ML20136H881

Requests Proprietary Meeting W/Nrc to Discuss Upper Plenum Injection Program.Proprietary Plan for Compliance & RA Wiesemann 851114 Ltr W/Affidavit Encl.Plan Withheld (Ref 10CFR2.790)

ML20136H881
Person / Time
Site:	Prairie Island
Issue date:	11/18/1985
From:	Musolf D NORTHERN STATES POWER CO., WISCONSIN ELECTRIC POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML19276C934	List: ML20136H881 ML20136H903
References
NUDOCS 8511250220
Download: ML20136H881 (27)

Text

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Northem States Power Company 414 Nicollet Matt Minneapolis. Minnesota 55401 Telephone (612) 330-5500 November 18, 1985 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISIAND NUCLEAR GENERATING PLANT Docket Nos. 50-28) License Nos. DPR-42 50-306 DPR-60 Request for Proprietary Meeting and Submittal of PROPRIETARY Material to be Reviewed at the November 22, 1985 Meeting The purpose of this letter is to request, on behalf of Northern States Power and Wisconsin Electric Power Companies, a proprietary meeting with the Staff to discuss our Upper Plenum Injection Program. An agenda for this meeting has been transmitted to the Prairie Island tac Project Manager. We plan to discuss the program status and review the features of our program which will demonstrate compliance with 10 CFR Part 50.46 and Appendix K.

To' facilitate a more productive meeting, a copy of our plan for compliance is enclosed for pre-meeting review. Since this plan, as well as other items to be discussed at the meeting, contain information proprietary to Westinghouse Electric Corporation, we request that this plan be withheld from public disclosure and the meeting be closed to the public. An affidavit in support of this request signed by Mr Robert A Wiesemann is attached. The affidavit sets forth the basis on which the information may be withheld from public disclosure and the basis on which the meeting can be closed to the public, and addresses with specificity the considerations listed in paragraph (b) (4) of Section 2.790 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.790 of the Commission's regulations.

ATTACHMENT 3 CONTAINS PROPRIETARY INFORMATION TO BE WITHHELD FROM Y{0 I

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4 PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR PART 2, SECTION 2.790

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Dir, NRR November 18, 1985 Page 2  ;

I Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should reference letter CAW-85-079 (Attachment 2) and should be addressed to R A Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P O Box 355, Pittsburgh, Pennsylvania 15230.

D M N= ,

David Musolf Manager - Nuclear Support Services DMM/TMP c: Regional Administrator-III, NRC NRR Project Manager, NRC l Resident Inspector, NRC G Charnoff Attachments:

1) Plan for Compliance with the Appendix K Required Features, L

Non-Proprietary Version

2) Letter from R A Wiesemann (Westinghouse) to H R Denton (NRC) dated November 14, 1985, " Application for Withholding Proprietary Information from Public Disclosure."

3) Plan for Compliance with the Appendix K Required Features, PROPRIETARY Version ATTACHMENT 3 CONTAINS PROPRIETARY INFORMATION TO BE WITHHELD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR PART 2 SECTION 2.790

Attachmsnt i Page 1 of 25

_ _A.C t e Plan for _ _

Model Compliance with the Appendix K Required Features This document describes how Westinghouse will _

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_ o.,C,e will be used to perform realistic PWR LOCA analysis. Some of the models and plant parameterassumptionsusedinthe[

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will serve as a base case from >

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]When performing the Appendix K licensing calculation, those specific models and i

assumptions which are required by 10CFR50.46 Appendix K[

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_ ]soastocomply with the requirements. This version of the code will then be used for the licensing calculations. The text below which is single snaced and indented istheAppendixKrequirement.[ ]4,E is then specifically addressed below each requirement.

AcDendix K Reauirement:

I. REQUIRED AND ACCEPTABLE FEATURES OF THE EVALUATION MODELS*

3 A. SOURCES OF HEAT DURING THE LOCA

< *By definition of 10CFR50.46(c)(2), "An evaluation model is the calculational framework for evaluating the behavior of the reactor system during a postulated loss-of-coolant accident (LOCA). It includes one or i more computer programs and all other information necessary for application of the calculational framework to a specific LOCA, such as nethematical models used, assumptions included in the programs, procedure for treating the program input and output information, specificatloa of those portions of analysis not included in computer programs, values of parameters, and all other information necessary to specify the calculational procedure."

• • Further, per 10CFR50.46(c)(1) LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system.

I 7889B:10/111385 11-1 1

Attachmant 1 Page 2 of 25 For the heat sources listed in paragraphs 1 to 4 below, it shall be assumed that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level (to allow for such uncertainties as instrumentation error), with the maximum peaking factor allowed by the technical specifications.

Method of ComDliance:

The power level used in the ECCS analysis is dif f erer.t f or each plant analyzed. The values are given in Table 15.6 of the 5atety Analysis Report (SAR) for each plant. The power level and peaking factor used comply with the above requirements.

ADDendiX K Requirement:

A range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results in the most severe calculated consequences, for the spectrum of postulated breaks and single failures analyzed.

Method of ComDliance:

a., C ,0.

ADDendiX K Requirement:

1. The Initial Stored Enerav in the Fuel. The steady-state temperature distribution and stored energy in the fuel before the hypothetical accident shall be calculated for the burn-up that yields the highest calculated cladding temperature (or optionally, the highest calculated stored energy). To accomplish this, the thermal conductivity of the U02 shall be evaluated as a function of burn-up and temperature, taking into consideration differences in initial' density, and the thermal conductance of the gap between the UO .and the cladding shall be evaluated as a function of the burn-up, taking into consideration fuel densification and expansion, the composition and pressure of the gases within the fuel rod, the initial cold gap dimension with its tolerances, and cladding creep.

7889B:10/111365 11-2

..= a Attachment 1 Page 3 of 25

2. Fission Heat. Fission heat shall be calculated using reactivity reactor kinetics. Shutdown reactivities resulting from temperatures and voids shall be given their minimum plausible values, including allowance for uncertainties, for the range of power distribution shapes and peaking factors indicated to be studied above. Rod trip )

i and insertion may be assumed if they are calculated to occur. l 1

3. Decay of Actinides. The heat from the radioactive decay of l

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actinides, including neptunium and plutonium generated during  !

operation, as well as isotopes of uranium, shall be calculated in accordance with fuel cycle calculation and known radioa:tive properties. The actinide deca: heat chosen shall be that appropriate for the time in the fuel cycle that yielos tne nignest calculated f uel temperature during the LOCA.

4. Fission Product Decay. The heat generation rates from radioactive decay of fission products shall be assumed to be equal to l.2 times the values for infinite operating time in the ANS Standard (Proposed American Nuclear Society Standards- " Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors".

Approved by Subcomittee ANS-5, ANS Standards Comittee, October 1971). The fraction of the locally generated gama energy that is deposited in the fuel (including the cladding) may be different from 1.0; the value used shall be justified by a suitable calculation.

5. Metal-Water Reaction Rate. The rate of energy release, hydrogen-generation, and cladding oxidation from the metal / water reaction shall be calculated using the Baker-Just equation (Baker L., Just, L.

C., " Studies of Metal Water Reactions at High Temperatures, III.

Experimental and Theoretical Studies of the Zirconium-Water Reaction," ANL-6548, page 7, May 1962) . The reaction shall be assumed not to be steam limited. For rods whose cladding is calculated to rupture during the LOCA, the inside of the cladding shall also be assumed to react after the rupture. The calculation of the reaction rate on the inside of the cladding shall also follow the Baker-Just equation, starting at the time when the cladding is

. calculated to rut *.ar.. and ext r din; e._nc the cladding inner circumference and axially no less than 1.5 inches eacn way from the location of the rupture, with the reaction assumed not to be steam limited.

Method of Compliance:

The above items 1-5 which deal with heat generation in the fuel and cladding willbemodeledin[

78898:10/111385 11-3

.. o Attachme't 1 Page 4 of 25 A,

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The core power at 1.02 times the nominal core power will be used. .

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Anoendix K Reauirement:

6. Reactor Internals Heat Transfer. Heat transf er f rom piping, vessel walls, and nonfuel internal hardware shall be taken into account.

Method of Compliance:

COBRA-TRAC models the heat transfer from internal structures to the fluid using a one-dimensional conduction formulation as shown in Volume 1 of the COBRA-TRAC manual ( }. Temperature dependent material properties are accounted for as well as any internal resistances or gaps. The heat transfer on the surfaces can range between single phase liquid or vapor to nucleate boiling, depending upon the wall temperature and the flow regime present. All the structural material in a node is accounted for such that the prcper metal heat release is calculated.

Anoendix K Reauirement: l

7. Pressurized Water Reactor Primary-to-Secondary Hea* Transfer.

Heat transferred between primary and secondary systems through heat exchangers (steam generators) shall be taken into account. (Not 1 applicable t:: Boiling Water Reactors.)

9 Method of Compliance:

The steam generator model, STGEN, for COBRA-TRAC is described in Volume 3 of the COBRA-TRAC manual (5) , and is a loop component from the TRAC-PD2 code (6) . The steam generator primary side and secondary side hydrodynamics are treated separately with the heat transfer coupling between the two sides.

7889B:10/111385 11-4

0 t Attachment 1 Page 5 of 25  ;

The initial steady-state heat transfer on the secondary side is forced convection and nucleate boiling while the primary side heat transfer is forced convection. Once the transient begins, the secondary side heat transfer will remain in a boiling mode or will become either natural circulation or forced circulation depending on the steam generator secondary circulation flow and the tube wall temperature.

The primary side heat transfer depends on the flow regime in the steam generator tubes as well as the calculated primary side surface temperature.

The selection logic is shown in Figure 1 f rom Volume 1 of the COBRA-TRAC manual (5) ,[

a. , C, e Apperidix K Requirement:

B. SWELLING AND RUPTURE OF THE CLADDING AND FUEL ROD THERMAL PARAMETERS Each evaluation model shall include a provision for predicting cladding swelling and rupture from consideration of the axial temperature distribution of the cladding and from the difference in

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pressure betwecq the inside a.d cutt'de of tFa cladding, both as f unctic:.. of tic.e. Tc De acceptable toe swe;iir.g and rupture calculations shall be based on applicable data in,such a way that the degree of swelling and incidence of rupture are not underestimated.

The degree of swelling and rupture shall be taken into account in calculations of gap conductance, cladding oxidation and embrittlement, and hydrogen generation.

The calculations of fuel and cladding temperatures as a function of time shall use values for gap conductance and other thermal parameters as functions of temperature and other applicable time-dependent variables. The gap conductance shall be varied in accordance with changes in gap dimension and any other applicable variables.

l I

7889B:10/111385 11-5

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Attachmant 1 l Page 7 of 25 l

l Method of Compliance:

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__ , o_, C ,4 C. BLOWDOWN PHENOMENA ADDendiX K Requirement:

1. Break r.haracteristics and Flow.

a. In analyses of hypothetical loss-of-coolant accidents, a spectrum of possible pipe break shall be considered. This spectrum shall include instantaneous double-ended breaks ranging in cross-sectional area up to and including that of the largest pipe in the primary coolant system. The analysis shall also include the effects of longitudinal splits in the largest pipes, with the split area equal to the cross-sectional area of the pipe.

Method of Compliance:

a.,c,e Appendix K Reouirement:

b. Discharge Model. For all times after the discharging fluid has been calculated to be two-phase in composition, the discharge rate shall be calculated by use of the Moody model (F. J. Moody, " Maximum Flow Rate of a Single Component, Two-Phase Mixture," Journal of Heat Transfer, Trans American Society of Mechanical Engineers, 87, No.1, February,1965). The calculation shall be conducted with at least three values of a discharge coefficient applied to the postulated break area, these values spanning the range f rom 0.6 to 1.0. If the results indicate that the maximum clad temperature for the hypothetical accident is to be found at an even lower value of the discharge coefficient, the range of discharge coefficients 7889B:10/111385 11-7

Attachment 1 Page 8 of 25 shall be extended until the maximum clad temperature calculated by this variation has been achieved.

l Method of Comoliance:

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ADoendix K Reauirement:

c. End of Blowdown. (Applies Only to Pressurized Water Reactors.) For postulated cold leg breaks, all emergency cooling water injected into the inlet lines or the reactor vessel during the bypass period shall in the calculations be subtracted from the reactor vessel calculated inventory.

This may be executed in the calculation during the bypass period, or as an alternative the amount of emergency core cooling water calculated to be injected during the bypass period may be subtracted later in the calculation from the water remaining in the inlet lines, downcomer, and reactor vessel lower plenum af ter the bypass period. This bypassing shall end in the calculation at a time designated as the "end of bypass," after which the expulsion or entrainment mechanisms responsible for the bypassing are calculated not to be effective. The end-of-bypass definition used in the calculation shall be justified by a suitable combination of analysis and experimental data. Acceptable methods for defining "end-of-bypass" include, but are not limited to, the following:

(1) Prediction of the blowdown calculation of downward flow in the downcomer for the remainder of the blowdown period; (2) Prediction of a threshold for droplet entrainment in the upward velocity, using local fluid conditions and a conservative critical Weber number.

Method of Compliance:

oL,C,R.

7889B:10/111385 11-8

. s Attachment 1 Page 9 of 25

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Appendix K Reauirement:

d. Noding Near the Break and the ECCS Injection Points. The noding in the vicinity of and including the broken or split sections of pipe and the points of ECCS injection shall be chosen to permit a reliable analysis of the thermodynamic history in these regions during blowdown.

Method of Compliance:

_ C , e.

The break model in TRAC-P02 has been verified against the Edwards pipe critical flow problem.

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s7889B:10/111385 11-9

Attachmtnt 1 Page 10 of 25 ADDendix K Requirement:

2. Frictional Pressure Drops. The frictional losses in pipes and other components including the reactor core shall be calculated using models that include realistic variation of friction factor with Reynolds number, and realistic two-phase friction multipliers that have been adequately verified by comparison with experimental data, or models that prove at least equally conservative with respect to maximum clad temperature calculated during the hypothetical accide.11. The modified Baroczy correlation for Two-Phase Pressure Drop," Chem. Enging. Prog. Symp.

Series, No. 64, Vol. 62,1965) or combination of the Thom correlation (Thom, J. R. S., " Prediction of Pressure Drop During Forced Circulation Boiling of Water," Int. J. of Heat & Mass Transfer, 7, 709-724, 1964) for pressures equal to or greater than 250 psia and the Martinelli-Nelson correlation (Martinelli, R. C. Nelson, D. B., " Prediction of' Pressure Dro During Forced Circulation Boiling of Water," Transactions of ASME, 695-702,1948) for pressures lower than 250 psia is acceptable as a basis for calculating realistic two-phase friction multipliers.

Method of ComDliance:

_. . a.,

4 ADDendix K Requirement:

3. Momentum Equation. The following effects shall be taken into account in the conservation of momentum equation: (1) temporal change of momentum, (2) momentum convection, (3) area change momentum flux, (4) momentum change due to compressibility, (5) pressure loss resulting from wall friction, (6) pressure loss resulting from area change, and (7) gravitational acceleration. Any omission of one or more of these terms under stated circumstances shall be justified by comparative analyses or by experimental data. ,

Method of Compliance:

There are full momentum equations in the governing set of equations for COBRA-TRAC which include, temporal change of momentum, convections of momentum, area changes which give momentum flux terms, compressibility 7889B:10/111385 11-10

Attachment 1 Page 11 of 25 effects, frictional losses, form losses due to area changes or geometry changes, interfacial shear effects and elevation pressure losses.

Anoendix K Reauirement:

4. Critical Heat Flux.

a. Correlations developed from appropriate steady-state and transient-state experimental data are acceptable for use in predicting the critical heat flux (CHF) during LOCA transients. The computer programs in which these correlations are used shall contain suitable checks to assure that the physical parameters are within the range of parameters specified for use of the correlations by their respective authors.

b. Steady-state CHF correlations acceptable for use in LOCA t.ansients include, but are not limited to, the following:

(1) W 3. L. S. Tong, " Prediction of Departure from Nucleate Boiling for an Axially Non-uniform Heat Flux Distribution," Journal of Nuclear Energy, Vol. 21, 241-248, 1967.

(2) B&W-2. J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J. Stanek, " Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water," Two-Phase Flow and Heat Transfer in Rod Bundles, ASME, New York, 1969.

(3) Hench-Levy. J. M. Healzer, J. E. Hench, E. Janssen, S. Levy, " Design Basis for Critical Heat Flux Condition in Boiling Water Reactors,"

APED-5186, GE Company Private report, July 1966.

(4) Macbeth. R. V. Macbeth, "An Appraisal of Forced Convection Burnout Data," Proceedings of the Institute of Mechanical Engineers, 1965-1966.

(5) Barnett. P. F. Barnett, " A Correlation of Burnout Data for Uniformerly Heated Annuli and Its Uses for Predi.cting Burnout in Uniformly Heated Rod Bundles," AEEW-R 463, 1966.

(6) Hughes. E. D. Hughes, "A Correlation of Rou dundle Critical Heat Flux for Water in the Pressure Range 150 to 725 psia," IN-1412, Idaho Nuclear Corporation, July 1970.

c. Correlations of appropriate transient CHF data may be accepted for use in LOCA transient analyses if comparisons between the data and the correlations are provided to demonstrate that the correlations predict values of CHF which allow for uncertainty in the experimental data throughout the range of parameters for which the correlations are to be used. Where appropriate, the comparisons shall use statistical uncertainty analysis of the data to demonstrate the conservatism of the transient correlation.

7889B:10/111385 11-11

Attachmsnt 1 Paga 12 of 25 Method of ComDliance:

COBRA-TRAC uses Griffith's(8) modification of the ZuberI9) pool boiling critical heat flux correlation for pool boiling situations and the Biasi(10) correlation for forced-convection critical heat flux. The Biasi correlation consists of two equations, one for low quality forced convection flow and one correlationforhighquality.{

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ADoendix K Reauirement:

d. Transient CHF correlations acceptable for use in LOCA transients include, but are not limited to, the following:

(1) GE transient CHF. B. C. Slifer, J. E. Hench, " Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors," NEDO-10329, General Electric Company, Equation C-32, April 1971.

e. Af ter CHF is first predicted at an axial fuel rod location during blowdown, the calculation shall not use nucleate boiling heat transfer correlations at that location subsequently during the blowdown even if the calculated local fluid and surface conditions would apparently justify the reestablishment of nucleate boiling. Heat transfer assumptions characteristic of return to nucleate boiling (rewetting) shall be permitted when justified by the calculated local fluid and surface conditions during the reflood portion of a LOCA.

Method of Compliance:

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ADDendix K Reauirement:

5. Post-CHF Heat Transfer Correlations.

a. Correlations of heat transfer from the fuel cladding to the surrounding fluid in the post-CHF regimes of transition and film boiling shall be compared to applicable steady-state and transient-state data using statistical correlation and uncertainty analyses. Such comparison shall demonstrate that the correlations predict values of heat transfer coefficient equal to or less than the mean value of the applicable experimental heat transfer data 78898:10/111385 11-12

Attachmznt 1 Page 13 of 25 i

throughout the range of parameters for which the correlations are to '

be used. The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the applicable data.

b. The Groeneveld flow film boiling correlation (equation 5.7 of D. C.

Groeneveld, "An Investigation of Heat Transfer in the Liquid Deficient Regime," AECL-3281, revised December 1969), the Dougall-Rohsenow flow film boiling correlation (R. S. Dougall and W. M. Rohsenow, " Film Boiling on the Inside of Vertical Tubes with Upward Flow of the Fluid at Low Qualities,: MIT Report Number 9079-26 Cambridge, Massachusetts, September 1963), and the Westinghouse correlation of steady-state transition boiling (" Proprietary Redirect / Rebuttal Testimony of Westinghouse Electric Corporation," U. S. A. E. Docket RM-50-1, page 25-1, October 26,1972) are acceptable for use in the post-CHF boiling regimes.

In addition the transition boiling correlation of McDonough Milich, and King (J. B. McDonough, W. Milich, E. C. King, " Partial Film Boiling with Water at 2000 psig in a Round Vertical Tube," MSA Research Corp.,

Technical Report 62 (NP-6976), (1958)) is suitable for use between nucleate and film boiling. Use of all these correlations shall be restricted as follows:

(1) The Groeneveld correlation shall not be used in the region near its low-pressure singularity.

(2) the first term (nucleate) of the Westinghouse correlation and the i

entire McDonough, Milich, and King correlation shall not be used during the blowdown after the temperature difference the clad and the saturated fluid first exceeds 300*F, (3) transition boiling heat transfer shall not be reapplied for the remainder of the LOCA blowdown, even if the clad superheat returns below 300*F, except for the reflood portion of the LOCA when justified by the calculated local fluid and surface conditions.

Method of Compliance:

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7889B:1D/111385 11-13

Attachmant 1 ,

Page 14 of 25 l

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Anoendix K Reauirement:

6. Pump Modeling. The characteristics of rotating primary system pumps (axial flow turbine, or centrifugal) shall be derived from a dynamic model that includes momentum transfer between the fluid and the rotating member, with variable pump speed as a function of time. The pump model resistance used for analysis should be justified. The pump model for the two-phase region shall be verified by applicable two-phase pump performance data.

For BWR's af ter saturation is calculated at the pump suction, the pump head may be assumed to vary linearly with quality, going to zero for one percent quality at the pump suction, so long as the analysis shows that core flow stops before the quality at pump suction reaches one percent.

Method of Compliance:

. ct ,

Appendix K Reauirement:

7. Core Flow Distribution During Blowdown. (Applies only to pressurized water reactors.)

a. The ' low rate through the hot region of the core during blowdown shall be calculated as a function of time. For the purpose of these calculations the hot region chosen shall not be greater than the size of one fuel assembly. Calculations of average flow and flow in the hot region shall take into account cross flow between regions and any flow blockage calculated to occur during blowdown as a result of cladding swelling or rupture. The calculated flow shall be smoothed to eliminate any calculated rapid oscillations (pericd less than 0.1 seconds).

Method of Compliance:

cu, 7889B:10/111385 11-14

Attachment 1 Page 15 of 25 ADDendix K Reauirement:

b. A method shall be specified for determining the enthalpy to be used as input data to the hot channel heatup analysis from quantities calculated in the blowdown analysis, consistent with the flow distribution calculations.

Method of Comoliance:

ou, ADDendix K Recuirement:

D. POST-BLOWDOWON PHENOMENA; HEAT REMOVAL BY THE ECCS

1. Single Failure Criterion. An analysis of possible failure modes of ECCS equipment and of their effects on ECCS performance must be made. In carrying out the accident evaluation the combination of ECCS subsystems assumed to be oDerative shall be there as=ilable a' er the most damaging single f ailurt of ECC5 ect pn.er. nas taken place.

Method of Compliance:

An analysis of possible failure modes has been made in WCAP-8340 EIII (Section 17.0) and WCAP-8471 E3(Section3.6).]

A,c 7889B:10/111385 11-15

Attachm:nt 1 Pagt 16 of 25 Anoendix K Reauirement:

2. Containment Pressure. The containment pressure used for evaluating cooling effectiveness during reflood and spray cooling shall not exceed a pressure calculated conservatively for this purpose. The calculation shall include the effects of operation of all installed pressure-reducing systems and processes.

Method of ComDliance:

-- --. a.,C.

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ADDendix K Requirement:

3. Calculation of Reflood Rate for Pressurized Water Reactors. The refilling of the reactor vessel and the time and rate of reflooding of the core shall be calculated by an acceptable model that takes into consideration the thermal and hydraulic characteristics of the core and of the reactor system.

Method of ComDliance:

a,t MM g mee.o e

Attachm nt i Page 17 of 25

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ADDendix K Requirement:

The primary system coolant pumps shall be assumed to have locked impellers if this assumption leads to the maximum calculated cladding temperature; otherwise the pump rotor shall be assumcd to be running free.

Method of ComD11ance:

a., c ADDendiX K Requirement:

The ratio of the total fluid flow at the core exit plane to the total liquid flow at the core inlet plane (carryover fraction) shall be used to determine the core exit flow and shall be determined in accordance with applicable experimental 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 Transfer (FLECHT) Group I Test Report," Westinghouse Report WCAP-7435, January 1970; "PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Group II Test Report," Westinghouse Report WCAP-7544, September 1970; "PWR FLECHT Final Report Supplement,"

Westinghouse Report WCAP-7931, October 1972).

Method of ComD11ance:

This specific requirement was written for a conventional PWR reflooding transient. - ' - " "

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7889B:10/111385 11-17

Attachminc 1 Pags 18 of 25 ADDendiX K Requirement:

The effects on reflooding rate of the compressed gas in the accumulator which is discharged following accumulator water discharge shall also be taken into account."

Method of Compliance:

d. , c.

Appendix K Recuirement:

4. Steam Interaction with Emeraency Core Coolina Water in Pressurized Water Reactors. The thermal-hydraulic interaction between steam and all emergency core cooling water shall be taken into account in calculating the core reflooding rate.

Method of Compliance:

ct,c.

Appendix K Reauirement:

During refill and reflood, the calculated steam flow in unbroken reactor coolant pipes shall be taken to be zero during the time that accumulators are discharging water into those pipes unless experimental evidence is available regarding the realistic thermal-hydraulic interaction between the steam and the liquid. In this case, the experimental data may be used to support an alternate assumption.

Method of Compliance:

The accumulator flow-steam flow interaction is calculated in directly from the conservation equations. The models for interfacial shear 18898:10/111385 11-18

Attachment 1 Page 19 of 25 and heat transfer have been verified against scaled experiments which had the cold leg interaction such as the Battelle 2/15 scale downcomer ECC bypess tests, Creare 1/15 scale downcomer ECC bypass tests, and the(~_

,- a.,c.

Annendix K Reauirmnent:

5. Refill and Reflood Heat Transfer for Pressurized Water Reactors.

For reflood rates of one inch per second or higher, reflood heat 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" Westinghouse Report WCAP-7665, 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) Group I Test Report," Westinghouse Report WCAP-7544, September 1970; "PWR FLECHT Final Report Supplement,"

Westinghouse Report WCAP-7931, October 1972) are not acceptable. New correlations or modifications 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 swelling or rupture as such blockage might affect both local steam flow and heat transfer.

l Method of Compliance:

__ ot, c_

l 78898:10/111385 11-19

Attachm:nt 1 Page 20 of 25

o. , C.

validate the applicability of UPI situations at low flows.

ADrendix K Requirement:

6. Convective Heat Transfer Coefficients for Boilina Water Reactor Full Rods Under Soray Coolina.

Method of ComDliance:

These requirements are not applicable to PWR's.

Appendix K Reauirement:

7. The Boilina Water Reactor Channel Box Under SDray Coolina.

Method of Compliance:

These requirements are not applicable to PWR's.

ADDendix K Reauirement:

II. REQUIRED DOCUMENTATION 1.a. A description of each evaluation model shall be furnished. The description shall be sufficiently complete to permit technical review of the analytical approach including the equations used, their approximations in dif f erence form, the assumptions made, and the values of all parameters

. or the procedure for their selection, as for example, in accordance with a specified physical law or empirical correlation.

Method of Compliance:

For the purposes of this review the primary description is contained in this ct,c, ADD ndix K Recuirement:

b. The description shall be sufficiently detailed and specific to require significant changes in the evaluation model to be specified in amendments of the description. For this purpose, a significant change is a change that would result in a calculated fuel cladding 7889B:10/111385 11-20

Attachmint 1 Page 21 of 25 temperature different by more than 20*F from the temperature calculated (as a function of time) for a postulated LOCA using the last previously accepted model."

Method of Compliance:

Procedures established for previous evaluation models will be followed. As described above, any model change which results in a change in peak cladding temperature of more than 20*F will be documented and submitted to NRC for review.

Anoendix K Reauirement:

c. A complete listing of each computer program, in the same form as used in the evaluation model, shall be furnished to the Nuclear Regulatory Commission."

Method of Comoliance:

Procedures established for previous evaluation models will be followed.

Specifically, controlled versions of each code used are maintained at Westinghouse and are available for inspection by NRC at any time.

Anoendix K Reauirement:

2. For each computer program, solution convergence shall be demonstrated by studies of system modeling or noding and calculational time steps."

Method of ComDliance:

Existing sensitivity studies are presented in the CO8RA-TRAC manua',I I which show the effect of varying timestep size and noding. These studies indicats that the timestep and noding values chosen are appropriate and lead to converged solutions.

Appendix K Requirement:

3. Appropriate sensitivity studies shall be performed for each evaluation model, to evaluate the effect on the calculated results of variations in [a] noding, [b] phenomena assumed in the calculation to 7889B:10/111385 11-21 1

r Attachm nt 1 Paga 22 of 25 1

predominate, including pump operation or locking, and [c] values of I parameters over their applicable ranges. For items to which results l are shown to be lensitive, the choices made shall be justified."

l Method of ComDliance:

- goc,C l ADDendix K Reauirement

4. To the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimental information."

Method of Compliance:

- Q.,C, e Appendix K Reauirement: .

5. General Standards for Acceptability - Elements of evaluation models reviewed will includi technical adequacy of the calculational methods, including compliance with required features of Section I of this Appendix K and provision of a level of safety and margin of conservatism comparable to other acceptable evaluation models, taking into account significant differences in the reactors to which they apply."

78898:10/111385 11-22

k

. .e 4 Attachment 1 Paga 23 of 25 Method of Como11ance: ,

l

_ a.,c,e i

4 78898:10/111385 11-23

Attachm:nt 1 Pagt 24 of 25 REFERENCES

1. Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8302, June 1974.

2. Bordelon, F. M., et al, The Westinghouse ECCS Evaluation Model:

Supplementary Information," WCAP-8471-P-A (Proprietary Version), WCAP-8472 (Non-Proprietary Version), January 1975.

3. Bordelon, F. M. et al, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974.

4. Young, M. Y. et al, 'BART: A Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary Version), WCAP-9695-A (Non-Proprietary Version), March 1984.

5. Thurgood, M. J., at 11 " COBRA / TRAC - A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems" NUREG-CR-3046, November 1982.

6. Liles, D. R., et al, " TRAC-PD2, On Advanced Best-Estimate Computer Program for Pressurized Water Reactor loss of Coolant Accident Analysis" NUREG-CR-2054, 1981.

7. Westinghouse Nuclear Safety, " Westinghouse Evaluation Model: 1981 Version," WCAP 9220-P-A, 1982.

8. Bjornaad, T. A., and P. Grif fith, "PWR Blowdown Heat Transf er," Thermal and Hydraulic AsDects of Nuclear Reactor Safety, Vol 1, pg 17-41, 1977, Published by ASHE.

9. Zuber, N. et 31, "The Hydodynamic Cruis in Pool Boiling of saturated and Subcooled liquids," Part II, No. 27, International Developments in Heat Transfer, International Heat Transfer Conference, Boulder Calorado,1961.

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Attachmint 1 Page 25 of 25

10. Biasi, h., gt 11. " Studies on Burnont, Part 3," Energia Nucleare,11. No.

9, 530-536, (1967).

11. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Version), July 1974,

12. Iguchi, T, it 11. " Data Report on Large Scale Reflood Test-96, CCTF Core II Test C2-13 (Run 72)", JAERI Memo 60-157 (July 1985).

13. Iguchi, T, it 11. " Data Report on large Scale Reflood Test-99, CCTF Core II Test C2-16 (Run 76)," JAERI Memo 60-158 (February 1985).

14. Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprittary Version), June 1974.

15. Colenbrander, H. G. C., Grimm, N. P., "Long Term Ice Condenser Containment Code-LOTIC Code," WCAP-8354 (Proprietary Version), WCAP-8355 (Non-Proprietary Version), July 1974.

16. Kelly, R. D., et al, " Calculational Model for Core Reflooding af ter a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), June 1974.

17. Hochreiter, L. E., et al. " Westinghouse Large Break LOCA Best Estimate Methodology, Volume I Model Description and Validation," WCAP-10924P December 1985.

18. NRC Reactor Systems Branch, " Emergency Core Cooling System Analysis Methods, SECY-83-472 (November 1983).

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