ML20196E162

Requests That Proprietary Addendum 2 to Rev 1 to WCAP-10924-P, Westinghouse Large Break LOCA Best Estimate Methodology... Be Withheld from Public Disclosure (Ref 10CFR2.790)

ML20196E162
Person / Time
Site:	Point Beach
Issue date:	11/29/1988
From:	Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Murley T Office of Nuclear Reactor Regulation
Shared Package
ML19295G789	List: ML20196E157 ML20196E162 ML20196E175
References
CAW-88-127, NUDOCS 8812090283
Download: ML20196E162 (21)

Text

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Weninghouse Electric Corporation Eurg Systems g@d Ben 355 PittSDurg5 Pev4fvm 1523}0355 i

November 29, 1988 CAW 88 127 Dr. Thomas Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

, APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC OISCLOSURE

Subject:

WCAP-10924 P Revision 1, Addendum 2 (Proprietary) and WCAP 12071 (Non Proprietary " Westingiouse large Break LOCA Best Estimate Methodology: Vo ume 2: Application to Two-Loop PWRs Equipped with Upper Head Injection, Addendum 2, PBNP Plant Specific Analysis"

Dear Dr. Murley:

The proprietary information for which withholding is being requested in the enclosed letter by Wisconsin Electric Power Company is further identified in

Affidavit CAW 88 127 signed by the cwner of the proprietary information, Westinghouse Electric Corporation. The afficiavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Comission and addressed with specificity the considerations listed in paragraph (b)(4) of 10CFP. Section 2.790 of the Comission's regulations. i Accordingly, this letter authorizes the utilization of the accompanying  !

affidavit by Wisconsin Electric Power Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW 88 127, and should be addressed to the undersigned.  ;

Very truly yours,

,i WESTINGHOUSE ELECTRIC CORPORATION  :

,e T l

~

R M L ' N esemann, ert .

U U(ManagerCE WL/

Regulatory & Legislative Affairs _,

I cc: E. C. Shomaker, Esq.

Office of the General Counsel, NRC SS12090283 GS1130 PDR ADOCK 05000266

'__ P PDC

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P PROPRIETARY INFORMATION NOTICE l

l l

TRANSMITTED HEREWITH ARE PROPRIETARY AND/OR NON PROPRIETARY VERSIONS OF l DOCUMENTS FURNISHED TO THE NRC IN CONNECTION WITH REQUESTS FOR GENERIC AND/0R PLANT SPECIFIC REVIEW AND APPROVAL.

IN ORDER TO CONFORM TO THE REQUIREMENTS OF 10CFR2.790 0F THE C0m ISSION'S

REGULATIONS CONCERNING THE PROTECTION OF PROPRIETARY INFORMATION SO SUBMITTED i TO THE NRC THE INFORMATION WHICH IS PROPRIETARY IN THE PROPRIETARY VERSIONS IS I CONTAINED WITHIN BRACKETS MD WHERE THE PROPRIETARY INFORMATION HAS BEEN [

DELTEED IN THE NON PROPRICTARY VERSIONS ONLY THE BRACKETS REMAIN, THE  !

INFORMATION THAT WAS CONTAINED WITHIN THE BRACKETS IN THE PROPRIETARY VERSIONS  ;

i HAVING BEEN DELETED. THE JUSTIFICATION FOR CLAIMING THE INFORMATION SO  !

4 DESIGNATED AS PROPRIETARY IS INDICATED IN BOTH VERSIONS BY MEANS OF LOWER CASE [

j LF,TTERS (a) THROUGH (g) CONTAINED WITHIN PARENTHESES LOCATED AS A SUPERSCRIPT j IMMEDIATELY FOLLOWING THE BRACKETS ENCLOSING EACH ITEM OF INFORMATION. THESE

! LOWER CASE LETTERS REFER TO THE TYPES OF INFORMATION WESTINGHOUSE CUSTOMARILY I HOLDSINCONFIDENCEIDENTIFIEDINSECTIONS(4)(ii)(a)THROUGH(4)(ii)(g)0FTHE f

AFFIDAVIT ACCOMPANYING THIS TRANSMITTAL PURSUANT TO 10CFR2.790(b)(1). I r

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CAW 88-127 AFFIDAVIT COPHONWEALTH OF PENNSYLVAN'A:

ss COUNTY OF ALLEGHENY:

] Before me, the undersigned authority, personally appeared Robert A. Wiesemann, who, being by me duly sworri according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation ("Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief: l

// -

81114111/ L

\

Robert A. Wiesemann, Manager  ;

Regulatory and Legislative Affairs [

Sworn to and subscribed before me this Mday  :

i of 7.*W*dW,' 1988, i ti l

o d d .- w.f i .o .

Notary Public l

wtwr sts LOWAE u PACA,NotMYMJeUC W.ACEct mao ALE"#ENf COUNTY l Wf cCMV$$CN [1P 4($ C(014, tggt l Wener.Poveiew Amew,4>WNstaw

__________________________________-_____________y_________

CAW 88 127 (1) I am Manager, Regulatory and Legislative Affairs, in the Nuclear and Advanced Techno1cgy Divisia, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems, Nuclear Fuel, and Power Generation Business Units.

(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding r.ccompanying this Affidavit.

(3) I have personal knowledse of the criteria and procedures utilized by the Westinghouse Energy Systems, Nuclear Fuel, and Power Generation Business Units in designating information as a trade secret, privileged or as confidential comercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Comission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(1) The information sought to be withheld from public disclosure is 9 owned and has been held in coafidence by Westinghouse.

e 3 CAW 88 127 (ii) The information is of a type customarily held in confidence by l Westinghouse and not customarily disclosed to the public.

Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that  !

connection, utilizes a system to determine when and whether to  :

hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required. i Under that system, information is held in confidence if it falls ,

in one or more of several types, the release of which might I result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of 4 e process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors  ;

without license from Westinghouse constitutes a competitive  ;

economic advantage over other companies, e i

(b) It consists of supporting data, including test data, i relative to a process (or component, structure, tool, j method, etc.), the application of which data secures a l competitive economic advantage, e.g., by optimization or {

improved marketability. .

8

eo. .

s .

CAW 88 127 (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or connercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(g) It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors.

It is, therefore, withheld from disclosure to protect the ,

Westinghouse compe+8t've position.

CAW 88 127 (b) It is information which is marketable in many ways. The extent to which such information is available to I

competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantaga to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

e 0

• 6- CAW 88-127 (iii) The information is being transmitted to the Coenission in confidence and, under the provisions of 10CFR Section

~

2.790, it is to be received in confidence by the Comission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method 1 to the best of cur knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in

' Westinghouse Ltrge Break LOCA Best Estimate Methodology:

Volume 2: Application to Two Loop PWR's Equipped with Upper Head Injection. Addendum 2 PBNP Plant Specific Analysis,"

WCAP 10924. Revision 1, Addendum 2 (Proprietary), for Point Beach Nuclear Plant Stations 1 and 2, being transmitted by i the Wisconsin Electric Company (WEPCO) letter and Application for Withholding Proprietary Information from Public Disclosure, C. W. Fay, V.P. Nuclear Power, WEPCO, to Document Control Desk, USNRC, November,1988. The proprietary information as submitted for use by Wisconsin Electric Cotpany for the Point Beach Nuclear Plant Units 1 J and 2 is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of compliance with 10CFR50.46 Appendix K rule and changes in technical specifications.

This information is part or that which will enable Westinghouse to:

O '

CAW 88 127 (a) Provide documentation of the analyses for Point Beach Nuclear Plant Units 1 and 2 to show compliance with ,

10CFR50.46 Appendix K using the SECY 83 472 approach and the NRC approved Westinghouse COBRA / TRAC code.

l (b) Estabitsh new technical specifications based on the Point Beach Nuclear Plant analysis for the maximum linear heat rate and hot assembly power as well as the maximum steam generator tube plugging levels allowed.

(c) Resolve the N?.C concern on the performance of the upper plenum injection system safety system in the unlikely event of a LOCA.

(d) Apply a new NRC approved methodology for large break LOCA analysis of the Point Beach Nuclear Plant.

(e) Assist the customer in obtaining NRC approval of his plant specific calculations and changes to the Point Beach Nuclear Plant Units technical specifications.

Further this information has substantial comercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for purposes of showing i adequate performance of the upper plenum injection system to resolve the NRC concern on this issue.

(b) Westinghouse can sell support and defense of this technology to its customers in the licensing process.

_---_---------------------------------y------

4 CAW.88-127 Public disclosure of this proprietary information is likely to cause substantial harm to the competitivo position of Westinghouse because it would enhance the ability of competitors to provide similar analytical documentation and licensing defense services for commercial power reactors without comensurate expenses. Also, pubite disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many

years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be i performed and a significant manpower effort, having the

] requisite talent and experience, would have to be expended l for the development verification, and licensing of i acceptable analytical methods which can be applied to this class of PWR's.

Further the deponent sayeth not.

I l

i l

l

POINT BEACH NUCLEAR PLANT TECHNICAL BASIS FOR EXEMPTION TO SELECTED APPENDIX K REQUIREMENTS

1. INTRODUCTION The emerger.cy core cooling system for all Westinghouse domestic two-loop i

pressurizedwaterreactorsinjectsthelowpressureEmergencyCoreCooling System (ECCS) cooling water directly into the upper plenum of the reactor in the event of a LOCA. Westinghouse three and four-loop pressurized water reactors inject the low pressure cooling water into cold legs where it flows into the downcomer and then into the lower plenum. In the past, Loss of Coolant Accident (LOCA) analysis for two-loop plants assumed that during reflood the low pressure water was injected into the lower plenum (core flooding from below) in the same manner as for the three-loop and four-loop plants. With this assamption, 10CFR50, Appendix K could be applied to the analyses without exception.

The NRC is concerned that the analytical assumption of low pressure water injecting into the lower plenum is unrealistic, and potentially non-conser-vative for two-loop pressurized water reactors (Reference 1). As a result of this concern, Wisconsin Electric Power Company, Northern States Power Company and Westinghouse Electric Corporation have developed a new LOCA model for plants with upper plenum injection (References 2 to 4) The new Upper Plenum Injection Best Estimate Methodology models the injection of low i

pressure ECCS water directly ir to the upper plenum.

In the process of reviewing this new model against the 10CFR50 Appendix K requirements, two Appendix K requirements were identified as not applicable i to two-loop plants with upper plenum injection. These two requirements, i

Sections I.D.3 and 5, were written for bottom flooding plants (i.e., cold leg injection plants) and compliance with these requirements for plants i with upper plenum injection would not serve the underlying purpose of the rule. The inapplicable requirements are:

1 Section I.D.3 Calculation of Core Exit Flow Based on Carryover Fraction i

Section 1.0.5 Calculation of Heat Transfer During Refill and

! Reflood.

Both of these requirements are imposed on the calculation for the refill

, and reflood portion of the transient. Section 2 below, describes the refill and reflood phases of the large break LOCA in cold leg injection and upper plenum injection plants. Section 3 contains the applicable Appendix K requirement, the basis or original intent of the requirement, and the proposed analysis methods to be used for upper plenum injection j plants.

i i

1  !

o . .

l 2. DESCRIPTION OF CALCULATED LOCA TRANSIENT Introduction In order to examine the different thermal-hydraulic behavior of a two-loop PWR with UPI for a postulated LOCA, a PWR LOCA transient with cold leg i injection is reviewed. The two-loop UPI PWR transient is then contrasted ,

tothecoldleginjectionPWR, [

i l Cold Lea Injection Plant j The large break LOCA transient includes three phases: blowdown, refill  !

j and reflood. Figure 1 shows the duration of each phase and the accumulator, l 1 low pressure and high pressure cooling water flow rates during each phase, j The timing and injection flot -ates in Figure 1 are from a licensing calcula- i tion for a double-ended cold -g guillotine break with a 0.4 discharge L coefficient for a cold leg in, etion plant. l 1

During blowdown, the vessel and loops depressurize and most of the fluid in l the vessel and loops goes out the break into the containment. Blowdown j; ands before the low pressure and high pressure cooling water injection is I assumed to start.

1 During refill, flow out of the break has ceased and the lower plenum and  !

I downcomer start to fill from water injected into the cold legs from the cold leg accumulator. The refill period lasts about 10 to 15 seconds and ends when the rising water level reaches the bottom of the core. The L 4

accumu'stors inject for the entire refill period, while the low pressure (

and high pressure cooling water start injecting near the end of the refill  ;

- period. As the lower plenum fills, it is assumed that there is only [

J radiation cooling in the core, and the fuel rods heat up nearly  :

I adiabatically. [

L l

! Morerecentcoldleginjectioncalculationswithrealisticmodelsindicate [

] that some flow and core cooling will occur dur:ng refill. However, at the t j time the rule was written, these calculations were not available and it l j was deemed prudent to require a conservative approach in this area.

[

! Reflood starts when the rising water level reaches the bottom of the core, f i and continues until the entire core is quenched (usually calculated to be }

j several hundred seconds after the start of reflood in large break LOCA j i calculations based on conservative licensing assumptions). The accumula- j 1 tors empty about 5 seconds after the start of reflood, so the low pressure  ;

j and high pressure systems provido the injection flow for the remainder of [

i the transient. Throughout the reflood period, the core refloods from flow  !

) entering the core from the lower plenum. l i

Upper Plenum Injection Plant The sequence of events of the large break I.0CA transient in the two-loop l

plantwithupperplenuminjectionissimilartothatcalculatedforacold f

! leginjectionplantintheblowdownphasesinceblowdownendsjustafter  !

l thelowheadsafetyinjectionbeginstoinjectintotheupperplenum,  !

1 Sensitivity calculations indicate that the assumption of maintaining i i

i  !

2-I

1

!- l

• . l off-site power yields hf0her calculated peak cladding temperatures. There-  !

j fore, there will be some high head safety injection into the cold legs during the end of blowdown, and the low head injection into the upper plenum .

i will begin once the system pressure drops below the low head SI pump shutoff

! head of approximately 145 psi. The refill and reflood phases have significant  ;

j differencesduetotheinjectionofthelowheadcoolingwaterintotheupper  !

i plenum. These differences are described below.  !

I I i Refill - With upper plenum injection, the low head safety injection starts f bcfore the end of blowdown, and begins to inject flow into the upper plenum, l which penetrates into the core. Since there is now a direct source of cooling (

1 water which flows down through the reactor core, core cooling is possible j during refill (References 5, 6 and 7). The accumulator flow and high head 1 safety injection flows are injected into the cold legs in the same fashion F

] as a cold leg injection PWR. j t

Reflood" - Accumulator injection into the cold legs continues for about the j i

first 5 seconds of reflood. During this period, core cooling occurs from (

l both bottom flooding resulting from accumulator injection and top flooding i 1 whichoccursfromupperplenuminjection. After accumulator injection ends.

however, water is added to the core mainly from above by water injected into [,

l the upper plenum. A smaller amount of high pressure cooling water is injected i into the cold legs. Recent detailed WCOBRA/ TRAC (Reference 4) calculations l l indicate that the UPI flow will easily penetrate lower power fuel assemblies 1

on the outside of the core and will flow down into the core. Since these j assemblies are at lower power, there is less steam generation. The UPI l 1 flow from the cold channels will crossflow below the quench front to the other

assemblies. The remainder of the core will be in a combination of coeurrent [

l upflow and countercurrent flow as the UPI flow in the upper plenum penetrates the upper core plate into the fuel region. The water accumulation rate into f the vessel plus the liquid entrainment rate up out of the core is smaller [

{ than the UPI delivery rate such that both the core and downcomer fill, even s i though the net core flooding rate is zero or negative. Heat is transferred '

to a cocurrent or countercurrent two phase mixture in the core which terminates the temperature rise at the core hot spot. This flow pattern has been observed in UPI simulation tests in the Japanese Cylindrical Core ,

Test facility (Reference 8) and in a thermal hydraulic calculation of a UPI l plant LOCA, performed by Sandia (Reference 9). l l

l 3. BAS!$ FOR EXEMPTION FROM APPENDIX K REQUIREMENTS AND PROPOSED~~

) ALTERNATE ANAC. E G IETH005 i

Carryover Fr',etion (Rule !.0.3) I

{

Appendix K Requirement -

t The ratio of the total fluid flow at the core exit plane to the total j liquid flow at the core inlet plane (carryover fraction) shall be used j t

1 "To permit comparison with the cold leg injection plant, the term >

l "reflood" is used here for the UPI plant to describe the period after [

the rising lower plenum water level reaches the bottom of the core.

However, as described above, the core may be flooded from above even before the "reflood" period starts, j, l

1 I

. l to determine the core vit flow and shall be determined in accordance with applicable experimentel data (for example, "PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report," Westinghouse Report WCAP-7665, April 1971; "PWR Full Length Emergency Cooling Heat Trantfor (FLECHT) Group ! Test Report," Westinghouse Report WCAP-7435, January 1970; "PWR FLECHT (full Length Emergency Cooling Heat Transfer)

Group !! Test Report," Westinghouse Report WCAP-7544, September 1970; "PWR FLECHT Final Report Supplement," Westinghouse Report WCAP-7931, .

October 1972).

Basis /Oricinal Intent of Requirement - The core flooding rate depends on the pressure drop tnrough the reactor coolant loops, the core liquid level i and the downcomer liquid level. As the downcomer level increases, it 2

forces more liquid into the core. Some of this liquid accumulates in the core as the vessel fills while a larger fraction of the core inlet mass flow is vaporized due to the core heat release. Vapor generated in the core  :

carries entrained liquid out of the core into the loops. Accordingly, accurate calculation of core flooding rate requires accurate calculation i I of core exit flow rate. When Appendix X was written, the NRC felt that i available codes could not accurately calculate the core exit flow rate.

As a result, Appendix K required core exit flow rate to be calculated i

using experimental data. Specifically, the core exit flow was determined

! from the code-calculated core inlet flow times a carryover fraction developed from FLECHT data. Using the terminology in Figure 2(A):

W core exit *I*N core inlet where f = carryover fraction, determined from FLECHT data as follows:

N i f, core inlet ~ "Neore/dt N

L,in

! where dHcore/dt is the core mass storage rate.  ;

The carryover fraction in the FLECHT tests ranged from about 0.8 to 0.9, i.e., 800 to 90% of the inlet flow was measured to leave the top of the Core. ,

The intent of this requirement was to ensure the flow exiting the core to  ;

the loops was calculated by the most appropriate means available. When -

Appendix K was written, the data-based calculation was considered more t appropriate than the code calculation for the bottom-flooding PWR, using the codes then available. (

I Why Requirement is inapplicable for UPI Plant - The exit flow calculation i method and the cited FLECHT data are for a bottom flooding situation, j where the liquid flow direction at the bottom of the core ("inlet plane")

is upward and the flow within the core at the top of the core ("exit flow")  !

is also coeurrent upward, as shown in Figure 2(A). In a plant with upper plenum injection, the liquid enters at the top of the core and exits at l

4 i  !

both the bottom (water) and at the top (steam and water) as shown in Figure 2(8) and can flow both in a countercurrent and cocurrent fashion in the core. Therefore, the definitions of inlet and exit are different in the two types of plants as well as the flow patterns in the core. To meet the intent of the Appendix K requirement, the liquid and steam flow from the core to the upper plenum is needed. For the cold leg injection plant, this is the core exit flow. The ratio of this exit flow to the inlet flow for the UPI case is significantly different than that in the bottom flooding situation, since the flow situation is markedly different. For example, in a typical CCTF UPI test (Reference 8) the reflood core exit steam mass flow (W,) was about 40% of the net liquid downflow at the top of the core (WL,down NL,up); most of the remainder, about 60%, went out the bottom of the core (W gge,) and then went up the downcomer to the ':old leg break as shown in Figure 2(B). The CCTF instrumentation did not permit separate determination of the liquid downfi m and liquid upflow at the tcp of the core. However, assuming the upward entrained water flow at the top of the core was small, the core exit flow (WL,up + Ws ) was about 40% of the inlet flow (WL,down) in the CCTF UPI tests, compared to 80% to 90% in the FLECHT bottom-flooding tests. Accordingly, both the definitions of "inlet" and "exit," and the relative magnitudes of the flows and flow directions and patterns, are significantly different in the two types of plants. Accor-dingly, the cited FLECHT data, and the prescribed method of calculating core exit flow do not apply to the UPI plant.

Proposed Analysis Methods for the UPI Plant - The intent of the Appendix K rule, accurate calculation of core exit flow, can be met by usin0 a code, which has been verified against appropriate experimental data, to

, calculate core exit flow rate.

The WCOBRA/ TRAC code (References 3 and 4) is an improved version of the COBRX/ TRAC code which has been recently developed to predict the thermal-hydraulic response of reactor systems to large and small break loss of coolant accidents. This code is a significant improvement over the codes that existed at the time Appendix K was written. WCOBRA/ TRAC uses a separated flow, two phase flow model in which there are three fields for the two phases; a continuous liquid field to model liquid film and low void fraction flows, a dispersed liquid field to model droplet flows, entrainment and de-entrainment; and a vapor field to model the gas phase. Each field has its own mass continuity equation and momentum equation. Within a given computational cell, the two liquid fields are assumed to be at the same temperature, while the vapor field can be at a separate temperaturet hence, there are two energy equations. The interactions between each field are modeled through interfacial heat, mass, and momentum transfer using locally calculated heat transfer and fluid drag relationships. Using this formula-WCOBRA/ TRAC can model the complexities of a two phase, nonequilibrium tion, flow sItuation such as that found in the PWR's equipped with UPI.The WCOBRA/ TRAC code calculates the amount of flow which penetrates down into the reactor vessel. It also predicts the net amount of steam upflow from the vessel to the loops, and it predicts what fraction, if any, of the water injected into the upper plenum is entrained out of the plenum into the loops. The WC0 ERA / TRAC formulation permits accurate calculation of inter-phase heat and mass transfer, entrainment, de entrainment, countercurrent 5-

$ flow, and liquid pooling such that steam and water flow c.ryuver into the g hot legs for PWR's with UPI can be accurately predicted.

To assess WCOBRA/ TRAC's capability for predicting the correct thermal-hydraulic behavior for upper plenum injection situations, WCOBRA/ TRAC has been compared to the Japanese Cylindrical Core Test Facility data which models the interaction effects of upper plenum injection in a large scale test facility (Reference 3). WCOBRA/ TRAC predicts the thermal-hydraulic effects of the upper plenum injection such that the carryover of steam and water into the hot legs is accurately calculated. The use of WCOBRA/ TRAC will meet the intent of Requirement I.D.3 of Appendix K.

Refill / Reload Heat Transfer (Rule !.0.5)

Appendix K Requirement -

For reflood rates of one inch per second or higher, reflood host transfer coefficients shall be based on applicable experimental data for unblocked cores including FLECHT results ("PWR FLECHT (full Length Emergency Cooling Heat Transfer) Final Report " West-inghouse Report WCAP 7655, April 1971). The use of a correlation derived from FLECHT data shall be demonstrated to be conservative for the transient to which it is applied; presently available FLECHT heat transfer correlations (PWR Full Length Emergency Cooling Heat Transfer (FLECHT) Croup 1 Test Report," Westinghouse Report WCAP-7544, September 1970; "PWR FLECHT Final Report Supplement,"

Westinghouse Report WCAP-7931, October 1972) are not acceptable. New correlavns or modificaticiis to the FLECHT heat transfer correlations are acceptable only after they are demonstrated to be conservative, by comparison with FLECHT data, for a range of parameters consistent with the transient to which they are applied.

During refill and during reflood when reflood rates are less than one inch per second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding welling or rupture as such blockage might affect both local steam flow and heat transfer.

Basis / Original intent of Requirement The rule prescribes heat transfer calculation methods for three cues; refill, reflood with flooding rate less than one inch per second, and reflood with flooding rate greater than one inch per second. For refill, the assumption of steam cooling is required in the rule because it was felt that there would be no water in the core during this period. For reflood, the requirements and the one inch per second threshold were chosen to ensure the effects of flow block-age were conservatively accounted for. At the time, a limited amount of flow blockage testing bad been performed in the FLECHT facility, using a perforated plate to simulate the flow blockage. Tests were performed at flooding rates of 0.6, 1.0, 2.0 and 6.0 inches per second. These tests indicated enhanced heat transfer due to blockage at flooding rates of one inch per second and higher because of increased turbulence and draplet break-up. The 0.6 inches per second flooding rate test indicated that the blocked bundle heat transfer was degraded relative to a similar unLlocked bundle at the same flooding rate. The degraded heat transfer was presumed 6-

i f

to be caused by liquid de entrainment at the blockage leaving only steam '

cooling heat transfer. Based on this data, the requirement for flooding rates greater than one inch per second was written to require that heat transfer coefficients be based on undistorted geometry data; this was judged acceptable since the FLECHT data indicated the data would be conservative if blockage were to occur. For flooding retes less than one inch per second, the assumption of steam cooling was required since the FLECHT data indicated this would be the flow regime if blockage were to occur; if there '

was no blockage, this assumption would be conservative, t More recent data, summarized in Appendix F of Reference 4, indicates that j there is no heat transfer penalty for flooding rates below 1 inch per second for blockage shapes which bour.J the most prototypical blockage '

geometries found in out of pile, and in pile experiments. The experiments also indicate that the flow above the quench front reasins two phase with  !

liquid entrainment down to flooding rates as low as .4 inch per second,  !

h such that steam cooling only does not occur during reflood,  ;

i Why Requirement is Inapplicable for a UPI Plant - Fer a PWR with upper plenum injection, the flow patterns and resulting heat transfer n.echanisms I are different than those assumed in the Appendix K rule. The specific

! differences are the following:  ;

(1) During refill in the UPI plant, the water injected into the upper i' i plenum will fall into the core and contribute to core cooling.

Therefore, the assumption of steam-cooling only during refill is t I

j inappropriate for the UPI plant. Further, the heat transfer mechanisms during rCfill are similar to those during reflood in the

! UPI plant, so it would be inconsistent to arbitrarily retain this

]

requirement. r 1 '

) (2) The one inch per-second' flooding rate threshold for steam cooling during reflood is based on bottom-flooding blockage heat transfer a data. This threshold is inappropriate for the UPI plant for two i i reasonst (a) the value of the threshold has no meaning for the UP! ,

) plant, because of the different flow situations (see discussion in '

J the "Carryover Fraction" section above), and (b) the local flow j patterns are dif ferent, so the behavior observed in the FLECHT  ;

3 reflood and blockage tests is not apprepriate. Specifically, the  !

4 FLECHT reflood and blockage data are for a bottom flooding  !

situation, with only cocurrent upward ste.im and water flow every- l where in the core (Figure 2(A)). Cooling is by dispersed coeurrent j

i' upflow film boiling and radiation. In the UPI plant, the net steam I flow is upward but the net flow of water is downward (Figure 2(B)). I g Further, the steam water flow patterns vary throughout the core

such that the rod surfaces are cooled by film boiling ano radiation j heat transfer resulting from a combination of cocurrent downflow, J cocurrcnt upflow, and countercurrent flow, as observed in the CCTF tests (References E, 6, 7, 8).

Proposed Analysis Methoos for UPI Plant - To meet the intent of Appendix K, which is to use the most applicable data for this situation, the WCOBRA/ TRAC

!l code has been verified against two independent sets of experimental data

- which model the upper plenum injection firs and heat transfer situation.

t i

The first series of tests which have been modeled by WCOBRA/ TRAC are the Westinghouse G-2 refill downflow and counterflow rod bundle film boiling experiwints (Reference 10). These experiments were performed as a full length 17x17 esi',0 house rod bundle array which had a total of 336 heated rode .r- ' action flow was from the top of the bundle and is scalable t- ' " njection flows. The pressures varied between 20-100 psi which is t'w e range for UPI top floodir.g situations. Esth cocurrent downflow f b i cnd countercurrent film boiling experiments were modeled u c ' TRAC. Both these flow situations are found in the calculate - 3ponse for a PWR with UPI.

In addition to modeling these separate effects tests, WCOBRA/ TRAC has been used to model the Japanese Cylindrical Core Test FacilTty experiments with upper plenum injection (References 5, 6, and 7). The tests whi,.:h have been modeled included test 72 which was a symmetrical UPI injection w;th maximum injection flow, test 59 which was minimum injection flow with a near ty symmetrical injection pattern, test 76 which was a minimum UPI injec-tion flow with a skewed U?! injection and test 54 which was a cold leg injection reference test for the UPI tests. A detailed three dimensional WCOBRA/ TRAC calculation, sponsored by the USNRC also was performed on test 72 (Reference 11). Coarser noded WCOBRA/ TRAC calculations were per-formed on tests 59, 72, and 76 using noding more typical of PWR evaluation model noding.

The results of these comparisons are documented in References 2 and 10 and show that WCOBRA/ TRAC does predict heat transfer behavior for these com-plex film Boiling situations as well as the system response for upper plenum injection situations.

The effect of flow blockage due to cladding burst is explicitly accounted for in WCOBRA/ TRAC with models which calculate cladding swelling, burst, and area reduction due to blockage. These models are based on previously 4 approved models used in current evaluation models (References 12 and 13) and on flow blockage modelt determined to be acceptable by the staff (Reference 14). The effect of flow blockage is accounted for from the time burst is calculated to occur. The fluid models in WCOBRA/ TRAC calcu-late flow diversion as a result of the blockage. Thus tIie intent of the rule, which requires that flow blockage effects to be taken into account, is met.

. 4. CONCLUSIONS The Westinghouse two-loop PWR's equipped with Upper Plentm Injection have unique features which rpake the applicati99 of certain Appendix K reflood requirements inappropriate. By using the best-estimate thea.lal-hydraulic comnuter code, WCOBRA/ TRAC, the intent of the Appendix K requirements can j be achieved. Therefore, it is proposed that the exemption from the in-appropriate reflood requirements be granted providing that WCOBRA/ TRAC is used to calculate the LOCA transient for PWR's equipped with UPI.

l s

1 8-

. - _ _ _ _ _ _ _ _ _ _ _ _ _ l

i REFERENCES

1) J. R. Miller to C. W. Fay, NRC letter dated February 13, 1985, entitled "Development of an Acceptable ECCS Evaluation Model which Includes the Effect of Upper Plenum Injection."

2) Letter dated April 19, 1985 from C. W. Fay, Wisconsin Electric Vice Prssident-Nuclear Power to H. R. Denton, Director, Office of Nuclear Reactor Regulation entitled "Analycis of Emergency Core Cooling.

System With Upper Plenum Injection Using SECY-83-472 Methodology, Point Beach Nuclear Plant Units 1 and 2."

3) Hochreiter, L. E. , Schwarz, W. R. , Takeuchi, K. , Tsai, C-X, and 8 Young, M. Y., "Westinghouse large-Break LOCA Best Estimate Methodology Volume 1: Model Description and Validation," WCAP-10924-P, April 1986 (Westinghouse Proprietary).

4) Dederer, S. I., Hochreiter, L. E., Schwarz, W. R., Stucker, D. L.,

Tsai, C-K, and Young, M. Y. , "Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWR's Equipped with Upper Plenum Injection," WCAP-10924, Volume 2, Revision 1 April 1988 (Westinghouse Proprietary).

5) Iguchi, T. , et al. , "Data Report on large Scale Reflood Test-99, CCTF Core-II Test C2-16 (Run 076)," JAERI-memo 60-158, February 1985, (JAERI-Proprietary).

6) Iguchi, T., et al., "Data Report on Large Scale Reflood Test-96, CCTF Core-II Test C2-13 (Run 072)," JAERI-memo 60-157, July 1985,

'(JAERI-Proprietary).

7) Iguchi, T., et al., "Data Report on Large Scale Reflood Test-79, CC)F Core-II Test C2-AS1 (Run 059)," JAERI-memo 59-447, February 1985, (JAERI Proprietary).

8) Iguchi, T. and Murao, Y. , "Experimental Study on Reflood Behavior in PWR with Upper Plenum Injection Type ECCS by Using COTF," J. Nuclear Science Technology 22(8), pp. 637 - 652 (August 1985).

9) Letter dated August 16, 1984 from L. Buxton (Sandia) to D. Langford (NRC).

10) Hochreiter, , et al., "G-2, 17x17 Refill Heat Transfer Tests and Analysis", WC. 4793, Au0ust 1976 (Westinghouse Proprietary). ,

11) Thurgood, M. J., and Wheeler, C. L.. "COBRA / TRAC Three-Dimensional Simulation of CCTF No Failure UPI Test C2-13 (Run 72)," Fate-86-108, Mar:b 386.

12) Young, M. Y., et al., "BART-A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A, March 1984 (Westinghouse Proprietary).

I 13) Bordelon, F. M., et al., "LOCTA-IV Program: Loss cf Coolant Transient Analysis," WCAP-8301, June 1984, i

14) Powers, D. A., and Meyer, R. O., "Cladding Swelling and Rupture Mode h

) for LOCA Analysis," NUREG-0630, April 1980.

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m AsSONES Legend: W = Hass Flow Rate  : -

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. UPI l 1

- 1 UPPER PLENUM UPPER PLENUM bpperPlenum 100P5 "Upper Plenum lo0PS -

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1 < r L,down "Core "Core .

CORE CORE n

. u W W L,in L. bottom (A) Bottom-Flooding Plant (B) UPI Plant PWR VESSEL FLOWS DURING REFLOOD

^

FIGURE 2

- - - - - _ _ _ _ _ _ _ _ _