ML20206B191
| ML20206B191 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1987 |
| From: | Reyes J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
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| Shared Package | |
| ML20206A796 | List: |
| References | |
| RTR-REGGD-1.157, TASK-RE, TASK-RS-701-4 REGGD-01.XXX, REGGD-1.XXX, NUDOCS 8704090011 | |
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U.S. NUCLEAR REGULATORY COMMISSION f
n %,{ OFFICE OF NUCLEAR REGULATORY RESEARCH r 87 Task RS 701-4
\ ) DRAFT REGULATORY GUIDE AND VALUE/ IMPACT STATEMENT D
Contact:
J. N. Reves (3011441-7A4n BEST-ESTIMATE CALCULATIONS OF EMERGENCY CORE COOLING SYSTEM PERFORMANCE A. INTRODUCTION Section 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," of 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," requires that light water nuclear reactors fueled with uranium oxide pellets within cylindrical zircaloy cladding be provided with emergency core cooling systems (ECCS) that e.re designed such that their calculated core cooling performance during a loss-of-coolant accident (LOCA) ~
conforms to acceptance criteria specified in paragraph (b) of S 50.46. Para-graph 50.46(b)(1) requires that the maximum calculated fuel element cladding temperature not be greater than 2200 F. In addition, paragraphs (b)(2) through (b)(5) of S 50.46 contain required limits for calculated maximum cladding oxida-tion and maximum hydrogen generation, require that calculated changes in core D geometry remain amenable to cooling, and require that long-term decay heat removal be provided.
The NRC staff has issued proposed amendments to S 50.46 and Appendix K of 10 CFR Part 50 to reflect the improved understanding of ECCS performance during reactor transients obtained through the extensive research performed since the promulgation of the requirements in January 1974. Paragraph 50.46(a)(1) would l
l permit licensees to use either Appendix K features or a realistic 1 evaluation model.
l l
2For the purposes of this guide, the terms "best-estimate" and " realistic" have the same meaning. The term "best-estimate" will be used to indicate that the techniques attempt to predict realistic reactor system thermal-hydraulic response and do not include artificial assumptions to provide a conservative bias.
This regulatory guide and the associated value/ impact statement are being tssued in draft fore to involve the public in the early stages of the development of a regulatory position in this area. They have not received complete staff review and do not represent an official NRC staff polition.
Pubite comunents are being solletted on both drafts, the guide (including any implementation schedule) and the value/ impact statement. Connents on the value/ impact statement should be accompanied by supporting data. Written comments may be sztattted to the Rules and procedures Branch. DDR. ADif U.S. Nuclear Regulatory Conselssion. Washington. DC 20555. Comments may also be delivered to Room 4000. Nryland National Bank Bu11 din 2 7735 Old c.corgetown Road. Bethesda. Nryland from 8:15 a.m. to 5:00 p.m. Copf es of comunents received may be enaatned at the PRC Public Document Room.1717 H 5treet NW.,
Washington. DC. Comuments will be most helpful if received by July 1, 1987.
Requests for single copies of draft guides (which may be reproduced) or for placement on an automatic distribution list for single copies of future draft guides in specific divisions should be nyde in writing to the U.S. Nuclear Regulatory Commission, Washington. DC 20555. Attention: Director. Division of Technical Information and Document Control.
8704090011 870331 PDR REGGD 01.XXX R PDR
These evaluation models2 must include sufficient supporting justification to demon-strate that the analytic techniques realistically describe the behavior of the reactor system during a postulated loss-of-coolant accident. Paragraph 50.46(a)(1) also requires that the uncertainty in the realistic evaluation model be quantified and considered when comparing the results of the calculations with the applicable limits in paragraph 50.46(b) so that there is a high probability that the criteria would not be exceeded.
This draft regulatory guide describes models,3 correlations,4 data, model evaluation procedures, and methods that are acceptable to the NRC staff for meeting the requirements for a realistic or best-estimate calculation of ECCS performance during a loss-of-coolant accident and for estimating the uncertainty in that calculation. Methods for incluaing the uncertainty in the comparisons of the calculational results to the criteria of paragraph 50.46(b), in order to meet the requirement that there be a high probability that the criteria would not be exceeded, are also described in this regulatory guide. Paragraph (a) of 6 50.46 also permits licensees to use evaluation models developed in conformance with Appendix K.
Other models, data, model evaluation procedures, and methods will be considered if they are supported by appropriate experimental data and technical I justification. Any models, data, model evaluation procedures, and methods listed as acceptable in this regulatory guide are acceptable in a generic sense only and would still have to be justified to the NRC staff as being appropriately applied and applicable for particular plant applications.
Appendix A to this draft guide lists models, correlations, data, and model evaluation procedures that the NRC staff considers acceptable for realistic calculations of ECCS performance.
Appendix B to this draft guide describes the acceptable features of best-estimate computer codes and acceptable methods for determining the uncertainty in the calculations.
- The term " evaluation model" refers to a nuclear plant systems computer code or any other analysis tool designed to predict the aggregate behavior of a loss-of-coolant accident. It can be either best-estimate or conservative and may contain many correlations or models.
3The term "model" refers to a set of equations derived from fundamental physical laws and designed to predict the details of a specific phenomenon.
'The term " correlation" refers to an equation having empirically determined constants such that it can predict some details of a specific phenomena for a limited range of conditions. l 2
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Any information collection activities mentioned in this draft regulatory D
l guide are contained as requirements in 10 CFR Part 50, which provides the regulatory basis for this guide. The information collection requirements in 10 CFR Part 50 have been cleared under OMB Clearance No. 3150-0011.
B. DISCUSSION The criteria set forth in S 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," of 10 CFR Part 50 and the calculational methods specified in Appendix K were formally issued in January 1974 after extensive rulemaking hearings and were based on the understanding of ECCS performance available at that time. In the years since those rules were promul-gated, the NRC, the nuclear industry, and several foreign institutions have conducted an extensive research program that has greatly improved the understanding of ECCS performance during a postulated loss-of-coolant accident. The methods specified in Appendix K are now known to be highly conservative; that is, the fuel cladding temperatures expected during a loss-of-coolant accident would be much less than the temperatures calculated using Appendix K methods. In addition to D
l showing that Appendix K is conservative, the ECCS research allows for quantifica-tion of that conservatism. The results of experiments, computer code development, and code assessment now allow more accurate calculations of ECCS performance j during a postulated loss-of-coolant accident, along with reasonable estimates of l uncertainty, than is possible using the Appendix K procedures.
l It is also known that some plants are being restricted in operating flexi-l bility by limits resulting from conservative Appendix K requirements. These l restrictions are preventing optimal operation of some plants. Based on the research performed, it is now known that these restrictions can be relaxed without adversely affecting safety through the use of more realistic calculations. The l Appendix K requirements divert both NRC and industry resources from matters that are more relevant to reactor safety to analyses with assumptions known to be overly conservative.
In recognition of the known conservatisms in Appendix K, the NRC adopted an interim approach, described in SECY-83-472,5 to accommodate industry requests 6Information Report from William J. Dircks to the NRC Commissioners, dated November 17, 1983, " Emergency Core Cooling System Analysis Methods," SECY-83-472.
D Available for inspection or copying for a fee in the NRC Public Document Room, 1717 H Street NW., Washington, DC.
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for improved evaluation models that would reduce reactor operating restrictions.
This interim approach was a step toward basing licensing decisions on realistic 1
calculations of plant behavior. Although the approach permits many best-estimate methods and models to be used for licensee submittals, it retains those features of Appendix K that are legal requirements.
The NRC staff has been consistently tending toward the rationale that safety is best served when decisions concerning the limits within which nuclear reactors are permitted to operate are based upon realistic calculations. This approach is currently being used in the resolution of almost all reactor safety issues (e.g. ,
anticipated transients without scram, pressurized thermal shock, operator guide-lines) and is now being applied to one of the last remaining major issues still treated in a prescriptive manner, the loss-of-coolant accident.
The proposed amendment to S 50.46 permits ECCS evaluation models to be based on best estimates and removes the arbitrary conservatisms contained in the required features of Appendix K for those licensees wishing to avail themselves of these improved methods. Because of the greatly improved knowledge of ECCS performance during loss-of-coolant accidents, best-estimate calculations with reasonable and quantifiable uncertainties can be performed. The NRC staff is proposing to amend S 50.46 of 10 CFR Part 50 to allow best-estimate, realistic methods to be used for the ECCS performance calculations in place of the evaluation models that use the required Appendix K features. This rule change would also require analyses of the uncertainty of the best-estimate calculations and require that this uncertainty be considered when comparing the results of the calculations to the limits of para-graph 50.46(b) so that there is a high probability that the criteria would not be exceeded. In this manner, more realistic calculations are available for regulatory decisions, yet appropriate conservatism will be maintained consistent with the accuracy of the calculation.
Many of the methods and models needed for a best-estimate calculation are the same as those used previously in ECCS analyses and, although licensees and applicants are well acquainted with them, explicit guidance on acceptable methods and models (based on NRC experience with its own best-estimate advanced codes such as TRAC-PWR, TRAC-BWR, RELAPS, COBRA, and FRAP) is both possible and appropriate.
Further, acceptable methods for the uncertainty analysis have been documented only in SECY-83-472. Therefore, the NRC staff decided that guidance in the form of a regulatory guide would be useful in order to document procedures for calculating ECCS performance that would be acceptable to the regulatory staff.
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l C. REGULATORY POSITION The features presented in Appendix B to this regulatory guide, " Description of Best-Estimate Calculations of Emergency Core Cooling System Performance and Estimation of the Calculational Uncertainty," would be acceptable to the NRC staff for demonstrating compliance with the proposed amendment to 10 CFR Part 50, paragraph 50.46(a)(1).
D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.
This draft guide has been released to encourage public participation in its development.
Licensees and applicants may propose means other than those specified in Section C of this guide for meeting applicable regulations. The guidanct provided in Section C would be used by the NRC staff in evaluating submittals in the following categories as an acceptable means of complying with the D
1 Commission's proposed regulations specified in Section A:
- 1. Applications for construction permits that make use of the provisions of the rule that allow the use of realistic models as an alternative j
to the features of Appendix K of 10 CFR Part 50.
- 2. Applications for operating licenses that make use of the provisions of the rule that allow the use of realistic models as an alternative l to the features of Appendix K of 10 CFR Part 50.
3 Operating reactor licensees will not be evaluated against the provi-sions of this guide except for their new submittals that make use of the provisions of the rule that allow the use of realistic models as an alternative to the features of Appendix K of 10 CFR Part 50.
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D APPENDIX A N0DELS, DATA, AND MODEL EVALUATION PROCEDURES CONSIDERED ACCEPTABLE FOR REALISTIC ECCS CALCULATIONS The NRC staff is soliciting public comments on the necessity and desir-ability of identifying specific models, data, and model evaluation procedures for fuels and thermal-hydraulic behavior in the final regulatory guide. The NRC staff would consider these models, data, and model evaluation procedures to be acceptable for realistic evaluations of ECCS performance provided their appli-cability to the specific use is demonstrated. This appendix lists specific models, data, and model evaluation procedures currently under consideration by the NRC staff. This list is not complete in that it does not represent all the models required to perform a calculation of ECCS performance or all the data that can be used to evaluate the models. It would not be practical to com-pile such a list because many models and data depend on the specific design of the plant and the particular transient to be calculated. Therefore, this appen-D dix is generally restricted to the realistic prediction of the physical proc-I esses identified in Appendix K to 10 CFR Part 50. The NRC staff is still evaluating the models, data, and model evaluation procedures listed in this l appendix in regard to their range of applicability and accuracy. Furthermore, licensees would not be limited to using only the models, data, and model evalua-tion procedures identified in this regulatory guide.
The first part of this appendix lists models, data, and model evaluation procedures that the NRC staff considers acceptable on initial stored energy in the fuel, materials properties, decay heat, and zircaloy oxidation. The second part of this appendix lists thermal-hydraulic model evaluation procedures and data pertaining to critical flow, ECC bypass, frictional pressure drop, post-CHF heat transfer, uncovered bundle heat transfer, and level swell that the NRC staff considers acceptable. Each section identifies specific capabilities that a model must have in order to perform realistic calculations and recommends a suitable data base for model evaluation. Some of the thermal-hydraulic sections identify models that have been used successfully to predict portions of the recommended data base. These models serve as examples and would have to be F assessed using the model evaluation procedures before the NRC staff would l
consider them acceptable.
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- 1. FUEL MATERIALS AND DECAY HEAT MODELS, DATA, AND DATA EVALUATION PROCEDURES
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, 1.1 Fuel and Cladding Material Properties for Use in Calculating Initial Stored Energy in the Fuel The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model used to calculate stored energy and heat transfer in fuel rods (Section 1.2.3.1 in Appendix B).
1.1.1 Model Evaluation Procedure A model to be used in ECCS evaluations to calculate internal fuel rod heat transfer should:
- be checked against several sets of relevant data recognize the effects of fuel burnup, fuel pellet cracking, fuel pellet relocation, cladding creep, and gas mixture conductivity.
The model described by Lanning (Ref. A-1) compared well to inpile fuel 4
temperature data.
1.1.2 Experimental Data The correlations and data of Reference A-2 would be considered acceptable for calculating the initial stored energy of the fuel and subsequent heat transfer.
1.2 Fission Product Decay Heat The model of Reference A-3 would be considered acceptable for calculating fission product decay heat (Sections 1.2.3.3 and 1.2.3.4 of Appendix B).
1.2.1 Model Evaluation Procedure The values of mean energy per fission (Q) and the models for actinide decay heat should be checked against a set of relevant data.
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- 1. 3 Metal-Water Reaction Rate The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model used to calculate the rates of energy release, hydrogen generation, and cladding oxidation (Section 1.2.3.5 of Appendix B).
1.3.1 Model Evaluation Procedure Correlations to be used to calculate metal-water reaction rates for the temperature range 1500*F - 1900 F should:
- be checked against a set of relevant data
- recognize the effects of steam pressure, preoxidation of the cladding, deformation during oxidation, and internal oxidation from both steam and U02 fuel.
The data of References A-4 and A-5 would be considered acceptable for
) calculating the rates of energy release, hydrogen generation, and cladding oxidation for cladding temperatures greater than 1900 F and less than 1500 F, respectively.
- 2. THERMAL-HYDRAULIC MODEL EVALUATION PROCEDURES AND DATA 2.1 ECC Bypass The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate ECC bypass during the blowdown phase of a loss-of-coolant accident (Sec-tion 1.2.5.2 of Appendix B).
2.1.1 Model Evaluation Procedure A correlation or model to be used in ECCS evaluations to calculate ECC bypass should:
- be checked against an acceptable set of relevant data
)
- recognize the effects of pressure, liquid subcooling, fluid conditions, hot walls, and system geometry.
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l Uncertainties and bias in the correlations or models used to calculate ECC bypass should be stated, as well as the range of their applicability.
For scaled-down PWR downcomers, correlations by Beckner and Reyes (Ref. A-6) compare well to the bypass data of References A-7 and A-8. Correlations of Sun (Ref. A-9) and Jones (Ref. A-10) compare well to counter-current flow limiting (CCFL) test data of interest to BWRs.
2.1.2 Experimental Data The following tests should be considered in establishing a set of data for scaled-down PWR downcomers:
BCL tests (Ref. A-7)
CREARE tests (Refs. A-7 and A-8)
For a full-scale PWR vessel, ECC bypass data will become available from the forthcoming upper plenum test facility (UPTF) experiments performed as part of the 2D/3D program sponsored by the Federal Republic of Germany, Japan, and the United States.
For BWRs, the following test should be considered in establishing an acceptable set of relevant data:
SSTF test data (Refs. A-11 through A-13) 2.2 Critical Flow The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate the discharge flow rate during a loss-of-coolant accident (Section 1.2.5.1 of Appendix B).
2.2.1 Model Evaluation Procedure Critical flow models to be employed in ECCS evaluations should:
- be checked against an acceptable set of relevant data A-4
- recognize thermal nonequilibrium conditions when the fluid is subcooled
- provide a means of transition from nonequilibrium to equilibrium conditions The uncertainties and bias of a correlation or model used to calculate critical flow should be stated, as well as their range of applicability.
The mechanistic thermal nonequilibrium and slip model of Richter (Ref. A-14) compares well to small- and large-scale test data (Ref. A-15).
2.2.2 Experimental Data An acceptable set of relevant critical flow data should cover the fluid conditions, geometries, and types of breaks pertinent to light water reactor loss-of-coolant accidents. The following tests should be considered in estab-lishing an acceptable set of relevant data:
Marviken tests (Ref. A-16)
Moby Dick experiments (Ref. A-17)
BNL critical flashing flows in nozzles (Ref. A-18)
Sozzi-Sutherland tests (Ref. A-19)
Edwards experiments (Ref. A-20)
Super Moby Dick experiments (Refs. A-21 and A-22)
For critical flow from small breaks under stratified conditions, currently acceptable test data for assessing models and codes include those reported by:
Anderson and Owca (Ref. A-23)
Reimann and Khan (Ref. A-24)
Schrock et al. (Refs. A-25 and A-26)
- 2. 3 Frictional Pressure Drop The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate the frictional pressure drop in pipes and other components (Section 1.2.7 of D Appendix B).
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2.3.1 Model Evaluation Procedure A model for frictional pressure drop to be used in ECCS evaluation should:
- be checked against a set of relevant data
- be consistent with models used for calculating gravitational and acceleration pressure drops. If void fraction models or correlations used to calculate the three components of the total pressure drop differ one from another, a quantitative justification must be provided.
Uncertainties and bias of a correlation or model should be stated as well as the range of applicability.
2.3.2 Experimental Data An acceptable set of relevant data should cover, as far as possible, the ranges of parameters (mass flux, quality, pressure, fluid physical properties, roughness, and geometries) that are found in actual plant applications. The following tests should be considered in establishing an acceptable set of rele-vant data: l Vertical tubes:
CISE tests (Refs. A-27 and A-28)
HTFS data bank
- Horizontal tubes HTFS data bank
- GE tests (Ref. A-29 and A-30)
Rod bundles:
HTFS data bank
- GE tests (Ref. A-31)
- The Heat Transfer Fluid Flow Services (HTFS) Data Bank, Harwell, UK, is already available to subscribers. The NRC is working with Harwell Laboratory to make these data sets available to the public. Contact Jose Reyes, Office of Nuclear Regulatory Research, Washington, DC 20555.
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2.4 Post-CHF Heat Transfer The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate the heat transfer from the fuel to the surrounding fluid while in the post-CHF regime (Section 1.2.10 of Appendix B).
2.4.1 Model Evaluation Procedure A model to be used in ECCS evaluations to calculate post-CHF heat transfer from rod bundles should:
- Be checked against an acceptable set of relevant data e Recognize effects of: liquid entrainment, thermal radiation, thermal nonequilibrium, low and high mass flow rates, low and high power densities, and saturated and subcooled inlet conditions.
The uncertainties and bias of models or correlations used to calculate g
post-CHF heat transfer should be stated as well as the range of their applica-bility.
2.4.2 Experimental Data The acceptable set of relevant data should cover power densities, mass flow rates, fluid conditions, and rod bundle geometries pertinent to LWR designs and applications. The following tests should be considered in establishing an acceptable set of relevant data:
ORNL tests (Refs. A-32 and A-33)
FLECHT SEASET tests (Ref. A-34)
INEL tests (Ref. A-35)
ORNL data bank (Ref. A-36) 2.5 Uncovered Bundle Heat Transfer The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate I
the heat transfer from the fuel to the vapor during periods when the fuel bundles are uncovered (Section 1.2.13.4 of Appendix B).
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2.5.1 Model Evaluation Procedure A correlation to be used in ECCS evaluations to calculate heat transfer ,
from uncovered rod bundles should:
- be checked against an acceptable set of relevant data recognize effects of radiation and of laminar, transition, and turbulent flows.
Uncertainties and bias in the correlation should be stated, as well as the range of applicability.
The correlation derived should include a stated procedure for correcting for radiative heat transfer and for estimating the vapor temperatures. The Hottel procedure cited in Reference A-37 is a satisfactory example.
The turbulent correlation may be of the general form:
Nu = A Re" Pr" for higher Reynolds numbers (Re), where the coefficients A, m, and n are modifications from the basic Dittus-Boelter form and may be functions of other variables. Pr represents the Prandtl number, and Nu is the Nusselt number. The physical properties may be defined as wall, film, or vapor values.
A distinction from, and transition to, laminar convection should be made, with a value of the laminar heat transfer being, for example, for rod bundles:
Nu = 8 for Re < 2000 Other forms and values, depending on the bundle geometry and flow condi- I tions, are also appropriate.
2.5.2 Experimental Data An acceptable set of relevant data should cover power densities, fluid conditions, and rod bundle geometries pertinent to LWR design and application.
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The following tests should be considered in establishing an acceptable set of relevant data:
ORNL-THTF tests (Refs. A-37 and A-38)
FLECHT SEASET tests (Refs. A-39 and A-40)
ORNL data base (Ref. A-36) 2.6 Level Swell The model evaluation procedure and data presented in this section would be considered acceptable for assessing a model or correlation used to calculate the two phase level in the reactor (Section 1.2.13.2 of Appendix B).
2.6.1 Model Evaluation Procedure A correlation or model to be used in ECCS evaluation to calculate level swell should:
- be checked against an acceptable set of relevant data D
- recognize the effects of depressurization, boil-off, power level, fluid conditions and system geometry.
l Uncertainties and bias of a correlation or model used to calculate level swell should be stated as well as the range of applicability.
l The correlation proposed by Chexal, Horowitz, and Lellouche (Ref. A-41) compares well to experimental data reported in References A-37, A-42, A-43, and A-44, 2.6.2 Experimental Data The following tests should be considered in establishing an acceptable set of relevant data:
l GE tests (Refs. A-42 and A-45)
ORNL tests (Refs. A-37 and A-43)
FLECHT-SEASET test (Ref. A-39)
D THETIS tests (Ref. A-44)
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REFERENCES {
Some of these references note where the document may be obtained.
Addresses of these sources are listed at the end of the reference list.
A-1 D. Lanning and M. Cunningham, " Trends in Thermal Calculations for Light Water Reactor Fuel (1971-1981)," in Ninth Water Reactor Safety Research Information Meetina, U.S. Nuclear Regulatory Commission (USNRC),
Washington, DC, NUREG/CP-0024, Vol. 3, March 1982.
A-2 USNRC, "MATPRO Version 11 (Revision 2): A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior,"
Washington, DC, NUREG/CR-0497, Rev. 2 (prepared for NRC by EG&G Idaho),
August 1981.
A-3 American Nuclear Society, "American National Standard for Decay Heat Power in Light Water Reactors," ANSI /ANS-5.1-1979, August 1979. (ANS, 555 North Kensington Avenue, La Grange Park, IL 60525.)
l A-4 J. V. Cathcart et al., " Zirconium Metal-Water Oxidation Kinetics: IV Reaction Rate Studies," Oak Ridge National Laboratory, Oak Ridge, TN, i ORNL/NUREG-17, August 1977. (Available from NTIS.)
A-5 R. R. Biederman et al., "A Study of Zircaloy-4 Steam 0xidation Reaction l
~
Kinetics," EPRI NP-734, Electric Power Research Institute, Palo Alto, CA l (prepared by Worcester Polytechnic Institute, Worcester, Massachusetts),
April 1978.
A-6 W. D. Beckner and J. N. Reyes, Research Information Letter No. 128, "PWR Lower Plenum Refill Research Results," USNRC, December 8, 1981. (Available in the NRC Public Document Room.)
A-7 W. D. Beckner, J. N. Reyes, R. Anderson, " Analysis of ECC Bypass Data,"
USNRC, Washington, DC, NUREG-0573, July 1979.
4 A-10 l
A-8 C. J. Crowley et al., "1/5-Scale Countercurrent Flow Data Presentation and Discussion," USNRC, Washington, DC, NUREG/CR-2106 (prepared by Creare Incorporated, New Hampshire), November 1981.
A-9 K. H. Sun, " Flooding Correlations for BWR Bundle Upper Tieplate and Bottom Side-Entry Orifices," in Multi-Phase Transport: Fundamentals, Reactor Safety, Applications, T. N. Veziroglu, Editor, Hemisphere Publishing Corp. ,1010 Vermont Avenue, Washington, DC, Vol.1,1979.
A-10 D. D. Jones, "Subcooled Counter-Current Flow Limiting Characteristics of the Upper Region of a BWR Fuel Bundle," General Electric Company, NEDG-NUREG-23549, July 1977. (Available in the NRC Public Document Room.)
A-11 J. A. Findlay, "BWR Refill-Reflood Program Task 4.4 - CCFL/ Refill System Effects Tests (30* Sector). Evaluation of Parallel Channel Phenomena,"
USNRC, Washington, DC, NUREG/CR-2566 (prepared for NRC by General Electric Company, GEAP-22044, EPRI NP-2373), November 1982.
A-12 D. G. Schumacher et al., "BWR Refill-Reflood Program Task 4.4 - CCFL/ Refill System Effects Tests (30 Sector). SSTF Systems Response Test Results,"
USNRC, Washington, DC, NUREG/CR-2568 (prepared for NRC by General Electric Company, GEAP-22046, EPRI NP-2374), April 1983.
! A-13 J. A. Findlay, "BWR Refill-Reflood Program Task 4.4 - CCFL/ Refill System Effects Tests (30 Sector). Evaluation of ECCS Mixing Phenomena," USNRC, Washington, DC, NUREG/CR-2786 (prepared for NRC by General Electric Company, GEAP-22150, EPRI NP-2542),'May 1983.
A-14 H. J. Richter, " Separated Two-Phase Flow Model: Application to Critical Two-Phase Flow," Electric Power Research Institute, Palo Alto, CA, EPRI Report NP-1800, April 1981.
A-15 D. Abdo11ahian et al., " Critical Flow Data Review and Analysis," Electric Power Research Institute, Palo Alto, CA, Report NP-2192, January 1982.
A-11
A-16 USNRC, "The Marviken Full Scale Critical Flow Tests, Summary Report" (Joint {
Reactor Safety Experiments in the Marviken Power Station, Sweden),
Washington, DC, NUREG/CR-2671, May 1982.
A-17 M. Reocreux, " Contribution to the Study of Two-Phase Steam-Water Critical Flow," Ph.D. Thesis, L'Universite Scientifique Medicale de Grenoble,1974.
(English translation available from NTIS, LIB /Trans-576.)
A-18 N. Abuaf et al. , "A Study of Nonequilibrium Flashing of Water in a Converging-Diverging Nozzle," USNRC, Washington, DC, NUREG/CR-1864 (prepared for NRC by Brookhaven National Laboratory), Vols.1-2, March 1982.
A-19 G. L. Sozzi and W. A. Sutherland, " Critical Flow of Saturated and Subcooled Water at High Pressure," General Electric Company, GE Report NED0-13418, 1975. (Available in the NRC Public Document Room.)
A-20 A. R. Edwards and T. P. O'Brien, " Studies of Phenomena Connected with the j Depressurization of Water Reactors," Journal of the British Nuclear Energy Society, Vol. 9, No. 2, April 1970.
A-21 C. Jeandey et al. , " Auto Vaporization d'Ecoulements Eau /Vapeur," Report TT, No. 163, Centre d' Etudes Nucleaires de Grenoble, Grenoble, France, July 1981. (Available in the NRC Public Document Room. The data section of this report is presently in English. The NRC English translation, No. 1861, will be available in the NRC Public Document Room in May 1987.)
A-22 C. Jeandey and L. Gros d'Aillon, " Critical Flows in a Short Super Moby Dick Pipe," Report TT/SETRE/71, Centre d' Etudes Nucleaires de Grenoble, Grenoble, France, September 1983. (NRC Translation 1401 available in the NRC Public Document Room.)
A-23 J. L. Anderson and W. A. Owca, " Data Report for the TPFL Tee / Critical Flow Experiments," USNRC, Washington, DC, NUREG/CR-4164 (prepared by EG&G Idaho, EGG-2377), November 1985.
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I A-24 J. Reimann and M. Khan, " Flow Through a Small Break at the Bottom of a Large Pipe with Stratified Flow," Nuclear Science and Engineering, Vol. 88, pp. 297-310, 1984.
A-25 V. E. Schrock et al. , " Steam-Water Critical Flow Through Small Pipes from Stratified Upstream Regions," in Heat Transfer 1986; C. L. Tien, V. P. Carey, and J. K. Ferrell, Editors, Hemisphere Publishing Corp.,
1010 Vermont Avenue, Washington, DC, Vol. 5, pp. 2307-2311, 1986.
A-26 V. E. Schrock et al., "Small Break Critical Discharge--The Roles of Vapor and Liquid Entrainment in a Stratified Two-Phase Region Upstream of the Break," USNRC, Washington, DC, NUREG/CR-4761 (prepared for NRC by Lawrence Berkeley Laboratory, LBL-22024), October 1986.
A-27 G. P. Gaspari, C. Lombardi, G. Peterlongo, " Pressure Drops in Steam Water Mixtures. Round Tube Vertical Upflow," Centro Informazioni Studi Esperienze, Milan, Italy, CISE-R83, 1964. (Available in the NRC Public Document Room.)
A-28 A. Alessadrini, G. Peterlongo, R. Ravetta, "Large Scale Experiments on Heat Transfer and Hydrodynamics with Steam-Water Mixtures. Critical Heat Flux and Pressure Drop Measurements in Round Vertical Tube at the Pressure of 51 kg/cm2 abs," Centro Informazioni Studi Esperienze, Milan, Italy, CISE-R86, 1963. (Available in the NRC Public Document Room.)
A-29 E. Janssen and J. A. Kervinen, "Two-Phase Pressure Drop Across Contractions and Expansions; Water-Steam Mixtures at 600 to 1400 psia," General Electric Company, GEAP-4622, 1964. (Available in the NRC Public Document Room.)
A-30 E. Janssen and J. A. Kervinen, "Two-Phase Pressure Drop in Straight Pipes and Channels; Water-Steam Mixtures at 600 to 1400 psia," General Electric Co., GEAP-4616, 1964. (Available in the NRC Public Document Room.)
l A-13
A-31 R. T. Lahey, B. S. Shiralkar, D. W. Radcliffe, "Two-Phase Flow and Heat
{
Transfer in Multirod Geometries; Subchannel and Pressure Drop Measurements in a Nine-Rod Bundle for Diabatic and Adiabatic Conditions," General Elec-tric Co., GEAP-13049, 1970. (Available in the NRC Public Document Room.)
A-32 G. L. Yoder et al., " Dispersed Flow Film Boiling in Rod Bundle Geometry -
Steady State Heat Transfer Data and Correlation Comparisons," USNRC, Washington, DC, NUREG/CR-2435 (prepared for NRC by Oak Ridge National Laboratory, ORNL-5822), April 1982.
A-33 D. G. Morris et al., " Dispersed Flow Film Boiling of High Pressure Water in a Rod Bundle," USNRC, Washington, DC, NUREG/CR-2183 (prepared for NRC by Oak Ridge National Laboratory, ORNL/TM-7864), September 1982.
A-34 N. Lee et al., "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Evaluation and Analysis Report," USNRC, Washington, DC, NUREG/CR-2256 (prepared for NRC by Westinghouse Electric Corporation, Westinghouse Report No. 10, EPRI NP-2013), November 1981.
A-35 R. Gottula et al., " Forced Convective, Nonequilibrium, Post-CHF Heat Transfer Experiment Data and Correlation Comparison Report," USNRC, Washington, DC, NUREG/CR-3193 (prepared for NRC by EG&G, Inc. , EGG-2245),
April 1985.
A-36 G. L. Yoder, " Rod Bundle Film Boiling and Steam Cooling Data Base and Correlation Evaluation," USNRC, Washington, DC, NUREG/CR-4394 (prepared for NRC by Oak Ridge National Laboratory, ORNL/TM-9628), August 1986.
A-37 T. M. Anklam et al. , " Experimental Investigations of Uncovered-Bundle Heat Transfer and Two-Phase Mixture Level Swell Under High-Pressure Low Heat-Flux Conditions," USNRC, Washington, DC, NUREG/CR-2456 (prepared for NRC by Oak Ridge National Laboratory, ORNL-5848), April 1982.
A-38 G. L. Yoder et al. , "High Dryout Quality Film Boiling and Steam Cooling Heat Transfer Data from a Rod Bundle," USNRC, Washington, DC, NUREG/CR-3502 (prepared for NRC by Oak Ridge National Laboratory, ORNL/TM-8794), January 1984.
A-14
A-39 S. Wong and L. E. Hochreiter, " Analysis of the FLECHT-SEASET Unblocked D Bundle Steam Cooling and Boiloff Tests," USNRC, Washington, DC, NUREG/CR-1533 (prepared for NRC by Westinghouse Electric Corporation, EPRI NP-1460, WCAP-9729), March 1981.
A-40 H. J. Loftus et al., "PWR FLECHT SEASET 21-Rod Bundle Flow Blockage Task Data and Analysis Report," USNRC, Washington, DC, NUREG/CR-2444, Vols. 1-2 (prepared for NRC by Westinghouse Electric Corporation, Westinghouse Report No. 11, EPRI NP-2014), September 1982.
A-41 B. Chexal, J. Horowitz, and G. Lellouche, "An Assessment of Eight Void Fraction Models for Vertical Flows," NSAC-107, Electric Power Research Institute, Palo Alto, CA, December 1986.
A-42 D. S. Seely and R. Muralidharan, "BWR Low Flow Bundle Uncovery Test and Analysis," USNRC, Washington, DC, NUREG/CR-2231 (prepared for NRC by General Electric Company, GEAP-24964, EPRI NP-1781), April 1982.
A-43 T. M. Anklam, "0RNL Small-Break LOCA Heat Transfer Test Series 1: Two-Phase Mixture Level Swell Results." USNRC, Washington, DC, NUREG/CR-2115 (prepared for NRC by Oak Ridge National Laboratory, ORNL/NUREG/TM-4), September 1981.
A-44 D. Jowitt, "A New Void Correlation for Level Swell Conditions," AEEW-R-1488, Winfrith UK, December 1981. (Available in the NRC Public Document Room.)
A-45 J. A. Findlay, "BWR Refill-Reflood Program Task 4.8 - Model Qualification Task Plan," USNRC, Washington, DC, NUREG/CR-1899 (prepared for NRC by General Electric Company, GEAP-24898, EPRI NP-1527), August 1981.
I A-15
ADDRESSES NUREG- and NUREG/CR-series documents are available from the Government Printing Office (GPO) and the National Technical Information Service (NTIS).
U.S. Government Printing Office National Technical Information Service Post Office Box 37082 Springfield, VA 22161 Washington, DC 20013-7082 Documents that are in the NRC Public Document Room are available for inspection or copying for a fee.
USNRC Public Document Room 1717 H Street NW.
Washington, DC A-16 E _ _ _ _ _ _ _
l APPENDIX B DESCRIPTION OF BEST-ESTIMATE CALCULATIONS OF EMERGENCY CORE COOLING SYSTEM PERFORMANCE AND ESTIMATION OF THE CALCULATIONAL UNCERTAINTY
- 1. BEST-ESTIMATE CALCULATIONS 1.1 General A best-estimate calculation employs modeling that attempts to realistically describe the physical processes in a nuclear reactor. There is no single approach to the extremely complex modeling of the processes occurring during a loss-of-coolant accident. The NRC has developed and performed assessments on several best-estimate advanced thermal-hydraulic transient codes, including TRAC-PWR, TRAC-BWR, RELAP5, COBRA, and the FRAP series of codes (Refs. B-1 through B-6). These codes predict the major phenomena observed for a broad rang.! of theriiial-hydraulic and fuel tests reasonably well. Licensees and appli-cants may use these codes or specific models within them to perform their best-estimate calculations. However, since the NRC staff has not performed the plant-specific uncertainty analysis required by the proposed revision to S 50.46 of 10 CFR Fart 50, the licensee must demonstrate that the codes or models are acceptable and applicable to the specific facility over the intended operating range and must quantify the uncertainty for the specific application. General features expected in a best-estimate calculation are described in this section.
A more detailed discussion of features that are considered ace.eptable for use ir best-estimate codes is given in section 1.2 of this rid.:trix.
A best-estimate model should provide a realist'C sct tion of a particu-lar phenomenon to the degree required or practical w.ch tiie currently available data and knowledge of the phenomenon. The model should be compared with appli-cable experimental data and should predict some mean of the data, rather than providing a bound to the data. The effects of all important variables should be considered. If it is not possible or practical to consider a particular phenomenon, the neglect of this phenomenon should not normally be treated by D including a bias in the analysis directly, but should be included as part of B-1
the model uncertainty. The importance of neglecting a particular phenomenon should be considered within the overall calculational uncertainty.
Careful consideration should be given to the range of applicability of the model when used in a best-estimate code. When comparing the model to the avail-able data, judgments of the applicability of the data to the situation that would actually occur in a reactor should be rtade. Correlations generally should not be extrapolated beyond the range over which they were developed or assessed.
If the model is to be extrapolated beyond the conditions for which valid data comparisons have been made, judgments should be made as to the effect of this extrapolation and should be considered in the uncertainty evaluation. The use of fundamental laws of physics, well-established data bases (e.g., steam tables),
and sensitivity studies should be used in estimating the uncertainty that results from extrapolation.
A best-estimate code contains all the models necessary to predict the important phenomena that might occur during a loss-of-coolant accident. Best-estimate code calculations should be compared with applicable experimental data (e.g. , separate effects tests and integral simulations of loss-of-coolant acci-dents) to determine the overall uncertainty and biases of the calculation. In addition to providing input to the uncertainty evaluation, integral simulation data comparisons should be used to ensure that important phenomena that are ex-pected to occur during a loss-of-coolant accident are adequately calculated.
The following paragraphs list some of the primary features that should be included in best-estimate thermal-hydraulic transient codes. In general, these features will have uncertainties associated with their use for predicting reac-tor system response. These uncertainties should be considered in the overall I
uncertainty analysis described in section 2 of this appendix.
The above discussion is an idealized definition of best-estimate codes.
In practice, best estimate codes may contain certain models that are simplified or contain conservatism to some degree. This conservatism may be introduced for the following reasons:
- 1. The model simplification or conservatism has little effect on the result, and therefore the development of a better model is not justified.
I B-2
- 2. The uncertainty of a particular model is difficult to determine and only an upper bound can be determined.
- 3. The particular application does not require a totally best-estimate calculation, so a bias in the calculation is acceptable.
The introduction of conservative bias or simplification in otherwise best-estimate codes should not, however, result in calculations that are unrealistic, do not include important phenomena, or contain bias and uncertainty that cannot be bounded. Therefore, any calculational procedure determined to be a best-estimate code in the context of this guide or for use under the proposed para-graph 50.46(a)(1)(1) should be compared with applicable experimental data to ensure that realistic calculations of important phenomena result.
1.2. Features of Best-Estimate Codes Some features that are acceptable for use in best-estimate codes are described in the following paragraphs. Models that address these features may be used with the basic requirement that a specific model is acceptable if it has been compared with applicable experimental data and shown to provide reason-able predictions. Reference B-7 provides a summary of the large experimental data base available upon which best-estimate models may be based. While inclu-sion in Reference B-7 does not guarantee that the data or model will be accept-able, the report represents a large body of data generally applicable to best-estimate models. For any models or correlations used in a best-estimate code, sufficient justification must be provided to substantiate that the code performs adequately for the classes of transients to which it is applied. In general, these features have uncertainties associated with their use for predicting reactor system response. These uncertainties should be considered as part of the overall uncertainty analysis described in section 2 of this appendix.
1.2.1 Basic Structure of a Code 1.2.1.1 Numerical Methods. A best-estimate code consists of a numerical scheme for solving the equations used to represent the various models. The D numerical scheme is in itself a complex process that can play an important role B-3
in the overall calculation. Careful numerical modeling, sensitivity studies, and evaluations of numerical error should be performed to ensure that the results of the calculations are representative of the models used in the code. Numer-ical simulations of complex problems such as those considered here treat the geometry of the reactor in an approximate manner, making use of discrete volumes or nodes to represent the system. Sensitivity studies and evaluations of the uncertainty introduced by noding should be performed. Numerical methods treat time in a discrete manner and the effect of time-step size should also be investigated.
- 1. 2.1. 2 Computational Models. A best-estimate code typically contains equations for conservation of mass, energy, and momentum of the reactor coolant and noncondensible gases, if important (e.g., air, nitrogen). Energy equations are also used to calculate the temperature distribution in reactor system struc-tures and the fuel rods. The required complexity of these equations will vary depending on the phenomena that are to be calculated and the required accuracy of the calculation. NRC staff experience with its own best-estimate computer codes has found that separate flow fields for different fluid phases, or types, and calculation of nonequilibrium between phases may be required to calculate some important phenomena (e.g., countercurrent flow, reflood heat transfer) to an acceptable accuracy. The NRC staff has also determined ~that certain phenom-ena require that the equations be solved in multiple dimensions. However, one-dimensional approximations to three-dimensional phenomena will be considered 1
acceptable if those approximations are properly justified. Other basic code l features include equations of state and other material properties. Sensitivity studies and comparisons to data should be performed to determine the importance of the simplifications used.
1.2.2 Initial and Boundary Conditions and Equipment Availability The heat generated by the fuel during a loss-of-coolant accident depends on the power level of the reactor at the time of the loss-of-coolant accident and the history of prior operation. The most limiting initial conditions expected over the life of the plant should be based upon sensitivity studies.
It is not necessary to assume initial conditions that are impossible to occur in combination. For example, beginning-of-life peaking factors and end-of-life l
l B-4
decay heat may be an unrealistic condition and, therefore, should not be con-sidered. Given the assumed initial conditions, relevant factors such as the actual total power, actual peaking factors, and actual fuel conditions should be calculated in a best-estimate manner.
Calculations should be performed that are representative of the spectrum of possible break sizes from the full double-ended break of the largest pipe to a size small enough that it can be shown that smaller breaks are of less conse-quence than those already considered. The analyses should also include the effects of longitudinal splits in the largest pipes, with the split area equal to twice the cross-sectional area of the pipe. The detail of break sizes con-sidered should be sufficient that the system response as a function of break size is defined well enough to confidently interpolate between calculations, without considering unexpected behavior between the break sizes. In ECCS per-formance calculations, the break should be assumed to occur instantaneously.
Other boundary and initial conditions and equipment availability should be based on plant technical specification limits. These other conditions include, but may not be limited to, availability and performance of equipment, automatic controls, and operator actions. It should be noted that Appendix A to 10 CFR Part 50 requires that a single failure be considered when analyzing safety system performance. In addition, Appendix A to 10 CFR Part 50 requires that the analysis consider the effect of using only onsite power and only off-site power.
1.2.3 Sources of Heat During a loss-of-Coolant Accident The sources of heat discussed below and the distribution of heat production should be accounted for.
1.2.3.1 Initial Stored Energy of the Fuel. The steady-state temperature distribution and stored energy in the fuel before the postulated accident should be calculated in a best-estimate manner for the assumed initial conditions, fuel conditions, and operating history. To accomplish this, the thermal conduc-tivity of the fuel pellets and the thermal conductance of the gap between the fuel pellet and the cladding should be evaluated. Thermal conductivity of fuel is a function of temperature and is degraded by the presence of gases in crack voids between fuel fragments. An acceptable model for thermal conductivity B-5
should be developed from the in pile test results for fuel centerline and off-center temperatures taking into account the conductivity of gases in crack voids.
Thermal conductance of the fuel-cladding gap is a strong function of hot gap size and the composition and pressure of the gases in the fuel rod. The calculation of hot gap size should take into account UO2 or mixed-oxide fuei swelling, densification, creep, thermal expansion and relocation, and cladding creep. Fuel swelling is a function of temperature and burnup. Fuel densifica-tion is a function of burnup, temperature, and initial density. Densification can result from hydrostatic stresses imposed on fuel during pellet-cladding mechanical interaction and should be considered. Fuel creep is a function of time, temperature, grain size, density, fission rate, oxygen-to-metal ratio,
- nd external stress. Fuel thermal expansion represents dimensional changes in unirradiated fuel pellets caused by changes in temperature. An acceptable model for fuel swelling should be based on in pile and out-of pile test data.
Cladding creep introduces compressive creep strain in cladding during steady-state operation reducing the gap between the fuel pellet and cladding. Cladding creep is a function of fast neutron flux (>l MeV), cladding temperature, hoop stress, and material. Cladding materials may be cold worked and stress relieved j or fully recrystallized, and there is a significant difference in the magnitude of creepdown between these materials. During pellet-cladding mechanical inter-action, cladding experiences deformation from tensile creep, which is signifi-cantly different from deformation caused oy compressive creep. An acceptable model for cladding tensile creep should be based on in-reactor tensile creep data.
Best-estimate fuel models will be considered acceptable provided they include essential data on the phenomena identified above and provided their technical basis is demonstrated with appropriate data and analyses.
1.2.3.2 Fission Heat. Fission heat should be included in the calculation and should be calculated using best-estimate reactivity and reactor kinetics calculations. Shutdown reactivities resulting from temperatures and voids should also be calculated in a best-estimate manner. The point kinetics formulation is considered an acceptable best-estimate method for determining fission heat in safety calculations for loss-of-coolant accidents. Other best-estimate models will be considered acceptable provided their technical basis is demonstrated B-6 l
with appropriate data and analyses. Control rod assembly insertion should be assumed if it is expected to occur.
1.2.3.3 Decay of Actinides. The heat from the radioactive decay of actinides, including neptunium and plutonium generated during operation as well as isotopes of uranium, should be calculated in accordance with fuel cycle calculations and known radioactive properties. The actinide decay heat chosen should be that appropriate for the operating history.
1.2.3.4 Fission Product Decay Heat. The heat generation rates from the radioactive decay of fission products, including the effects of neutron capture, should be included in the calculation and should be calculated in a best-estimate manner. The energy release per fission (Q value) should also be calculated in a best-estimate manner. Best-estimate methods will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
1.2.3.5 Metal-Water Reaction Rate. The rate of energy release, hydrogen h
F generati n, and clad ing xid tion fr m the reacti n f the zircal y cl dding with steam should be included in the calculation in a best-estimate manner.
Best-estimate model:: will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses. For rods calculated to rupture their cladding during the loss of-coolant accident, the oxidation of the inside of the clading should be included in the calculation in a best-estimate manner.
1.2.3.6 Heat Transfer from Reactor Internals. Heat transfer from piping, vessel walls, and internal hardware should be included in the calculation and should be calculated in a best-estimate manner. Models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
l 1.2.3.7 Primary to Secondary System Heat Transfer (not applicable to boiling water reactors). Heat transferred between the primary and secondary systems through the steam generators should be considered in the calculation and should be calculated in a best-estimate manner. Models will be considered D acceptable provided their technical basis is demonstrated with appropriate data l and analyses.
B-7
l.2.4 Swelling and Rupture of the Cladding, Fuel Rod Thermal Parameters {
A calculation of the swellihg and rupture of the cladding resulting from the temperature distribution in the cladding and from the pressure difference between the inside and outside of the cladding, both as a function of time, should be included in the analysis and should be performed in a best-estimate manner. The degree of swelling and rupture should be taken into account in calculations of gap conductance, cladding oxidation and embrittlement, hydrogen generation, and in calculating heat transfer and fluid flow outside of the cladding. The calculations of fuel and cladding temperatures as a function of time should use values of gap conductance and other thermal parameters as func-tions of temperature and time. Best-estimate methods to calculate the swelling of the cladding should take into account spatially varying cladding temperatures, heating rates, anisotropic material properties, asymmetric deformation of clad-ding, and fuel rod thermal and mechanical parameters. Best-estimate methods will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
1.2.5 Blowdown Phenomena 1.2.5.1 Break Characteristics and Flow. In analyses of hypothetical loss-of-coolant accidents, a spectrum of possible break sizes should be considered as indicated in section 1.2.2 above. The discharge flow rate should be calcu-lated with a critical flow rate model that considers the fluid conditions at I the break location, upstream and downstream pressures, and break geometry. The critical flow model should be justified by comparison to applicable experimen-tal data over a range of conditions for which the model is applied. The model should be a best estimate calculation, with uncertainty in the critical flow rate included as part of the uncertainty evaluation. Best-estimate models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
1.2.5.2 ECC Bypass. The best-estimate code should contain a calculation of the amount of injected cooling water that bypasses the vessel during the blowdown phase of the loss of-coolant accident. The calculation of ECC bypass should be a best estimate calculation using analyses and comparisons with ap-plicable experimental data. Although it is clear that tha dominant processes B-8
governing ECC bypass are multidimensional, single-dimensional approximations justified through sufficient analysis and data may be acceptable. Best-estimate methods will be considered acceptable provided their technical basis is demon-
! strated with appropriate data and analyses. Cooling water that is not expelled, but remains in piping or is stored la parts of the vessel, should be calculated in a best-estimate manner based on applicable experimental data.
1.2.6 Noding Near the Break and ECCS Injection Point The break location and ECCS injection point are areas of large thermal nonequilibrium and contain phenomena that are often difficult to calculate.
The results of these calculations are often highly dependent on the noding.
Sufficient sensitivity studies should be performed on the noding and other important parameters to ensure that the calculations provide realistic results.
L 2.7 Frictional Pressure Drop The frictional losses in pipes and other components should be calculated using models that include variation of friction factor with Reynolds number and account for two phase flow effects on friction. Best-estimate models will be D considered acceptable provided their technical basis is demonstrated with appro-priate data and analyses.
1.2.8 Momentum Equation The following effects should be taken into account in the two phase con-I servation of momentum equation: (1) temporal change in momentum, (2) momentum convection, (3) area change momentum flux, (4) momentum change from compress-ibility, (5) pressure loss resulting from wall friction, (6) pressure loss resulting from area change, and (7) gravitational acceleration. Models will be considered acceptable provided their technical basis is demonstrated with appro-priate data and analyses.
1.2.9 Critical Heat Flux Best-estimate models developed from appropriate steady-state or transient experimental data should be used in calculating critical heat flux (CHF) during loss-of-coolant accidents. The codes in which these models are used should contain suitable checks to ensure that the range of conditions over which these D correlations are used are within those intended. Research has shown that CHF i
B-9
" .-___-____n________________________-___________-______l
O is highly dependent on the fuel rod geometry, local heat flux, and fluid condi-tions. After CHF is predicted at an axial fuel rod location, the calculation may use nucleate boiling heat transfer correlations if the calculated local fluid and surface conditions justify the reestablishment of nucleate boiling.
Best estimate models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
1.2.10 Post-CHF Heat Transfer Models of heat transfer from the fuel to the surrounding fluid in the post-CHF regimes of transition and film boiling should be best-estimate models based on comparison to applicable steady-state or transient data. Any models used should be evaluated to demonstrate that they provide acceptable results over the applicable ranges. Best-estimate models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses.
1.2.11 Pump Modeling The characteristics of rotating primary system pumps should be derived from a best-estimate dynamic model that includes momentum transfer between the fluid q and the rotating member, with variable pump speed as a function of time. The pump model resistance and other empirical terms should be justified through comparisons with applicable data. The pump model for the two phase region should be verified by comparison to applicable two phase performance data. Pump coast-down following loss of power should be treated in a best estimate manner. A locked rotor following a large-break loss-of-coolant accident need not be assumed unless it is calculated to occur.
1.2.12 CoreFlowDistributionDurinc[dlowdown The core flow through the hottest region of the core during the blowdown should be calculated as a function of time. For purposes of these calculations, the hottest region of the core should not be greater than the size of one fuel assembly. Calculations of the flow in the hot region should take into account any cross-flow between regions and any flow blockage calculated to occur during the blowdown as a result of cladding swelling or rupture. The numerical scheme should ensure that unrealistic oscillations of the calculated flow do not result.
G B-10 d
L 1.2.13 Post-Blowdown Phenomena 1.2.13.1 Containment Pressure. The containment pressure used for evalu-ating cooling effectiveness during the post-blowdown phase of a loss-of-coolant !
accident should be calculated in a best-estimate manner that includes the effects of containment heat sinks. The calculation should include the effects of opera-ting all pressure-reducing equipment assumed to be available as discussed in section 1.2.2 of this appendix.
1.2.13.2 Calculation of Post-Blowdown Thermal Hydraulics for Pressurized Water Reactors. The refilling of the reactor vessel and the ultimate reflooding of the core should be calculated by a best-estimate model that takes into con-sideration the thermal and hydraulic characteristics of the core, the emergency core cooling systems, and the primary and secondary reactor systems. The primary coolant _ pumps should be assumed to be operating in the expected manner based on the assumptions of section 1.2.11 when calculating the resistance offered by the pumps to fluid flow. Models will be considered acceptable provided their technical basis is demonstrated through comparison with appropriate da'ta and analyses. The total fluid flow leaving the core exit (carryover) should be calculated using a best-estimate model that includes the effect of cross-flow on carryover and core fluid distribution. Thermal-hydraulic phenomena asso-ciated with unique emergency core cooling systems, such as upper plenum injec-tion and upper head injection, should be accounted for. The effects of the
, compressed gas in the accumulator that is discharged during accumulator water discharge should be included in the calculation. Any model or code used for this calculation should be assessed against applicable experimental data.
l 1.2.13.3 Steam Interaction with Emergency Core Cooling Water in Pressur-ized Water Reactors. The thermal-hydraulic interaction between the steam or two phase fluid and the emergency core cooling water should be taken into account in calculating the core thermal hydraulics and the steam flow through the reactor coolant pipes during the time the accumulators are discharging water.
1.2.13.4 Post-Blowdown Heat Transfer for Pressurized Water Reactors.
During the refilling of the reactor vessel and ultimate reflooding of the core, D the heat transfer data should be based on a best-estimate calculation of the B-11
fluid flow through the core, accounting for unique emergency core cooling systems, and should include the effects of any flow blockage calculated to ,
occur as a result of cladding swelling or rupture. Heat transfer calculations that account for two phase conditions in the core during refilling of the reactor vessel should be justified through comparisons with experimental data.
Models will be considered acceptable provided their technical basis is demon-strated through comparison with appropriate data and analyses.
1.2.14 Convective Heat Transfer Coefficients for Boiling Water Reactor Rods Under Spray Cooling Models will be considered acceptable provided their technical basis can be justified with appropriate data and analyses. These models should contain the following:
- 1. Following the blowdown period, convective heat transfer coefficients should be determined based on the calculated fluid conditions and heat transfer modes within the bundle and the calculated-rod temperatures.
~
l
- 2. During the period following the flashing of the lower plenum fluid, but prior to ECCS initiation, heat transfer models should include cooling by steam flow or by a two phase mixture, if calculated to occur.
- 3. Following initiation of ECCS flow, but prior to reflooding, heat transfer should be based on the actual calculated bundle fluid condi-tions and best-estimate heat transfer models that take into account rod-to-rod variations in heat transfer.
- 4. After the two phase reflood level reaches the level under considera-tion, a best-estimate heat transfer model should be used that includes the effects of any flow blockage calculated to occur as a result of cladding swelling or rupture.
- 5. Thermal-hydraulic models that do not calculate multiple channel effects should be compared with applicable experimental data or more detailed B-12
calculations to ensure that all important phenomena are adequately calculated.
1.2.15 Boiling Water Reactor Channel Box Under Spray Cooling Following the blowdown period, heat transfer from the channel box and wet-ting of the channel box should be based on the calculated fluid conditions on both sides of the channel box and should make use of best-estimate heat trans-fer and rewetting models that have been compared with applicable experimental data.
1.2.16 Special Considerations for Small-Break Loss-of-Coolant Accidents in Pressurized Water Reactors The slower, small-break loss-of-coolant accident leads to fluid conditions characterized by separation of the fluid phases versus the more homogeneous fluid conditions that would result from rapid large-break loss-of-coolant acci-dent transients. Phenomena that would occur in a pressurized water reactor during a small-break loss-of-coolant accident would, therefore, be significantly different from those phenomena that would occur during a large-break loss-of-D coolant accident. The distribution of liquid throughout the reactor system, in addition to the total liquid inventory, is of increased importance for the small-break loss-of-coolant accident. A number of special factors must be given increased consideration in small-break loss-of-coolant accident calcula-tions to correctly calculate phenomena influenced by the liquid inventory distribution.
Break flow may be greatly influenced by the location and specific geometry of the break. For a break in a horizontal pipe containing stratified flow, the quality of the break flow will be a strong function of the assumed location of the break on the pipe (e.g., top or bottom). Small-break loss-of-coolant acci-dent calculations should, therefore, include various assumed break locations in the spectrum of breaks analyzed. The assumed operating state of the reactor
, coolant pump will also influence the distribution of liquid throughout the system and the amount of liquid lost through the break. The pump operation assumptions used in the calculations should be the most likely, based on operat-l ing procedures, with appropriate consideration of the uncertainty of the pump operation during an actual event. Level depression in the core region and B-13
subsequent core heatup may be influenced by liquid holdup in the steam genera-tor tubes, manometric effects of liquid in the piping and loop seal region, and liquid levels relative to vent paths for steam through upper plenum bypass flow paths and vent valves. Steam generator heat transfer under " reflux" or " boiler-condensor" modes of operation may also strongly influence core inventory through level depression and the effect on total system pressure and thus ECCS flow.
These phenomena should be carefully considered in the calculation, as well as sensitivity studies of the importance of these effects for use in the uncer-tainty evaluation.
Hest transfer from an uncovered core under high pressure conditions typi-cal during a small-break loss-of-coolant accident may include contributions from both convective and radiation heat transfer to the steam. Models will be considered acceptable provided their technical basis is demonstrated through comparison with appropriate data and analyses.
1.2.17 Other Features of Best-Estimate Codes No list of best-estimate code features could be all-inclusive because the important features of a best-estimate code may vary depending on the transient {
to be calculated and the required accuracy of the calculation. Therefore, no attempt has been made to construct an exhaustive list of best-estimate code features. Rather, features that were identified as important for inclusion in Appendix K were used as a basis for the above list. These features are not necessarily any more or less important than other code features, but were high-lighted because it is necessary to give specific examples of how current best-estimate models may vary from methods traditionally used in evaluating models or codes using the various Appendix K conservatisms. In addition, models have not been included for certain areas if it was felt that the best model would be highly dependent on the specific plant design or the specific transient under consideration. The NRC staff believes that good examples of best-estimate thermal-hydraulic transient codes are those developed by the NRC (e.g., TRAC-PWR, TRAC-BWR, RELAPS, COBRA, and FRAP). Although these codes are subject to further improvement, based on their ongoing use and assessment, they currently provide reasonable best-estimate calculations of a loss-of-coolant accident in a full-scale light water reactor. This is substantiated in the code develop-ment and assessment literature published by the NRC and its contractors over l
the past several years.
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It is possible, however, to describe in general how other features of best-l estimate codes should be constructed. Two basic criteria should be applied:
1.2.17.1 Completeness. Best-estimate codes should contain models in sufficient detail to predict phenomena that are important to the desired result of the calculation (e.g., peak cladding temperature). Simplifications are acceptable as long as code uncertainties or bias do not become so large so that they would cast doubt on the actual behavior that would occur and the true effect of assumed initial and boundary conditions (e.g., equipment sizing, safety system settings). Comparisons of the overall calculations to integral experi-ments should be performed to ensure that important phenomena can be predicted and to help in making judgments on the effect of code simplifications. Consid-eration should also be given to the uncertainty and validity of the experiment to ensure that meaningful comparisons are being made.
1.2.17.2 Data Comparisons. Individual best-estimate models should be compared to applicable experimental data to ensure that realistic results are l
predicted and that relevant experimental variables are included. Uncertainty analyses are required to ensure that a major bias does not exist in the models and that the model uncertainty is small enough to provide a realistic estimate of the effect of important experimental variables. Uncertainty analyses should also consider experimental uncertainty to ensure that meaningful comparisons are being made.
- 2. ESTIMATION OF OVERALL CALCULATIONAL UNCERTAINTY 2.1 General The term " uncertainty," when applied to best-estimate thermal-hydraulic transient codes, is used at two levels. At the lower or more detailed level, it refers to the degree to which an individual model, correlation, or method used within the code represents the physical phenomenon it addresses. These individual unctirtainties, when taken together, are the " code uncertainty."
The combined uncertainty associated with individual models (i.e., code uncertainty) within the best-estimate codes does not account for all the D uncertainty associated with its use. In addition to the code uncertainty, 8-15
various other sources of uncertainty are introduced when attempting to use {
best-estimate codes to predict f011-scale plant thermal-hydraulic response.
These sources of uncertainty include those associated with the experimental data used in the code assessment process (including applicability of the data to full-scale reactors), the input boundary and initial conditions, and the fuel behavior. Additional sources of uncertainty stem from the use of simpli-fying assumptions and approximations. A careful statement of these assumptions and approximations should be made, and the uncertainty associated with them should be taken into account. Therefore, the "overall calculational uncertainty" is defined as the uncertainty arrived at when all the contributions from the sources identified above, including the code uncertainty, are taken into account.
A 95% probability level would be considered acceptable by the NRC staff i for comparing best estimate predictions to the applicable limits of paragraph 1
50.46(b) to meet the requirement of proposed paragraph 50.46(a)(1)(i) to show that there is a high probability that the criteria would not be exceeded. The basis for selecting the 95% probability level is primarily for consistency with standard engineering practice in regulatory matters involving thermal hydraulics.
Many parameters, most notably the departure from nucleate boiling ratio (DNBR),
have been found acceptable at the 95% probability level by the NRC staff in the past. It is useful, however, to examine this criterion on a risk basis. To illustrate by way of an example, assume that the probability of a large pipe rupture is approximately 1 x 10 4 per reactor year. Assume also that a hypo-thetical large-break loss-of-coolant accident calculation results in a best-estimate peak cladding temperature of 2000 F with a standard deviation of 122 F.
When overall calculational uncertainties are considered, the probability of exceeding the 2200 F limit in paragraph 50.46(b) would be 0.05.* Thus, the probability of a large pipe rupture resulting in a peak cladding temperature exceeding 2200 F would be 5 x 10 8 Only the hottest rods would exceed the 2200 F limit under these assumptions, and the 2200 F limit is itself a conserv-ative requirement chosen to prevent significant embrittlement of the cladding to guarantee that a coolable geometry is maintained. The accident at TMI-2 resulted in a significant percentage of the core exceeding the 2200 F limit for
- This assumes that the temperature distribution is normal with no bias (z = 1.64 standard deviations for a single-sided test at the 95% level). These assump-tions may or may not be valid for a given evaluation model, and assumptions such as these would have to be clearly stated and justified.
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1 long periods of time.
Despite the significant damage, the core still remained coolable.
Therefore, if it is further assumed that gross core damage occurs in only one in ten cases where the 2200 F limit is exceeded, the overall probabil-ity of gross core damage would be on the order of 5 x 10 7 This probability level is well below many other significant contributors to risk, and therefore the use of a 95% probability level should ensure that large-break loss-of-coolant accidents are not important contributors to risk.
This 95% probability level would also be applied to small-break loss-of-coolant accidents that have a higher probability than large breaks. The domi-nant factors influencing risk from small-break loss of-coolant accidents include equipment availability and operator actions.
Calculational uncertainties are much less important than factors such as operator recognition of the event, the availability of equipment, and the correct use of this equipment. The use of a best estimate calculation with reasonable and quantifiable uncertainty is expected to provide a reduction in the overall risk from a small-break loss of-coolant accident by providing more realistic calculations with which to evaluate operator guidelines and to determine the true effect of equipment availability.
D code uncertainty evaluation that would be required by paragraph 50.46(a)(1). This uncertainty evaluation should make use of statis-tically based methods to determine the code uncertainty and the bias (if any).
For a calculation of this complexity, a completely rigorous mathematical treat-ment of the statistics is neither practical nor required. In many cases, approximations and assumptions may be made to make the overall calculational uncertainty evaluation possible. A careful statement of these assumptions and approximations should be made so that the NRC staff can make a judgment as to i
the validity of the uncertainty evaluation.
The purpose of the uncertainty evaluation is to provide assurance that for postulated loss of-coolant accidents a given plant will not, with a probability of 95% or more, exceed the applicable limits specified in paragraph 50.46(b).
- 2. 2 Code Uncertainty This regulatory guide makes a distinction between the terms " code uncer-tainty" and "overall calculational uncertainty." The latter term is defined in section 2.1 of this appendix and includes the contributions to the uncertainty I
8-17 l
\
The features of the code uncertainty (i.e.,
described in sections 2.2 and 2.3.
the contribution to the overall uncertainty caused by the models and numerical methods used) are described in this section.
The code uncertainty should be evaluated through direct data comparison with relevant integral systems and separate effects experiments at different scales. In this manner, an estimate of the uncertainty attributable to the combined effect of the models and correlations within the code can be obtain for all scales and different phenomena. Comparison to a sufficient number of integral systems experiments, from different test facilities and different scales, should be made to ensure that a reasonable estimate of code uncertainty When necessary, separate effects experiments should and bias has been obtained.
be utilized to establish code uncertainty for specific phenomena (e.g., compari-sons to cylindrical core test facility data to ascertain code uncertainty in Code comparisons should account modeling upper plenum injection performance).
for limitations of the measurements and calibration errors.
These comparisons should be performed for important key parameters to For large-break demonstrate the overall best-estimate capability of the code.
loss-of-coolant accidents, the most important key parameter is peak cladding temperature, which is one of the criteria of paragraph 50.46(b) and has a direct In addition, a code uncertainty evaluation influence on the other criteria.
should be performed for any other important parameter for the transient of interest to evaluate compensating errors. For small-break loss-of-coolant acci-dents, the cladding temperature response is again the most important parameter, but the ability of the codes to predict overall system mass and reactor vessel inventory distribution should also be statistically examined.
In evaluating the code uncertainty, it will be necessary to evaluate the code's predictive ability over several time intervals, since different processes For example, in large-break and phenomena occur at different time intervals.
loss-of-coolant accident evaluations, separate code uncertainties may be required for the peak cladding temperature during the blowdown and post-blowdown phases.
Justification for treating these uncertainties individually, or methods for combining these uncertainties, should be provided.
The experimental information used to determine code uncertainty will usually be obtained from facilities that are much smaller than nuclear power reactors.
Applicability of these results should be justified for larger scales.
The g effects of scale can be assessed through comparison to available large-scale W
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l l
long periods of time. Despite the significant damage, the core still remained coolable. Therefore, if it is further assumed that gross core damage occurs in only one in ten cases where the 2200 F limit is exceeded, the overall probabil-ity of gross core damage would be on the order of 5 x 10 7 This probability level is well below many other significant contributors to risk, and therefore the use of a 95% probability level should ensure that large-break loss-of-coolant accidents are not important contributors to risk.
This 95% probability level would also be applied to small-break loss-of-coolant accidents that have a higher probability than large breaks. The domi-nant factors influencing risk from small-break loss-of-coolant accidents include cquipment availability and operator actions. Calculational uncertainties are much less important than factors such as operator recognition of the event, the cvailability of equipment, and the correct use of this equipment. The use of a best estimate calculation with reasonable and quantifiable uncertainty is cxpected to provide a reduction in the overall risk from a small-break loss-of-coolant accident by providing more realistic calculations with which to evaluate operator guidelines and to determine the true effect of equipment availability.
This section describes the features that should be included in the overall code uncertainty evaluation that would be required by the proposed revision to paragraph 50.46(a)(1). This uncertainty evaluation should make use of statis-tically based methods to determine the code uncertainty and the bias (if any).
For a calculation of this complexity, a completely rigorous mathematical treat-ment of the statistics is neither practical nor required. In many cases, approximations and assumptions may be made to make the overall calculational uncertainty evaluation possible. A careful statement of these assumptions and cpproximations should be made so that the NRC staff can make a judgment as to the validity of the uncertainty evaluation. The purpose of the uncertainty evaluation is to provide assurance that for postulated loss-of-coolant accidents a given plant will not, with a probability of 95% or more, exceed the applicable limits specified in paragraph 50.46(b).
- 2. 2 Code Uncertainty This regulatory guide makes a distinction between the terms " code uncer-tainty" and "overall calculational uncertainty." The latter term is defined in I
section 2.1 of this appendix and includes the contributions to the uncertainty B-17
l described in sections 2.2 and 2.3. The features of the code uncertainty (i.e.,
the contribution to the overall uncertainty caused by the models and numerical ,
methods used) are described in this section.
The code uncertainty should be evaluated through direct data comparison with relevant integral systems and separate effects experiments at different scales. In this manner, an estimate of the uncertainty attributable to the combined effect of the models and correlations within the code can be obtained for all scales and different phenomena. Comparison to a sufficient number of integral systems experiments, from different test facilities and different scales, should be made to ensure that a reasonable estimate of code uncertainty and bias has been obtained. When necessary, separate effects experiments should be utilized to establish code uncertainty for specific phenomena (e.g., compari-sons to cylindrical core test facility data to ascertain code uncertainty in modeling upper plenum injection performance). Code comparisons should account for limitations of the measurements and calibration errors.
These comparisons should be performed for important key parameters to demonstrate the overall best estimate capability of the code. For large-break loss-of-coolant accidents, the most important key parameter is peak cladding temperature, which is one of the criteria of paragraph 50.46(b) and has a direct influence on the other criteria. In addition, a code uncertainty evaluation should be performed t e any other important parameter for the transient of interest to evaluate compensating errors. For small-break loss-of-coolant acci-dents, the cladding temperature response is again the most important parameter, but the ability of the codes to predict overall system mass and reactor vessel inventory distribution should also be statistically examined.
In evaluating the code uncertainty, it will be necessary to evaluate the code's predictive ability over several time intervals, since different processes and phenomena occur at different time intervals. For example, in large-break loss-of-coolant accident evaluations, separate code uncertainties may be required for the peak cladding temperature during the blowdown and post-blowdown phases.
Justification for treating these uncertainties individually, or methods for combining these uncertainties, should be provided.
The experimental information used to determine code uncertainty will usually be obtained from facilities that are much smaller than nuclear power reactors.
Applicability of these results should be justified for larger scales. The effects of scale can be assessed through comparison to available large-scale 8-18
separate effects tests and through comparison to integral tests from various-h y size facilities. If there are scaling problems, particularly if predictions are nonconservative, the code should be improved for large-scale nuclear plants.
Codes not having scaling capability will not be acceptable if their predictions are nonconservative.
2.3 Other Sources of Uncertainty When a best estimate methodology is used to predict reactor transients, sources of uncertainty other than the limitations in the individual models and numerical methods (i.e., code uncertainty) are introduced. The following contributors to the overall calculational uncertainty should also be considered in the uncertainty analysis.
2.3.1 Initial and Boundary Conditions and Equipment Availability When a plant input model is prepared, certain relationships describing the plant boundary and initial conditions and the availability and performance of equipment are defined. These include factors such as initial power level, pump I performance, valve activation times, and control systems functioning. Uncer-tainties associated with the boundary and initial conditions and the character-ization and performance of equipment should be accounted for in the uncertainty evaluation. It is also acceptable to limit the variables to be considered by setting their values to conservative bounds.
2.3.2 Fuel Behavior Variability of the results of plant transient calculations can result from uncertainties associated with fuel behavior, which are not included in the comparisons of code results with integral experiments since most integral tests use electrical heater rods. This uncertainty includes many effects such as fuel conductivity, gap width, gap conductivity, and peaking factors. These uncertainties should be quantified and used in the determination of the overall calculational uncertainty.
2.3.3 Other Variables There may be individual models within the best-estimate code whose effect may not have been evaluated by the comparison to the integral systems data.
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For example, since most integral systems experiments use electrically heated q rods, uncertainties associated with the prediction of core decay heat and cladding-metal-water reaction will not have been evaluated. In addition, to demonstrate the overall adequacy of the predictive ability of the best estimate code, it may be necessary to use empirical break-discharge coefficients in order to obtain a reasonable break flow. The uncertainty in the individual models that have not been evaluated via the integral systems data comparisons should be quantified and used in the determination of overall code uncertainty.
2.4 Statistical Treatment of Overall Calculational Uncertainty The methodology used to obtain an estimate of the overall calculational uncertainty at the 95% probability limit should be provided and justified. If linear independence is assumed, suitable justification should be provided. In examining the influence of the individual parameters on code uncertainty, justi-fication should be provided for the assumed distribution of the parameter and the range considered.
A one-sided probability interval should be used to establish the 95% prob-ability limit. Consideration of the 95% probability level and the distribution curve used for the key parameter should be provided. In lieu of a detailed consideration of these effects, use of two standard deviations for evaluating the 95% probability level for normal distributions is acceptable.
The evaluation of the peak cladding temperature at the 95% probability level need only be performed for the worst-case break identified by the break spectrum analysis in order to demonstrate conformance with paragraph 50.46(b).
However, in order to use this approach, justification must be provided that demonstrates that the overall calculational uncertainty for the worst case bounds the uncertainty for other breaks within the spectrum. It may be neces-sary to perform separate uncertainty evaluations for large- and small-break loss-of-coolant accidents because of the substantial difference in system thermal-hydraulic behavior. The proposed paragraph 50.46(a)(1)(i) would require showing with a high probability that all the criteria of paragraph 50.46(b) will not be exceeded, not just the peak cladding temperature criterion. However, since the other criteria are strongly dependent on peak cladding temperature, explicit consideration of the probability of exceeding the other criteria may l
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I l
not be required if it can be demonstrated that meeting the temperature criterion D ,
at the 95% probability level ensures with a similar or greater probability that the other criteria will not be exceeded.
2.5 NRC Approach to LOCA Uncertainty Evaluation Chapter 4 of the " Compendium of ECCS Research for Realistic LOCA Analysis" !
(Ref. B-7) presents a methodology that is being considered for evaluating a best-estimate LOCA analysis code and determining the overall calculational uncertainty in peak cladding temperature predictions.
I B-21
References B-1 U.S. Nuclear Regulatory Commission (USNRC), " TRAC-PF1/M001: An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Thermal- l Hydraulic Analysis," Washington, DC, NUREG/CR-3858 (prepared for NRC by Los Alamos National Laboratory, LA-10157-MS), July 1986.
l B-2 USNRC, " TRAC-BD1/M001: An Advanced Best-Estimate Computer Program for Boiling Water Reactor Transient Analysis," Washington, DC, NUREG/CR-3633, 4 Vols. (prepared for NRC by EG&G, Inc., EGG-2294), April 1984.
B-3 Idaho National Engineering Laboratory, "RELAP5/M002 Code Manual," EGG-2396, Vol. 1-2 (NUREG/CR-4312), August 1985. (Available in the NRC Public Document Room and from EG&G Idaho, Inc. , P.O. Box 1625, Idaho Falls, Idaho 83415.)
B-4 USNRC, " COBRA / TRAC - A Thermal Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems," Washington, DC, I
NUREG/CR-3046 (prepared for NRC by Pacific Northwest Laboratory, PNL-4385),
Vol. 1-5, March 1983.
B-5 L. J. Siefken et al., "FRAP-T6: A Computer Code for the Transient Analysis of 0xide Fuel Rods," USNRC, Washington, DC, NUREG/CR-2148 (prepared for NRC t,y EG&G, Inc., EGG-2104), May 1981.
B-6 G. A. Berna et al., "fRAPCON-2: A Computer Code for the Calculation of l Steady State Thermal-Mechanical Behavior of 0xide Fuel Rods," USNRC, Washington, DC, NUREG/CR-1845 (prepared for NRC by Battelle Memorial Institute), January 1981.
B-7 USNRC, " Compendium of ECCS Research for Realistic LOCA Analysis,"
Washington, DC, NUREG-1230 (to be published April 1987).
l l
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REGULATORY ANALYSIS A separate regulatory analysis was not prepared for this draft regulatory guide. A draft regulatory analysis that examines the costs and benefits of the rule as implemented by the guide was prepared for the proposed amendments to 10 CFR Part 50, which provides the regulatory basis for this guide. The draft cnalysis is available for inspection and copying for a fee in the NRC Public Document Room, 1717 H Street NW., Washington, DC.
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