceptance for Referencing of Licensing Topical Report

~

WCAP-10325, " Westinghouse LOCA Mass and Energy Release

'Model for Containment Design - March 1979 Version" The staff- has completed its review of the subject . topical report, _ submi ed by Westinghouse Electric Corporation letters' MS-TMA 2075 ( April 25,197 nd NS-EPR-2948 (October 4, 1984). We find the report acceptable for _ erencing in: licensing actions to the extent specified and under the limita) ons delineated in the report and the associated NRC evaluation whic)(is enclosed.

.The evaluation defines the basis- for acceptance-of the report /

We do~not intend to repeat our review of the matters desc bed in the' report and found acceptable when the report is referenced in 1 ensing actions except

to. assure that the material. presented is applicable t the specific plant

involved. Our acceptance applies only to the matter described in the report.

InaccordancewithproceduresestablishedinNURJG-0390,itisrequestedthat-Westinghouse publish' accepted versions.of thisfreport, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions should incorporate this letter and he appropriate evaluation between the. title page and the abstract. The acc ted versions shall include an -A

'(designating. accepted) following the rep rt identification symbol.

-Should our criteria or regulations c nge such that our conclusions as to the -

-acceptability of the report are in idated, Westinghouse and/or the applicants referencing the topical report wi be expected to revise and resubmit their respective documentation, or sub ': justification for the continued effective applicability of the topical r ort with revision of their respective documentation.

Sincerely, Charles E. Rossi, Assistant Director Division of PWR Licensing - A

Enclosure:

Distribution:

l As stated (w/o enclosure) (with enclosure)

PDR LHulman CBerlinger RBosnak

!" DCS CERossi RBallard WMinners BClayton CMcCracken EDoolittle TNovak

/ REmch JWShapaker BBuckley PWR-A R/F JDosa CLi PTam Central File l- RDiggs JCalvo P0'Connor PSB Read File

~

L OFC- :PSB:PWR-A :ACtPS . WR-A:AD:PWR-A

__.__;____________::] I p ...:____________:____________:_.._________:._________.

a er...b_c_hracken :CRossi  :  :  :

NAME :CLi:arte :M

/87 :1/ : __ _h_/ / 87 : 1/ /87 :

iDATE :1/ 1/ /87 :1/

. OFFICIAL RECORD COPY