ML20116A164

Nonproprietary Upper Plenum Injection LOCA Program Plan Using SECY-83-472 Methodology

ML20116A164
Person / Time
Site:	Point Beach
Issue date:	03/31/1985
From:	NORTHERN STATES POWER CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP., WISCONSIN ELECTRIC POWER CO.
To:
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ML19269B430	List: ML20116A136 ML20116A148 ML20116A164
References
PROC-850331-01, NUDOCS 8504240385
Download: ML20116A164 (46)

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.{ . WESTINGHdUSEPROPRIETARYCLASS.3 UPI LOCA PROGRAM PLAN USING SECY 83-472 METHODOLOGY Prepared by Northern State Power Wisconsin Electric Power Westinghouse Electric Corporation l

l March 1985 i

0504240385ggj4d$66 l PDR AM PDR I P

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• WESTINGHOUSE PROPRIETARY CLASS 3 UPI LOCA PROGRAM PLAN USING SECY 83-472 METHODOLOGY CONTENTS

(

l Section Title PAge 1 Summary 1 2 Introduction / Background 2 3 UPI LOCA Plan 4 3.1 Introduction c, E 4

3.2 [ ] Code Description and Validation Program 5 3.3 Program Objective 14 3.4 Detailed Work Plan 15 4 Discussion of NRC Concerns 29 5 Program Plan Schedule, and Proposed Meeting 37 with NRC Staff 6 Conclusion 39

'l 7 References 40 c,E.

AppendixA-E [) Nomenclature 42 i

, WESTINGHOUSE PROPRIETARY CLASS 3 t

1.0

SUMMARY

This plan was prepared to address the NRC's concerns regarding the capability of the existing licensing evaluation model to accurately predict the performance of the upper plenum injection (UPI) emergency core cooling system for a postulated Loss of Coolant Accident (LOCA). The objective of'this plan

• will be to [

_. a.

_. This plan will cover cores with [

. ]'The thermal-hydraulic calculational tool which will be used in this program is the [ ] dde. This program plan also includes a new method of assessing the LOCA margin for two-loop PWRs with UPI using the-statistical methodology as described in SECY-83-472.

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, WESTINGHOUSE PROPRIETARY CLASS 3

?

2.0 INTRODUCTION

/8ACKGROUND

, After the core cooling hearings of 1972 to 1973,II} Westinghouse developed a series of computer codes which predicted the performance of the emergency core cooling systems for those PWRs with cold leg pumped safety injection and cold leg accumulators. These computer codes and models were subsequently reviewed and approved for use on all Westinghouse PWRs with cold leg injection. It was felt that the two-loop plants with upper plenum injection (UPI) would be covered by the approved evaluation models at this time.

The ECCS evaluation model for two-loop plants equipped with UPI originally '

assumed that water injected into the upper plenum did not interact with steam and entrained water risi5g from the core and the UPI water was assumed to be delivered directly to the lower plenum through the cold leg injection -

l locations. This approach was defended on the basis that a potential negative UPI impact of increased steam binding would be balanced by the positive UPI

. impact of improved core heat transfer.

In November 1977 the NRC issued a report (Reference 2) describing a model they developed based on recent experimental data. Their report concluded that the then-current Westinghouse approach [ .

]QBy calculating increased steam generation due to top injection without giving credit in the hot channel for improved core heat transfer, the NRC model necessarily produced lower flooding rates and higher peak cladding temperatures than the then-current Westinghouse model (all other factors being equal).

This NRC model was then used to provide bases for continued operation of the two-loop plants, resulting in penalties in peak clad temperature to be applied to results of the old model (References 3 and 4).

4 e

f 2

1

, WESTINGHOUSE PROPRIETARY CLASS 3 The NRC reviewed the Westinghouse revised model and found that it was acceptable as an interin basis for continued operation; however, they did not regard the model as a final accepted model and required that the two-loop utilities continue to work to produce a final acceptable ECCS evaluation model or make a plant hardware change to eliminate UPI.

In addition to the review of the Westinghouse UPI evaluation model, the NRC has also sponsored UPI ECCS experiments in the Japanese cylindrical core test facility (CCTF) to examine UPI phenomena, as well as better estimate PWR calculations using the TRAC code to examine UPI effects and to contrast the PWR response with UPI to a plant with cold leg injection. The tests and TRAC calculations did show thermal-hydraulic phenomena which were not accounted for explicitly in the Westinghouse evaluation model. However, there was no conclusion that the UPI emergency core cooling system was inadequate; rather, the NRC felt that the evaluation model which was used to estimate its perfomance had an unacceptable level of uncertainty and that this evaluation model should be improved. In the February 1985 letters to the two-loop i

utilities (5) the NRC identified ten areas where improved modeling was felt necessary.

A program plan has been developed, using the[' ]c,E. computer code, to address each of these above areas. Thisplanwillalso[ .

...... . [willresultinanacceptablemodelwhichmeetstheAppendixK requirements.

l

. 3

P WESTINGHOUSE PROPRIETARY CLASS 3 3.0 UPI LOCA MARGIN PLAN

3.1 INTRODUCTION

This program will utilize the Westinghouse [

]"whichisspecificallydesignedtobeE * *

}

, . . NresentedinSECY-83-472.

The SECY-83-472 document states that there are three areas of conservatism in the current licensing models: the required Appendix K conservatism, the conservatism imposed by the NRC staff to cover uncertainties, and conservatism

! imposed by the industry in many cases to reduce the cost for the analysis.

Based on a review of the available experimental data and the best-estimate computer code calculations, the NRC staff has concluded that there is more than sufficient safety margin to assure adequate performance of the emergency core cooling system and that this excess margin can be reduced without an l ,

adverse effect on plant safety. Therefore, in the SECY-83-472 approach it is proposed that the licensee would utilize a realistic model of the PWR to calculate the plant response to a LOCA at the most realistic or most probable level (50% probability) and at a more conserv'ative 95% probability level. The calculation at the 95% probability would include uncertainties such as power level, fuel initial temperature, nuclear parameters, and computer code

uncertainties. The realistic PWR model and the uncertainty analysis can be performed on a generic PWR model which covers a class of similar plants, that 1s, two , three , or four-loop PWRs such that all plants of that class are l

bounded by the generic plant model. The third calculation the licensee would have to perform would be to use a plant-specific realistic best-estimate model including the required Appendix K features, such as 1971 ANS decay heat + 20%

maximum stored energy, no rewet during blowdown, etc. This would be a new Appendix K calculation. The NRC staff will accept this new Appendix K calculation provided that the peak cladding temperature (PCT) calculated with the Appendix K calculation is greater than the PCT c91culated at the 95 percent probability level but below the licensing limit of 2200*F.

4

ifESTINGHdySE PROPRIETARY CLASS 3 1

The NRC staff interpretation of these results is that the required features of Apperdix K have sufficient margin to cover all uncertainties inherent in a LOCA analysis combined at a 95% probability level. Such a series of calculations would provide an acceptable licensing basis by conforming to the Appendix K requirements and is estimated to result in 500*F peak clad temperature margin which can be used for operational flexibility, low leakage .

loading patterns to address pressurized thermal shock concerns, upratings, and more economical fuel designs.

3.2 E JbDEDESCRIPTIONANDVALIDATIONPROGRAM c

TheC, 3 c,Eomputer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to

l small-and large-break loss-of-coolant accidents and other anticipated transients. It is derived from the merging of COBRA-TF for the reactor vessel and TRAC-PD2 for the loop components.

The COBRA-TF computer code provides a two-fluid, three-field representation of two-phase flow. Each field is treated in three dimensions and is compressible. Continuous vapor, continuous liquid and entrained liquid drop are the three fields. The conservation equations for each of the three fields l

l and for heat transfer from and within the solid structures in contact with the fluid are solved using a semi-implicit, finite-difference numerical technique on an Eulerian mesh. The COBRA-TF vessel model features extremely flexible noding for both the hydrodynamic mesh and the heat transfer solution. This flexibility provides the capability to model the wide variety of geometries encountered in vertical components of nuclear reactor primary systems. The flexible noding scheme allows representation of single hot asserd/ lies with their unique hardware and power characteristics while permitting coarser noding for the remainer of the core. This reduces the run time of the calculations.

TRAC-PD2 is a systems code designed to model the behavior of the entire reactor primary system. It features special models for each component in the l

l l

l 5

WESTINGHOUSE PROPRIETARY CLASS 3 system. These include accumulators, pumps, valves, pipes, pressurizers, steam generators, and the reactor vessel. With the exception of the reactor vessel, the thermal-hydraulic response of these components to transients is treated with a five-equation drift flux representation of two-phase flow. The vessel

component of TRAC-PD2 is somewhat restricted in the geometries that, can be .

modeled and cannot treat the entrainment of liquid drops from the continuous liquid phase directly, and thus, is replaced with the COBRA-TF vessel model.

The TRAC-PD2 vessel module has been removed and COBRA-TF has been implemented as the new vessel component in TRAC-PD2. The resulting code is [ 3c,E The vessel component in[ ]C.E.has the extended capabilities provided by the three-field representation of two-phase flow and the flexible noding. The code has been assessed against a variety of two-phase flow data from ,

experiments simulating important phenomena anticipated during postulated accidents and transients. O

~

Multiphase flows consisting of two or more fluids are separated by moving phase interfaces. The liquid and vapor phase properties are assumed to change discontinuous 1y across these interfaces. In general, the phases present can be any combination of liquid, solid, or gas. The flow pattern can take any one of a wide variety of forms, such as bubbly flow, droplet flow, gas-particle flow, and stratified flow. Exact conservation equations can be derived for each phase, and jump conditions relating variables on each side of the phase interface can be formulated, much as in single-phase shock wave theory. With appropriate initial and boundary conditions, these equations could theoretically be analytically solved for the exact motion of each phase and the phase interfaces. Except in a few simple cases, however, an exact analysis of multiphase flow is impossible because of its complex and essentially chaotic nature. The quantities of interest are the average behavior of each phase in the control volumes. Thus, most work in multiphase flow is done using average equations across the control volume.

C The average conservation equations used in the vessel module of [ ]a,E.re derived following the methods of Ishii . The average used is a simple Eulerian time average over a time interval, At, assumed to be long enough to 6

? WESTINGHOUSE PROPRIETARY CLASS 3 smooth out the random fluctuations present in a multiphase flow but short enough to preserve any gross unsteadiness in the flow. The resulting average equations can be cast in either the mixture form or the two-fluid form.

Because of its greater physical appeal and broadsr range of application, the two-fluidapproachisusedasthefoundationfor[ 3CEi The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on another i are accounted for by interfacial interaction terms appearing in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.

CE Thethree-fieldformulationusedinthevesselmoduleof( 3isa straightforward extension of the two-fluid model. The fields included are vapor, continuous liquid, and entrained liquid. Dividing the liquid phase into two fields is the most convenient and physically reasonable way of

.. handling flows where the liquid can appear in both film and droplet form. In I such flows the motion of the droplets can be quite different from the motion of the film, so a single set of average liquid phase equations cannot adequately describe the liquid flow or the interaction between liquid and vapor.

Use of a third field, droplet flow, permits more accurate modeling of ,

thermal-hydraulic phenomena such as entrainment, deentrainment, fallback, liquid pooling and flooding. These effects are significant when modeling the upper plenum injection system performance.

c ThephasicconservationequationsfortheE ] v,E.

essel model describe the time-averaged behavior of a phase, which can be any phase in a multiphase flow. The phasic conservation equations are general within the assumptions listed below.

l 1. Gravity is the only body force.

7

j , WESTINGHOUSE PROPRIETARY CLASS 3

2. There is no volumetric heat generation in the fluid.

3. Radiation heat transfer is limited to rod to drop and rod to steam.

4. The pressure is the same in all phases.

5. The viscous dissipation can be neglected in the enthalpy formulation for l the energy equation.

Four mass conservation equations are required for the vapor, continuous

liquid, entrained liquid, and noncondensible gas mixture. In vectcr form they -

l are, respectively, as follows:

. -s -+

E "v v + Y * ("vPU P

v v) = r"' + V

• Gf

, - * ,, 4 g a gp, + V - (a gp,Ug ) = -r, - S"' + V

• Gf a a

. g a,pg + 7 - (a,p,U,) = -r,' + S"'

3 gopgg + V * (a p Uy ) = r g + V G,T l The individual terms in the equations become l

( Rate of + Rate of mass = Rate of mass + Rate of mass + Rate of mass change gain by gain by gain by efflux due to of mass convection interfacial entrainment void drift transfer or chemical -

reaction 8

\

, WESTINGHOUSE PROPRIETARY CLASS 3 Also, a gas mixture transport equation (for each species of noncondensible gas) is solved explicitly at the end of each times step. The nomenclature used for these equations is given in Appendix A.

l l

Two energy conservation equations, in which the liquid and the entrained .

liquid are assumed to interact at a rate sufficient to nearly maintain i equilibrium, are specified for the vapor-gas mixture and the combined liquid h (a yp ygh yg) + 7 * (a yp yghyg D y) = I"' h, + q$y + QQ - V " (ay g fg)

I i *

• a "

iY (ag + a,)p gg h + 7 * (a gggg p h U ) + V * (a,p tg h U,) = r"'h + f q$g + Q

- V * (a gT)

The individual terms in the equations become Time rate + Convection = Energy transport + Interfec'.a1 + Wall heat - Turbulent of change due to phase heat flux heat flux change transfer The use of a single energy equation for the combined continuous liquid and liquid droplet fields implies that both fields are at the same temperature.

Ir nr, ions where both liquid droplets and liquid films are present, this e

assumption can be justified in view of the large rate of mass transfer between the two fields, tending to draw both to the same temperature.

Three momentum equations are solved in COBRA-TF, allowing the liquid and entrained liquid fields to flew with different velocities relative to the vapor phase. They are as follows:

• ** .+ .* ,,,.*,,, .,

77 (ayp ygU y) + 7 * (a yp ygUU y y) = a yVP + ay p g t, - ty tg + (r"'U) av av

+Va (a y [g) 77 (aggg p U u ) * ~ *t 9P + agg p U ) + 7 * (agggt pg t,g + ty -(r U) g - (S"'U) tv

+7* (og f) 9

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WESTINGHOUSE PROPRIETARY CLASS 3 L h(=,,,0,)+ v- (=,,,D,0,) =

., TP + e,,,g -I + Ty ev l - (r 0) + (s, 0) .

3*

The individual tems in the equations become l

Rate of + ltate of = Pressure + Gravity + Wall l change of momentum gradient force shear momentum change by .

convection

+ Interfacial + Interfacial + Interfacial + Momentum + Momentum drag between drag between momentum exchange exchange vapor and vapor and exchange due to due to continuous drops entrainment turbulence liquid COBRA-TF was developed for use with either rectangular Cartesian or subchannel coordinates. This allows a fully three-dimensional treatment in geometries amenable to description in a Cartesian coordinate system. For more complex or irregular geometries, the user may select the subchannel formulation (which neglects some of the momentum flux terns in the above equations) or a mixture l

of the two. The subchannel approach has been used by Wheeler et al., for bundle thernal-hydraulic analysis by the COBRA series of Codes (10) ,g comparison of the subchannel and Cartesian momentum equations is given in Section 2.3 of the Thurgood reportI I.

One of the important features of the COBRA-TF vessel model is that the governing equations form a complete set. No terms are omitted particularly in the momentum equations where wall shear, momentum exchange due to turbulence and all the interfacial drop terms are represented. The COBRA-TF vessel model 10 __

WESTINGHO)SE PROPRIETARY CLASS 3 I also has two energy equations such that thermodynamic non-equilibrium between the two-phases can be accounted for. [ .

} a, ,

c,E The currently planned program for [ ..] code validation will [, ,

. . . . 3." System effects tests simulate most of the components of the PWR acting together during the transient and can be used to compare predicted event times and peak clad temperatures, as well as other local data. Single effects tests will provide data for specific scaled components of the PWR and can be used to E ' ] ct which for component models have a particularly important effect on[

3a c, E To examine [ ]blowdownperformance,the[

] @l1 be predicted with [ ].Nhese tests are the best available system experiments to assess code performance during the blowdown

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transient due to the relatively large scale of the facility and its nuclear core. The following quantities will be compared:

- CL The following are the most important phenomena affecting the prediction of peak clad temperature during blowdown:

a.

Separateeffecttestswillbeanalyzedtovalidatethe[

fests will be used

]"The{

11

WESTINGHOUSE PROPRIETARY CLASS .3 to test the prediction of the [ 3* ,

, Predictions will be made of the[ ]Indthesewillbecomparedtodata.

The[ .

t . ) a.will be used to assess the [ " ,

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]a The [. ]otwillbecomparedto{ }o-.

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generatedusinga{

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Substantial validations exist for the TRAC PD-2 break model. Although uncertaintiesexistinthemodel,itisplannedin[

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Multidimensional effects may be important guring blowdown. The multidimensionalcapabilitiesof{ ]'will be used to[

]9- The E

]ct, E.

l To assess the predictive capability of [. ] for{

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fihe[ b, There are integral systems effects tests which model a simulated PWR response for a reflood transient for a cold leg injection transient; or a top-down flooding transient for[ ]a. The keyparameterstobecomparedinclude,[ ,

)* ( , ].1The [ ,]Q-are the largest scale system simulations that can be modeled and will provide

usefulassessmentof[ ]GF 12

1

. WESTINGHOUSE PROPRIETARY CLASS 3 i

n. 1 Inadditiontothe[ ] the [ I

]Y11 be simulated with[ Itkassessthe

'These experiments are low pressure top flooding experiments which can characterize the hot assembly thermal-hydraulic characteristics for a UPI situation. The key parameters to be examined in these experiments [ .

]Dheseexperimentswill help quantify the[ ] ]R Since accumulator ECC condensation effects can influence the delivery and ECC bypass offacts; [ '] ' ill be used to model the[ ,

] These comparisons should help reduce uncertainty in the [ i

]D C,E Itiscurrentlyplannedtoincludethe( ,

validation [ .

3"Inaddition,the[ ]c arisions with the [

Additionalverificationof[ ] with the [

Ja 13

WESTINGHOUSE PROPRIETARY CLASS 3 3.3 PROGRAM OBJECTIVE The objective of this program is to address the NRC concerns on the effectiveness of the upper plenum injection (UPI) emergency core cooling system to mitigate the consequence of a large break Loss of Coolant Accident .

(LOCA). Thisoffortwill[ .

. ,, ]4This plan will cover

[

]CIn addition the proposed plan will [ _

3" S

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1 14

WESTINGHOUSE PROPRIETARY CLASS 3 3.4 DETAILED WORK PLAN Task 1. Set.upa[ .

Thistaskwill[ . . ,.-

[Thetwo-loopplantparameters[

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,. WESTINGHOUSE PROPRIETARY CLASS 3 Task 2.[_..

]CL C-The[

]"The { - .

In b .. .

Q

] Ihere will also be

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Therewillbeanumberof{

]G wo T options currently exist for the i .

t .

osas

].b The[ .

]G fhe end result of these calculations will be to develop a{

].4The work scope includes:

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WESTINGHOjiSE PROPRIETARY CLASS 3

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? tfESTINGHOUSE PROPRIETARY CLASS 3 Task 3. Run[ . ----

This task will use the two-loop [

ny(

fTheresultsofthesecalculationswillestablishthe C'

3a This task includes: a u

M>

18

.o _ . _ _ .

, WESTINGHOUSE PROPRIETARY CLASS 3 Task 4. Perform Plant Sensitivity Studies This task will determine the relative [ ,

The work scope includes:

- .-, a r

19

. WESTINGHOUSE PROPRIETARY CLASS .3 Task 5. Perform Sensitivity Studies of Model Parameters In this task the effect on the calculated peak clad temperature of the [

3 The ,

work scope includes:

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20

. WESTINGHOUSE PROPRIETARY CLASS 3 Task 6. Perform {"  !

]

This task will perform additional [

t'

]"Thiseffortincludes: _

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? WESTINGHOUSE PROPRIETARY CLASS 3 Task 7. Prepare Report on Best-Estimate Code This task will prepare a technical report for transmittal to the NRC on the [ ] N report will specifically describethe{ .

..... _.._ . . .. -]"The work scope includes:

L. -

i .

b) suDmit Inis report to Ine MNL,.

22

.o tfESTINGH6USE PROPRIETARY CLASS 3 Tast 8. Perform Analysis of [ U' Thistaskwillmodifythe[. ,

] O In addition, [ ,

]4This taskwillpermitanalysisof-[

3 4E The specific work scope includes:

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WESTINGHOUSE PROPRIETARY CLASS 3 Task 9.[

. 3A usingthetablesof[ . .

]a,Tothiscalculated[

e The specific work scope incli. des:

24

.- WESTINGHOUSE PROPRIETARY CLASS .3 ,

i Task 10. [

)

i Repeat Task 9 using the [. -

1

]a. Add to the [

i )

Thetotal[

]0.The specific work scope includes:

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-- - - ,.,r-_ - .-+.-----,,r,-- --.,--,---,---.------y--..----.-------3 - , - . - .yv.- , y---- , - - - - -. - - , --. y--4-m.

WESTINGH0USE PROPRIETARY CLASS .3 Task 11. Perform a plant specific { 31 The[.. , ,

~

]1The models and assumptions which Will be added are [ ,

]4khis calculation C

]Otherplant-specific [

']"Thespecificworkscopeincludes:

~ -

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26 y= ---e -- -- - - - - , , ~w , _ __ _, , _ _ , _ _ _ . , , , , __ , , , _ _ _ , _ _ , , ,, _ _ _ _ __ _ _ _ _ _ _ _ __

WESTINGHOUSE PROPRIETARY CLASS 3 Task 12.Performa{ ]&

This task will calculate the [., ,

. 3 The purpose of this [

show that the once burned fuel in a mixed core is less limiting than -

j 1

. . . i JaThehott a.

] The specific work scope includes:

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i

=

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_- . . . . . . - . . _ - ... . . . - _ - . _ _ . - . . . . . - . - = _ . _ - - - . . _ _ _ . - . - . , - _ - _ - _ _ . - , . .

, WESTINGHOUSE PROPRIETARY CLASS 3 Task 13. Prepare a Report Describing the [ i.

]O'b A technical report will be prepared which describes the results of eachofthe[. t

].aThereport

{ - .. .

In addition, all (,

a,E

]The work scope includes:

CL l

l d) Submit this technical report to the NRC.

t l 28 1

WESTINGHOUSE PROPRIETARY CLASS 3 i

4.0 DISCUSSION OF NRC CONCERNS The thermal hydraulic development effort presented in Section 3 will result in E

. 3 utilizing the [*

] hh amount of C

- a 3

In addition, it is felt that the [

]e,Epermit a more accurate calculation of -

the core behavior such that the [

]6NRCconcernsonthecurrentWestinghouseUPImodelidentifiedin the February 1985 letter to the two-loop utilities (5) are addressed below with a brief discussion of how the approach given in Section 3 will address the concern.

1. NRC Concern - Metal Heat Transfer to UPI Water Heat transfer from upper plenum metal to UPI water is calculated by a lumped thermal capacitance model. It is necessary to show that either (a) the coefficients used in the model are correct (b) sensitivity studies show that results are insensitive to the values chosen for the coefficients, or (c) the coefficients selected are conservative for all cases and conditions of interest.

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29

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. WESTINGHOUSE PROPRIETARY CLASS 3 GL, E ,C n

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2. NRC Concern - Unoer Plenum Injection Flow Distribution Westinghouse (W) assumes that safety injection (SI) forces ECC water into the upper plenum and this water covers only part of the core. The percent core coverage is important because the M UPI EM uses different analyses for the covered and uncovered core regions. ,

Acceptance of the partial core coverage concept would require explanation of experimental data from the Semiscale and CCTF test facilities which indicate upper plenum pooling occurs in the test facilities and hence, may occur in UPI plants. It would also require determination and analysis of worst case coverage conditions and consideration of steam-water interactions.

c E ,C.

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, WESTINGHOUSE PROPRIETARY CLASS.3

3. NRC Concern - Decay Heat and Stored Enerov This section of the Westinghouse EM submittal says, in full, that i "The decay heat calculation uses ANS plus 20 percent. The initial stored energy at the beginning of reflood is based on the stored energy at the -

end of blowdown plus a calculation that adds the decay heat generated through the refill period to the core stored energy."

In evaluating the significance of this portion of the proposed model, the staff notes that the docketed EM submittals indicate that the refill period includes 14.45 seconds of pumped UPI before bottom of core recovery (80CREC).

The proposed Westinghouse EM ignores this UPI water during refill, which should be conservative relative to vessel inventory. However, it is necessary -

to address the effect of this 14.45 seconds of pumped UPI on (a) stea a generation and steam flow or steam binding, (b) changes in fuel and metal wall

. stored energy prior to 80CREC, and (c) reflood, this being done with (d) 1 and 2 trains of SI operable. Alternately. (e) it is necessary to show that the proposed treatment is conservative and that injection of UPI water before

BOCREC will not cause steam binding and delay BOCREC.

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a.

4. NRC Concern - Fuel Rod With Too Quench Front -

Operation of a safety injection train injects ECCS water into the upper plenum. The Westinghouse model assumes that this UPI water covers a fixed percentage of the core, and that it will flow downward into that part of the core and cause a top-down quench.

i 31 1

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

{ -, WESTINGHOUSE PROPRIETARY CLASS 3 l Westinghouse assumed that the time required for the top-down quench to reach i the midplane of a fuel bundle was a specific linear function of the UPI mass flow rate per fuel assembly. Adequate technical justification of the function describing the top-down quench would require (a) auch more specific i

identification of the experimental data used to develop the quench-time function (b) showing that the data applies to the UPI situation, and (c)

• showing that the flow per assembly is a valid correlation parameter.

I Westinghouse stated that if used a quench-time function which is a ' bound to l

1 the [ experimental data) since more rapid quenching increases heat transfer to l UPI water and results in more steam generation." This more rapid steam j generation is presumably detrimental because it will impede reflood. However,

) to justify use of this bounding quench function, it would be necessary to show l that an overestimate of top-down quench speed (and/or of UPI flow per -

l assembly) will not cause an underestimate of peak cladding temperature.

1 a, E, c, i

5. NRC Concern - Core Heat Transfer Model The Westinghouse non-UPI EM uses very simple heat generation and heat transfer

{ models to determine core exit fluid conditions. Westinghouse has made what appear to be relatively simple modifications to its non-UPI EM to account for heat transfer in the top-down quench region. No adequate justification is .

given for using this simplified model under UPI conditions.

l -

1 4 J i

i 1

32

WESTINGHOUSE PROPRIETARY CLASS 25

- -n cLgE,CL

_J i _- ,

6. NRC Concern - Core Steam Generation

• It is necessary to clarify the location of the UPI water, and whether part of it is held up in a pool in the upper plenum, and it is necessary to demonstrate that the correlation for carryover from core to upper plenum is applicable to two-loop UPI plants.

The staff notes that water can be injected into the upper plenum at a rate which may be more thar.10 times as large as the rate which water can be injected into the intact cold leg, and that these flows determine the peak cladding temperature after the accumulators are empty. The licensees must submit a reflood and refill model which considers these flows and relevant thermal and hydraulic characteristics. The licensee must specifically show how and v:hy the water injected into the intact cold leg is or is not bypassed throughout the transient.

ra (1, E , C.

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7. NRC Concern - Condensation

~

The licensee must explain and justify (a) what happens if the falling subcooled UPI water can condense all the rising steam in the covered region, and (b) equations and assumptions coupling the covered and uncovered regions.

33

s WESTINGHOUSE PROPRIETARY CLASS 3 a , E , C-

8. NRC Concern - Vertical Entrainment Steam may be generated by the UPI water falling through the core. This steam is assumed to entrain and carry part of the UPI water upward. Westinghouse has not shown that the experimental data on entrainment were applicable to the conditions in a UPI plant, or that there exists a reasonable technical basis for extrapolating the correlation outside the data base.

CL, E, c.

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WESTINGHOUSE PROPRIETARY CLASS 3 s

l

9. NRC Concern - Horizontal Entrainment  ;

j Westinghouse assumed that 1.675 of the ECCS water which is injected into the

! upper plenum is entrained by horizontally moving steam which carries this 1.675 of UPI water into the hot legs. The Westinghouse entrainment model has not been proven valid because the model is based on (a) entrainment test data

which have not been demonstrated to be applicable to the conditions in the upper plenum of a UPI plant, and (b) air flow tests which have not been documented enough to be reviewed. Any new submittal will have to address the i

data from the CCTF experiments which imply that a frothy mixture exists up to j the hot leg nozzle.

-. O , E , C.

l i

l

10. NRC Concern - Total Steam Addition Due to UPI No theoretical justification has been given for assuming that there is no

interaction of the steasewater mixture rising vertically from the bottom quench front and the steam-water mixture rising vertically from the falling 4 UPI liquid. Further, the quality of each steam-water mixture is calculated by a different method, and there has been no discussion of the effects of different flow rates even though the two flow rates may be vastly different.

i 2

35

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WESTINGHOUSE PROPRIETARY CLASS 3 6

G . E C.

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, WESTINGHOUSE PROPRIETARY CLASS 3 5.0 SCHEDULE AND PROPOSED MEETINGS WITH THE NRC STAFF Through the u,se of a more[ ,

] it is believed that the NRC technical concern on the[ .

, 3"Also.[... .

o.

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]The[ .

- ]a,6A schedule for the program plan is given in Figure 1.

Meetings with the NRC staff are planned during this program and will provide theStaffwith{.-

].b These meetings will also serve to

. ]iey assumptions [

]RThree meetings are proposed:

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WESTINGHOUSE PROPRIETARY CLASS 3

6.0 CONCLUSION

A detailed plan has been developed to address the NRC concerns on the performance of the upper plenum injection emergency core cooling system for a postulated large-break LOCA transient. This plan utilizes [ .

e

]'i'h$iendresultofthis effort will be to [ .

j 1

I 39

i WESTINGHOUSE PROPRIETARY CLASS 3 7.0 REFERENGil

1. Rulemaking Hearing: " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Reactors". Docket R. M.- 50-1 (1973).

2. NRC Staff Report, Analysis Branch Division of Systems Safety, ' Safety Evaluation Report on ECCS Evaluation Model for Westinghouse Two-loop Plants," November 1977.

~

3. . . . . , . .

]

~

]CL

5. Letter dated February 12.1985 J. R. Miller (NRC) to D. M. Musolf (NSP) titled, " Development of an Acceptable ECCS Evaluation Model Which. Includes the Effect of Upper Plenum Injection Prairie Island Nuclear Generating Plant Units Nos. 1 and 2".

6. NRC Staff Report, " Emergency Core Cooling System Analysis Methods,"

USNRC - SECY-83-472. November 1983.

(c,E

7. Thurgood, M. J., et al. , "[ ]- A Thermal Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems," NUREG-CR-3046, Volumes 1-5, March 1983.

8. " TRAC-PD2: An Advanced Best Estimate Computer Program for Pressurized Water Reactor Loss of Coolant Accident Analysis' Los Alamos National Laboratory Report, LA-8709-MS, NUREG-CR-2054 (1981).

9. Ishii, M., " Drift Flux Model and Constitutive Equations for Relative Motion Between Phases in Various Two-Phase Flow Regimes," ANL-77-47, October 1977.

40 l

, WESTINGHOUSE PROPRIETARY CLASS 3

10. Wheeler, C. L., et. al. " COBRA-IV-I: An Interim Version of CO8RA For Thermal-Hydraulic Analysis, Rod Bundle Nuclear Fuel Elements and Core,"

BNWL-1962, March 1973.

11. Reeder D. L., ' LOFT System and Test Description" NUREG/CR-0247, .

July 1978.

12. McIntyre, 8. A. et. al. " Full Scale Controlled Transient Heat Transfer itsts-Data Report," EPRI-NP-1810, April 1981.

13. [ -a 14.E . . .

15. Hirano, K., et al., Quick-Look Report on Large Scale Reflood Test-23",

JAERI - Memo 9767, October 1981.

16. Blaisdell, J. A., Waring J. P. and L. E. Hochreiter, "PWR FLECHT-SET Phase A Report, WCAP-8238, December 1973.

17. Hochreiter, L. E., et al., " Mixing of ECC Water With Steam: 1/3 Scale Test and Sununary". EPRI 294-2, June 1975.

, 18.{

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APPENDIX A 46 NOMENCLATURE

[ ]

e 42

A _, a _

l i

APPENDIX A TWO FLUID PNASIC CONSERVATION EQUATIONS

APPENDIX NOMENCLATURE e internal energy l f body force

y ac[elgration of gravity

h enthalpy I phase. interface

} identity tensor k thermal conductivity 1/Lj surface area concentration for the jth interface 1/Ls total surface area concentration -

M interfacial momentum exchange g interfacial drag force

[ interfacial momente exchange due to phase change i mass filow rate n unit exterior nomal vector P pressure a conduction heat flux 6 volumetric heat generation rate .

T S turbulent heat flux gj' interfacial heat transfer per unit volme

$ phase interface stress tensor I turbulent (Reynolds) stress tensor T temper'ture a

1 stress vector t time at averaging time interval U fluid velocity 3 interface velocity '

V velocity of volme V A.1

1 r

).

I l

r V arbitrary volisme x position vector .

Greek Symbols 8V boundary of V v normal displacement speed of interface j . .

3 a void fraction s phase function -

r not rate of interfacial mass transfer per unit volisme i A net rate of interfacial energy transfer per unit volume 1 second viscosity coefficient y viscosity e density -

I surface tension vector g viscous stress tensor (stress deviator)

$ arbitrary fluid property

,$ubscripts 1 ' phase 1 2 phase 2 Ig phase 1 interfacial limit 2g phase 2 interfacial limit l I interfacial

! j Jth interface k phase k .

kg phase k interfacial limit Superscripts ,

I i interfacial surface average j s material T turbulent t transpose t A.2