Regulatory Guide 1.157

(Task Rs 701-4), Best-Estimate Calculations of Emergency Core Cooling System Performance

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U.S. NUCLEAR REGULATORY COMMISSION May 1989 REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.157 (Task RS 701-4)

BEST-ESTIMATE CALCULATIONS OF EMERGENCY CORE COOLING

SYSTEM PERFORMANCE

A. INTRODUCTION

use either Appendix K features or a realistic1 evaluation model. These realistic evaluation models2 Section 50.46, "Acceptance Criteria for Emer must include sufficient supporting justification to gency Core Cooling Systems for Light-Water Nuclear demonstrate that the analytic techniques employed Power Reactors," of 10 CFR Part 50, "Domestic Li realistically describe the behavior of the reactor censing of Production and Utilization Facilities," re system during a postulated loss-of-coolant accident.

quires that light-water nuclear reactors fueled with Paragraph 50.46(a)(1) also requires that the uranium oxide pellets within cylindrical zircaloy clad uncertainty in the realistic evaluation model be ding be provided with emergency core cooling systems quantified and considered when comparing the results (ECCS) that are designed in such a way that their of the calculations with the applicable limits in calculated core cooling performance after a postu paragraph 50.46(b) so that there is a high probability lated loss-of-coolant accident (LOCA) conforms to that the criteria will not be exceeded.

certain criteria specified in paragraph 50.46(b). Para This regulatory guide describes models, 3 correla graph 50.46(b)(1) requires that the calculated maxi tions, 4 data, model evaluation procedures, and meth mum temperature of fuel element cladding not be ods that are acceptable-to the NRC staff for meeting

~ greater than 2200'F. In addition, paragraphs the requirements for a realistic or best-estimate calcu

50.46(b) (2) through (b) (5), which contain required lation of ECCS performance during a loss-of-coolant limits for calculated maximum cladding oxidation and accident and for estimating the uncertainty in that maximum hydrogen generation, require that calcu lated changes in core geometry remain amenable to 1 For the purpose of this guide, the terms "best-estimate" and "realis cooling and that long-term decay heat removal be tic" have the same meaning. Both terms are used to indicate that the provided. techniques attempt to predict realistic reactor system thermal-hydraulic response. Best-estimate is not used in a statistical sense in this guide.

2 On September 16, 1988, the NRC staff amended IThe term "evaluation model" refers to a nuclear plant system com puter code or any other analysis tool designed to predict the aggregate the requirements of § 50.46 and Appendix K, behavior of a reactor during a loss-of-coolant acciden

t. It can be either

"ECCS Evaluation Models" (53 FR 35996), so that best-estimate or conservative and may contain many correlations or these regulations reflect the improved understand models.

ing of ECCS performance during reactor transients MIhe term "model" refers to a set of equations derived from funda that was obtained through the extensive research mental physical laws that is designed to predict the details of a specific phenomenon.

performed since the promulgation of the original 4 requirements in January 1974. Paragraph The term "correlation" refers to an equation having empirically de termined constants such that it can predict some details of a specific phe

50.46(a)(1) now permits licensees or applicants to nomenon for a limited range of conditions.

USNRC REGUIATORY GUIDES The guides are issued In the following ten broad divisions:

Regulatory Guides are Issued to describe and make available to the pub lic methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the 1. Power Reactors 6. Products staff in evaluating specific problems or postulated accidents, or to pro 2. Research and Test Reactors 7. Transportation vide guidance to applicants. Regulatory Guides are not substitutes for 3. Fuels and Materials Facilities 8. Occupational Health regulations, and compliance with them is not required. Methods and 4. Environmental and Siting 9. Antitrust and Financial Review solutions different from those set out In the guides will be acceptable If 5. Materials and Plant Protection 10. General they provide a basis for the findings requisite to the issuance or continu ance of a permit or license by the Commission.

Copies of issued guides may be purchased from the Government Printing This guide was issued after consideration of comments received from Office at the current GPO price. Information on current GPO prices may the public. Comments and suggestions for Improvements In these be obtained by contacting the Superintendent of Documents, U.S.

guides are encouraged at all times, and guides will be revised, as ap Government Printing Office, Post Office Box 37082, Washington, DC

propriate, to accommodate comments and to reflect new Information or 20013-7082, telephone (202)275-2060 or (202)275-2171.

experience.

Written comments may be submitted to the Regulatory Publications Issued guides may also be purchased from the National Technical Infor Branch, DFIPS, ARM, U.S. Nuclear Regulatory Commission, Washing mation Service on a standing order basis. Details on this service may be ton, DC 20555. obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA 22161.

calculation. Methods for including the uncertainty in vided information that allows for quantification of the comparisons of the calculational results to the cri that conservatism. The results of experiments, com teria of paragraph 50.46(b), in order to meet the re puter code development, and code assessment allow quirement that there be a high probability that the more accurate calculations, along with reasonable es criteria would not be exceeded, are also described in timates of uncertainty, of ECCS performance during this regulatory guide. Paragraph 50.46(a) also per a postulated loss-of-coolant accident than is possible mits licensees to use evaluation models developed in using the Appendix K procedures.

conformance with Appendix K.

It was also found that some plants were being re Other models, data, model evaluation proce stricted in operating flexibility by limits resulting from dures, and methods will be considered if they are conservative Appendix K requirements.,Based on the supported by appropriate experimental data and research performed, it was determined that these re technical justification. Any models, data, model strictions could be relaxed through the use of more evaluation procedures, and methods listed as accept realistic calculations without adversely affecting able in this regulatory guide are acceptable in a ge safety. The Appendix K requirements tended to di neric sense only and would still have to be justified to vert both NRC and industry resources from matters the NRC staff as being appropriately applied and ap that are relevant to reactor safety to analyses with as plicable for particular plant applications. sumptions known to be nonphysical.

The regulatory position in this regulatory guide In recognition of the known conservatisms in Ap lists models, correlations, data, and model evaluation pendix K, the NRC adopted an interim approach in procedures that the NRC staff considers acceptable 1983, described in SECY-83-472,5 to accommodate for realistic calculations of ECCS performance. It industry requests for improved evaluation models for also provides a description of the acceptable features the purpose of reducing reactor operating restrictions.

of best-estimate computer codes and acceptable This interim approach was a step in the direction of methods for determining the uncertainty in the calcu basing licensing decisions on realistic calculations of lations. plant behavior. Although the approach permits many The Advisory Committee on Reactor Safeguards "best-estimate" methods and models to be used for has been consulted concerning this guide and has licensee submittals, it retains those features of Ap concurred in the regulatory position. pendix K that are legal requirements.

Any information collection activities discussed in The current revision of § 50.46 permits ECCS

this regulatory guide are contained as requirements in evaluation models to be fully "best-estimate" and re

10 CFR Part 50, which provides the regulatory basis moves the arbitrary conservatisms contained in the for this guide. The information collection require required features of Appendix K for those licensees ments in 10 CFR Part 50 have been cleared under wishing to use these improved methods. Safety is best OMB Clearance No. 3150-0011. served when decisions concerning the limits within which nuclear reactors are permitted to operate are based upon realistic calculation

B. DISCUSSION

currently being used in the resolution of almost all reactor safety issues (e.g., anticipated transients with The criteria set forth in § 50.46, "Acceptance out scram, pressurized thermal shock, and operator Criteria for Emergency Core Cooling Systems for guidelines) and is now available for one of the last Light-Water Nuclear Power Reactors," and the cal remaining major issues still treated in a prescriptive culational methods specified in Appendix K were manner, the loss-of-coolant accident.

promulgated in January 1974 after extensive rulemaking hearings and were based on the under The NRC staff amended § 50.46 of 10 CFR Part standing of ECCS performance available at that time. 50 to allow realistic methods to be used for the ECCS

In the years following the promulgation of those performance calculations in place of the evaluation rules, the NRC, the nuclear industry, and several for models that use the required Appendix K features.

eign institutions have conducted an extensive pro This rule change also requires analysis of the uncer gram of research that has greatly improved the un tainty of the best-estimate calculations and requires derstanding of ECCS performance during a that this uncertainty be considered when comparing postulated loss-of-coolant accident. The methods the results of the calculations to the limits of para specified in Appendix K were found to be highly con graph 50.46(b) so that there is a high probability that servative; that is, the fuel cladding temperatures ex pected during a loss-of-coolant accident would be "6Information Report from William J. Dircks to the Commissioners, much lower than the temperatures calculated using dated November 17,1983, "Emergency Core Cooling System Analy Appendix K methods. In addition to showing that sis Methods," SECY-83-472. Available for inspection or copying for a fee in the NRC Public Document Room, 2120 L Street NW.,

Appendix K is conservative, the ECCS research pro- Washington, DC.

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the criteria will not be exceeded. In this manner, the phenomenon. The model should be compared more realistic calculations are available for regulatory with applicable experimental data and should predict decisions, yet an appropriate degree of conservatism the mean of the data, rather than providing a bound would be maintained. to the data. The effects of all important variables should be considered. If it is not possible or practical Many of the methods and models needed for a to consider a particular phenomenon, the effect of S best-estimate calculation are the same as those used ignoring this phenomenon should not normally be previously for evaluation model analyses. Although treated by including a bias in the analysis directly, but licensees and applicants are well acquainted with should be included as part of the model uncertainty.

them, explicit guidance on acceptable methods and The importance of neglecting a particular phenome models (based on NRC experience with its own best non should be considered within the overall calcula estimate advanced codes such as TRAC-PWR, tional uncertainty.

TRAC-BWR, RELAP5, COBRA, and FRAP) would Careful consideration should be given to the be useful. Further, the NRC has not previously pub lished acceptable methods for uncertainty analyses. range of applicability of a model when used in a best Therefore, guidance on methods acceptable to the estimate code. When comparing the model to data, NRC staff for calculating ECCS performance and for judgments on the applicability of the data to the situ estimating the uncertainty are provided in the follow ation that would actually occur in a reactor should be made. Correlations generally should not be extrapo ing Regulatory Position.

lated beyond the range over which they were devel oped or assessed. If the model is to be extrapolated

C. REGULATORY POSITION

beyond the conditions for which valid data compari sons have been made, judgments should be made as

1. BEST-ESTIMATE CALCULATIONS to the effect of this extrapolation and the effect A best-estimate calculation uses modeling that should be accounted for in the uncertainty evalu attempts to realistically describe the physical proc ation. The fundamental laws of physics, well esses occurring in a nuclear reactor. There is no established data bases (e.g., steam tables), and sensi unique approach to the extremely complex modeling tivity studies should be used to assist in estimating the of these processes. The NRC has developed and as uncertainty that results from extrapolation.

sessed several best-estimate advanced thermal A best-estimate code contains all the models nec hydraulic transient codes. These include TRAC essary to predict the important phenomena that might PWR, TRAC-BWR, RELAP5, COBRA, and the occur during a loss-of-coolant accident. Best-estimate

. FRAP series of codes (References 1 through 6). code calculations should be compared with applicable These codes reasonably predict the major phenom experimental data (e.g., separate-effects tests and in ena observed over a broad range of thermal tegral simulations of loss-of-coolant accidents) to de hydraulic and fuel tests. Licensees and applicants termine the overall uncertainty and biases of the cal may use, but are not limited to, these codes and the culation. In addition to providing input to the specific models within them to perform best-estimate uncertainty evaluation, integral simulation data com calculations of emergency core cooling system parisons should be used to ensure that important phe (ECCS) performance. Since the NRC staff has not nomena that are expected to occur during a loss-of performed the plant-specific uncertainty analysis re coolant accident are adequately predicted. This is an quired by the revised § 50.46 of 10 CFR Part 50, the idealized characterization of a best-estimate code. In licensee must demonstrate that the code and models practice, best-estimate codes may contain certain used are acceptable and applicable to the specific fa models that are simplified or that contain conserva cility over the intended operating range and must tism to some degree. This conservatism may be intro quantify the uncertainty in the specific application. duced for the following reasons:

General attributes expected in a best-estimate calcu lation are described here in Regulatory Position 1; 1. The model simplification or conservatism has little effect on the result, and therefore the special considerations for thermal-hydraulic best development of a better model is not estimate codes are presented in Regulatory Position justified.

2; and specific examples of features that are consid ered acceptable best-estimate models are given in 2. The uncertainty of a particular model is diffi Regulatory Position 3. Other models or correlations cult to determine, and only an upper bound will be considered acceptable if their technical basis is can be determined.

demonstrated with appropriate data and analysis. 3. The particular application does not require a A best-estimate model should provide a realistic totally best-estimate calculation, so a bias in calculation of the important parameters associated the calculation is acceptable.

with a particular phenomenon to the degree practical The introduction of conservative bias or simplifi with the currently available data and knowledge of cation in otherwise best-estimate codes should not,

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however, result in calculations that are unrealistic, or nodes to represent the system. Sensitivity studies that do not include important phenomena, or that and evaluations of the uncertainty introduced by contain bias and uncertainty that cannot be bounded. noding should be performed. Numerical methods Therefore, any calculational procedure determined treat time in a discrete manner, and the effect of to be a best-estimate code in the context of this guide time-step size should also be investigated.

or for use under paragraph 50.46(a)(i) should be compared with applicable experimental data to en 2.1.2 Computational Models sure that the calculation of important phenomena is A best-estimate code typically contains equations realistic. for conservation of mass, energy, and momentum of

2. CONSIDERATIONS FOR THERMAL the reactor coolant and noncondensible gases, if im portant (e.g., air, nitrogen). Energy equations are HYDRAULIC BEST-ESTIMATE CODES also used to calculate the temperature distribution in Some features that are acceptable for use in reactor system structures and in the fuel rods. The best-estimate codes are described in the following required complexity of these equations will vary de paragraphs. Models that address these features may pending on the phenomena that are to be calculated be used with the basic proviso that a specific model is and- the required accuracy of the calculation. NRC

acceptable if it has been compared with applicable staff experience with its own best-estimate computer experimental data and shown to provide reasonable codes has indicated that separate flow fields for dif predictions. Reference 7, "Compendium of ECCS ferent fluid phases, or types, and calculation of non Research for Realistic LOCA Analysis" equilibrium between phases may be required to calcu (NUREG-1230), provides a summary of the large ex late some important phenomena (e.g.,

perimental data base available, upon which best countercurrent flow, reflood heat transfer) to an ac estimate models may be based. While inclusion in ceptable accuracy. The NRC staff has also deter Reference 7 does not guarantee that the data or mined that certain phenomena require that the equa model will be acceptable, the report describes and tions be solved in multiple dimensions. However, references a large body of data generally applicable one-dimensional approximations to three to best-estimate models. NUREG-1230 also provides dimensional phenomena will be considered accept documentation of NRC studies of the effect of reac able if those approximations are properly justified.

tor power increase on risk, background information Other basic code features include equations of state on the ECCS rule, and a description of the methodol and other material properties. Sensitivity studies and ogy developed by NRC for estimating thermal comparisons to data should be performed to deter hydraulic transient code uncertainty. mine the importance of the simplifications used.

For any models or correlations used in a best 3. BEST-ESTIMATE CODE FEATURES

estimate code, sufficient justification must be pro vided to substantiate that the code performs ade 3.1 Initial and Boundary Conditions and Equipment Availability quately for the classes of transients to which it applied. In general, the features of best-estimate The heat generated by the fuel during a loss-of thermal-hydraulic transient codes have uncertaintie coolant accident depends on the power level of the associated with their use for predicting reactor syster reactor at the time of the loss-of-coolant accident and response. These uncertainties should be considered on the history of operation. The most limiting initial as part of the overall uncertainty analysis described in conditions expected over the life of the plant should Regulatory Position 4. be based on sensitivity studies. It is not necessary to assume initial conditions that could not occur in com

2.1 Basic Structure of Codes bination. For example, beginning-of-life peaking fac tors together with end-of-life decay heat do not re

2.1.1 Numerical Methods quire consideration. Given the assumed initial A best-estimate code uses a numerical scheme conditions, relevant factors such as the actual total for solving the equations used to predict the thermal power, actual peaking factors, and actual fuel condi hydraulic behavior of the reactor. The numerical tions should be calculated in a best-estimate manner.

scheme is, in itself, a complex process that can play The calculations performed should be represen an important role in the overall calculation. Careful tative of the spectrum of possible break sizes from the numerical modeling, sensitivity studies, and evalu full double-ended break of the largest pipe to a size ations of numerical error should be performed to en small enough that it can be shown that smaller breaks sure that the results of the calculations are represen are of less consequence than those already consid tative of the models used in the code. Numerical ered. The analyses should also include the effects of simulations of complex problems, such as those con longitudinal splits in the largest pipes, with the split sidered here, treat the geometry of the reactor in an area equal to twice the cross-sectional area of the approximate manner, making use of discrete volumes pipe. The range of break sizes considered should be

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sufficiently broad that -the system response as a func creep strain in cladding during steady-state operation, tion of break size is well enough defined so that inter reducing the gap between the fuel pellet and clad polations between calculations, without considering ding. Cladding creep is a function of fast neutron unexpected behavior between the break sizes, may flux (>1 MeV), cladding temperature, hoop stress, be made confidently. and material. Cladding materials may be cold worked and stress-relieved or fully recrystallized, and Other boundary and initial conditions and equip there is a significant difference in the magnitude of ment availability should be based on plant technical creepdown between these materials. During pellet specification limits. These other conditions include, cladding mechanical interaction, cladding experi but may not be limited to, availability and perform ences deformation from tensile creep, which is sig ance of equipment, automatic controls, and operator nificantly different from that caused by compressive actions. Appendix A to 10 CFR Part 50 requires that creep. An acceptable model for cladding tensile a single failure be considered when analyzing safety creep should be based on in-reactor tensile creep system performance and that the analysis consider data.

the effect of using only onsite power and only offsite power. Best-estimate fuel models will be considered ac ceptable provided the models include essential phe

3.2 Sources of Heat During a Loss-of-Coolant nomena identified above and provided their technical Accident basis is demonstrated with appropriate data and Models should account for the sources of heat analyses.

discussed below and the distribution of heat 3.2.1.1 Model Evaluation Procedure for production. Stored Energy and Heat Transfer in Fuel Rods. A

model to be used in ECCS evaluations to calculate

3.2.1 Initial Stored Energy of the Fuel internal fuel rod heat transfer should:

The steady-state temperature distribution and a. Be checked against several sets of relevant stored energy in the fuel before the postulated acci data, and dent should be calculated in a best-estimate manner b. Recognize the effects of fuel burnup, fuel for the assumed initial conditions, fuel conditions, pellet cracking and relocation, cladding and operating history. To accomplish this, the ther creep, and gas mixture conductivity.

mal conductivity of the fuel pellets and the thermal conductance of the gap between the fuel pellet and The model described by Lanning (Ref. 8) com

'* the cladding should be evaluated. Thermal conduc pared well to in-pile fuel temperature data. Best tivity of fuel is a function of temperature and is de estimate models will be considered acceptable graded by the presence of gases in crack voids be provided their technical basis is demonstrated with tween fuel fragments. An acceptable model for appropriate data and analyses.

thermal conductivity should be developed from the 3.2.1.2 Experimental Data for Stored in-pile test results for fuel centerline and off-center Energy in Fuel Rods and Heat Transfer. The temperatures, taking into account the conductivity of correlations and data of Reference 9 are acceptable gases in crack voids. for calculating the initial stored energy of the fuel and Thermal conductance of the fuel-cladding gap is subsequent heat transfer.

a strong function of hot gap size and of the composi 3.2.2 Fission Heat tion and pressure of the gases in the fuel rod. The Fission heat should be included in the calculation calculation of hot gap size should take into account and should be calculated using best-estimate reactiv U0 2 or mixed-oxide fuel swelling, densification, ity and reactor kinetics calculations. Shutdown reac creep, thermal expansion and fragment relocation, tivities resulting from temperatures and voids should and cladding creep. Fuel swelling is a function of also be calculated in a best-estimate manner. The temperature and burnup. Fuel densification is a func point kinetics formulation is considered an accept tion of burnup, temperature, and initial density. Den able best-estimate method for determining fission sification can result from hydrostatic stresses imposed heat in safety calculations for loss-of-coolant acci on fuel during pellet-cladding mechanical interaction dents. Other best-estimate models will be considered and should be considered. Fuel creep is a function of acceptable provided their technical basis is demon time, temperature, grain size, density, fission rate, strated with appropriate data and analyses. Control oxygen-to-metal ratio, and external stress. Fuel ther rod assembly insertion may be assumed if it is ex mal expansion represents dimensional changes in pected to occur.

unirradiated fuel pellets caused by changes in tem perature. An acceptable model for the above fuel pa 3.2.3 Decay of Actinides rameters should be based on in-pile and out-of-pile The heat from radioactive decay of actinides, in test data. Cladding creep introduces compressive cluding neptunium and plutonium generated during

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operation as well as isotopes of uranium, should be 3.2.6 Heat Transfer from Reactor calculated in accordance with fuel cycle history and Internals known radioactive properties. The actinide decay Heat transfer from piping, vessel walls, and inter heat chosen should be appropriate for the facility's nal hardware should be included in the calculation operating history. Best-estimate models will be con and should be calculated in a best-estimate manner.

sidered acceptable provided their technical basis is Heat transfer to channel boxes, control rods, guide demonstrated with appropriate data and analyses. tubes, and other in-core hardware should also be considered. Models will be considered acceptable

3.2.4 Fission Product Decay Heat provided their technical basis is demonstrated with appropriate data and analyses.

The heat generation rates from radioactive decay of fission products, including the effects of neutron 3.2.7 Primary to Secondary Heat Transfer capture, should be included in the calculation and (Not Applicable to Boiling Water should be calculated in a best-estimate manner. The Reactors)

energy release per fission (Q value) should also be Heat transferred between the primary and secon calculated in a best-estimate manner. Best-estimate methods will be considered acceptable provided their dary systems through the steam generators should be considered in the calculation and should be calcu technical basis is demonstrated with appropriate data lated in a best-estimate manner. Models will be con and analyses. The model in Reference 10 is consid sidered acceptable provided their technical basis is ered acceptable for calculating fission product decay demonstrated with appropriate data and analyses.

heat.

3.3 Reactor Core Thermal/Physical Parameters

3.2.4.1 Model Evaluation Procedure for Fission Product Decay Heat. The values of mean 3.3.1 Thermal Parameters for Swelling energy per fission (Q) and the models for actinide and Rupture of the Cladding and decay heat should be checked against a set of Fuel Rods relevant data. A calculation of thle swelling and rupture of the cladding resulting from the temperature distribution

3.2.5 Metal-Water Reaction Rate in the cladding and from the pressure difference be tween the inside and outside of the cladding, both as The rate of energy release, hydrogen generation, a function of time, should be included in the analysis and cladding oxidation from the reaction of the zir and should be performed in a best-estimate manner.

caloy cladding with steam should be calculated in a The degree of swelling and rupture should be taken best-estimate manner. Best-estimate models will be into account in the calculation of gap conductance, considered acceptable provided their technical basis cladding oxidation and embrittlement, hydrogen gen is demonstrated with appropriate data and analyses.

eration, and heat transfer and fluid flow outside of For rods calculated to rupture their cladding during the cladding. The calculation of fuel and cladding the loss-of-coolant accident, the oxidation of the in temperatures as a function of time should use values side of the cladding should be calculated in a best of gap conductance and other thermal parameters as estimate manner.

functions of temperature and time. Best-estimate methods to calculate the swelling of the cladding

3.2.5.1 Model Evaluation Procedure for should take into account spatially varying cladding Metal-Water Reaction Rate. Correlations to be used temperatures, heating rates, anisotropic material to calculate metal-water reaction rates at less than or properties, asymmetric deformation of cladding, and equal to 1900'F should: fuel rod thermal and mechanical parameters. Best estimate methods will be considered acceptable pro a. Be checked against a set of relevant data, vided their technical basis is demonstrated with ap and propriate data and analyses.

3.3.2 Other Core Thermal Parameters b. Recognize the effects of steam pressure, pre oxidation of the cladding, deformation dur As necessary and appropriate, physical and ing oxidation, and internal oxidation from chemical changes in in-core materials (e.g., eutectic both steam and U0 2 fuel. formation, phase change, or other phenomena caused by material interaction) should be accounted The data of Reference 11 are considered accept for in the reactor core thermal analysis. Best-estimate able for calculating the rates of energy release, hydro models will be considered acceptable provided their gen generation, and cladding oxidation for cladding technical basis is demonstrated with appropriate data temperatures greater than 1900 'F. and analyses.

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3.4 Blowdown Phenomena For critical flow from small breaks under strati fied conditions, currently acceptable test data for as

3.4.1 Break Characteristics and Flow sessing models and codes include those reported by:

In analyses of hypothetical loss-of-coolant acci dents, a spectrum of possible break sizes should be "* Anderson and Owca (Ref. 21)

-- ' considered, as indicated in Regulatory Position 3.1.

The discharge flow rate should be calculated with a

"* Reimann and Khan (Ref. 22)

critical flow rate model that considers the fluid "* Schrock et al. (Refs. 23 and 24)

conditions at the break location, upstream and down stream pressures, and break geometry. The critical 3.4.2 ECC Bypass flow model should be justified by comparison to ap The best-estimate code should contain a calcula plicable experimental data over a range of conditions tion of the amount of injected cooling water that by for which the model is applied. The model should be passes the vessel during the blowdown phase of the a best-estimate calculation, with uncertainty in the loss-of-coolant accident. The calculation of ECC by critical flow rate included as part of the uncertainty pass should be a best-estimate calculation using evaluation. Best-estimate models will be considered analyses and comparisons with applicable experimen acceptable provided their technical basis is demon tal data. Although it is clear that the dominant proc strated with appropriate data and analyses. esses governing ECC bypass are multidimensional, single-dimensional approximations justified through

3.4.1.1 Model Evaluation Procedure for sufficient analysis and data may be acceptable. Best Discharge Flow Rate. Critical flow models to be estimate methods will be considered acceptable pro employed in ECCS evaluations should:

vided their technical basis is demonstrated with ap a. Be checked against an acceptable set of rele propriate data and analyses. Cooling water that is not vant data, expelled, but remains in piping or is stored in parts of the vessel, should be calculated in a best-estimate b. Recognize thermal nonequilibrium conditions manner based on applicable experimental data.

when the fluid is subcooled, and

3.4.2.1 Model Evaluation Procedure for ECC Bypass. A correlation or model to be used to c. Provide a means of transition from nonequi librium to equilibrium conditions. evaluate ECC bypass should:

The uncertainties and bias of a correlation or a. Be checked against an acceptable set of rele model used to calculate critical flow should be stated, vant data, and as well as their range of applicability.

b. Recognize the effects of pressure, liquid sub The mechanistic thermal nonequilibrium and slip cooling, fluid conditions, hot walls, and sys model of Richter (Ref. 12) compares well to small tem geometry.

and large-scale test data (Ref. 13). Uncertainties and bias in the correlations or models used to calculate ECC bypass should be

3.4.1.2 Experimental Data for Discharge stated, as well as the range of their applicability.

Flow Rate. An acceptable set of relevant critical flow data should cover the fluid conditions, geometries, For scaled-down PWR downcomers, correlations and types of breaks pertinent to light-water reactor by Beckner and Reyes (Ref. 25) compared well to loss-of-coolant accidents. The following tests should the bypass data of References 26 and 27. Correla be considered in establishing an acceptable set of tions of Sun (Ref. 28) and Jones (Ref. 29) compare relevant data: well to counter-current flow limiting (CCFL) test data of interest to BWRs.

"* Marviken tests (Ref. 14)

3.4.2.2 Experimental Data for ECC Bypass.

"* Moby Dick experiments (Ref. 15) The following tests should be considered in establishing a set of data for scaled-down PWR

"* Brookhaven critical flashing flows in nozzles downcomers:

(Ref. 16)

"* Battelle Columbus test (Ref. 26)

"* Sozzi-Sutherland tests (Ref. 17)

"* Creare test (Refs. 26 and 27)

"* Edwards experiments (Ref. 18) For a full-scale PWR vessel, ECC bypass data will become available from the forthcoming upper

"* Super Moby Dick experiments (Refs. 19 and plenum test facility (UPTF) experiments performed

20) as part of the 2D/3D program sponsored by the

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Federal Republic of Germany, Japan, and the United * Horizontal tubes States. GE tests (Refs. 35 and 36)

For BWRs, the following test should be consid * Rod bundles ered in establishing an acceptable set of relevant GE tests (Ref. 37)

data:

3.7 Momentum Equation

0 SSTF test data (Refs. 30 through 32) The following effects should be taken into account in the two-phase conservation of momentum

3.5 Noding Near the Break and ECCS Injection equation: (1) temporal change in momentum, Point (2) momentum convection, (3) area change momen The break location and ECCS injection point are tum flux, (4) momentum change due to compressibil areas of high fluid velocity and complex fluid flow ity, (5) pressure loss resulting from wall friction, and contain phenomena that are often difficult to cal (6) pressure loss resulting from area change, and culate. The results of these calculations are often (7) gravitational acceleration. Best-estimate models highly dependent on the noding. Sufficient sensitivity will be considered acceptable provided their technical studies should be performed on the noding and other basis is demonstrated with appropriate data and important parameters to ensure that the calculations analyses.

provide realistic results.

3.8 Critical Heat Flux

3.6 Frictional Pressure Drop Best-estimate models developed from appropri ate steady-state or transient experimental data should The frictional losses in pipes and other compo be used in calculating critical heat flux (CHF) during nents should be calculated using models that include loss-of-coolant accidents. The codes in which these variation of friction factor with Reynolds number and models are used should contain suitable checks to en account for two-phase flow effects on friction. Best sure that the range of conditions over which these estimate models will be considered acceptable pro correlations are used are within those intended. Re vided their technical basis is demonstrated with ap search has shown that CHF is highly dependent on propriate data and analyses.

the fuel rod geometry, local heat flux, and fluid con

3.6.1 ditions. After CHF is predicted at an axial fuel rod Model Evaluation Procedure for Frictional Pressure Drop location, the calculation may use nucleate boiling heat transfer correlations if the calculated local fluid A model for frictional pressure drop to be used and surface conditions justify the reestablishment of in ECCS evaluation should: nucleate boiling. Best-estimate models will be consid ered acceptable provided their technical basis is dem a. Be checked against a set of relevant data, onstrated with appropriate data and analyses.

and

3.9 Post-CHF Blowdown Heat Transfer b. Be consistent with models used for calculat Models of heat transfer from the fuel to the sur ing gravitational and acceleration pressure rounding fluid in the post-CHF regimes of transition drops. If void fraction models or correlations and film boiling should be best-estimate models based used to calculate the three components of the on comparison to applicable steady-state or transient total pressure drop differ one from another, a quantitative justification must be provided. data. Any model should be evaluated to demonstrate that it provides acceptable results over the applicable Uncertainties and bias of a correlation or model ranges. Best-estimate models will be considered ac should be stated as well as the range of applicability. ceptable provided their technical basis is demon strated with appropriate data and analyses.

3.6.2 Experimental Data for Frictional Pressure Drop 3.9.1 Model Evaluation Procedure for An acceptable set of relevant data should cover, Post-CHF Heat Transfer as far as possible, the ranges of parameters (mass A model to be used in ECCS evaluation to calcu flux, quality, pressure, fluid physical properties, late post-CHF heat transfer from rod bundles should:

roughness, and geometries) that are found in actual plant applications. The following tests should be con a. Be checked against an acceptable set of rele sidered in establishing an acceptable set of relevant vant data, and data:

b. Recognize effects of liquid entrainment, ther

• Vertical tubes mal radiation, thermal nonequilibrium, low CISE test (Refs. 33 and 34) and high mass flow rates, low and high power

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densities, and saturated and subcooled inlet variables. Pr represents the Prandtl number, and Nu conditions. is the Nusselt number. The physical properties may The uncertainties and bias of models or correla be defined as wall, film, or vapor values.

tions used to calculate post-CHF heat transfer should A distinction from, and transition to, laminar be stated as well as the range of their applicability. convection (i.e., Re <2000) should be made, with a value of the laminar heat transfer for rod bundles

3.9.2 Experimental Data for Post-CHF that is appropriate for the applicable bundle geometry Heat Transfer and flow conditions.

The acceptable set of relevant data should cover Other forms and values, depending on the bun power densities, mass flow rates, fluid conditions, dle geometry and flow conditions, are also appropri and rod bundle geometries pertinent to light-water re ate.

actor designs and applications. The following tests should be considered in establishing an acceptable set 3.9.3.2 Experimental Data for Heat of relevant data: Transfer from Uncovered Rod Bundles. An acceptable set of relevant data for post-CHF heat

"* ORNL tests (Refs. 38 and 39) transfer from uncovered rod bundles should cover power densities, fluid conditions, and rod bundle

"* FLECHT-SEASET tests (Ref. 40) geometries pertinent to light-water reactor design and application. The following tests should be considered

"* INEL tests (Ref. 41) in establishing an acceptable set of relevant data:

"* ORNL data base (Ref. 42) "* ORNL-THTF tests (Refs. 43 and 44)

"* FLECHT-SEASET tests (Refs. 45 and 46)

3.9.3 Post-CHF Heat Transfer from Uncovered Bundles "* ORNL data base (Ref. 42)

During some time periods of small-break loss-of

3.10 Pump Modeling coolant accidents and during portions of large breaks prior to reflood, partial or complete core uncovering The characteristics of rotating primary system may be calculated to occur. Under these circum pumps should be derived from a best-estimate dy stances, special considerations for calculating heat namic model that includes momentum transfer be transfer are necessary. tween the fluid and the rotating member, with vari able pump speed as a function of tim

e. The pump

3.9.3.1 Model Evaluation Procedures for model resistance and other empirical terms should be Heat Transfer from Uncovered Rod Bundles. A justified through comparisons with applicable data.

correlation to be used in ECCS evaluations to The pump model for the two-phase region should be calculate heat transfer from uncovered rod bundles verified by comparison to applicable two-phase per should: formance data. Pump coastdown following loss of power should be treated in a best-estimate manner. A

a. Be checked against an acceptable set of rele locked rotor following a large-break loss-of-coolant vant data, and accident need not be assumed unless it is calculated to occur. Best-estimate models will be considered ac b. Recognize the effects of radiation and of laminar, transition, and turbulent flows. ceptable provided their technical basis is demon strated with appropriate data and analyses.

Uncertainties and bias in the models and correla tions used to calculate post-CHF heat transfer should 3.11 Core Flow Distribution During Blowdown be stated, as should the range of their applicability. The core flow through the hottest region of the The correlation derived should include a stated core during the blowdown should be calculated as a procedure for correcting for radiative heat transfer function of time. For the purpose of these calcula and for estimating the vapor temperatures. The Hot tions, the hottest region of the core should not be greater than the size of one fuel assembly. Calcula tel procedure cited in Reference 43 is a satisfactory example. tions of the flow in the hot region should take into account any cross-flow between regions and any flow The turbulent correlation may be of the general blockage calculated to occur during the blowdown as form: a result of cladding swelling or rupture. The numeri cal scheme should ensure that unrealistic oscillations Nu = A Rem Prn of the calculated flow do not result. Best-estimate for higher Reynolds numbers (Re), where the coeffi models will be considered acceptable provided their cients A, m, and n are modifications from the basic technical basis is demonstrated with appropriate data Dittus-Boelter form and may be functions of other and analyses.

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3.12 Post-Blowdown Phenomena effects of the compressed gas in the accumulator fol lowing accumulator water discharge should be in

3.12.1 Containment Pressure cluded in the calculation. Any model or code used The containment pressure used for evaluating for this calculation should be assessed against applica cooling effectiveness during the post-blowdown phase ble experimental data. Reference 7 describes a large of a loss-of-coolant accident should be calculated in a body of refill/reflood thermal-hydraulic data obtained best-estimate manner and should include the effects from the 2D/3D program that is appropriate for of containment heat sinks. The calculation should in consideration.

clude the effects of operation of all pressure-reducing equipment assumed to be available. Best-estimate 3.12.2.2 Experimental Data for Post models will be considered acceptable provided their Blowdown Thermal Hydraulics. The following tests technical basis is demonstrated with appropriate data should be considered when establishing an acceptable and analyses. set of relevant data:

3.12.2 Calculation of Post-Blowdown

"* GE tests (Refs. 48 and 51)

Thermal Hydraulics for Pressurized

"* ORNL tests (Refs. 43 and 49)

Water Reactors The refilling of the reactor vessel and the ulti "* FLECHT-SEASET test (Ref. 45)

mate reflooding of the core should be calculated by a best-estimate model that takes into consideration the "* THETIS tests (Ref. 50)

thermal and hydraulic characteristics of the core, the emergency core cooling systems, and the primary and 3.12.3 Steam Interaction with Emergency secondary reactor systems. The model should be ca Core Cooling Water in Pressurized pable of calculating the two-phase level in the reactor Water Reactors during the postulated transient. Best-estimate models The thermal-hydraulic interaction between the will be considered acceptable provided their technical steam or two-phase fluid and the emergency core basis is demonstrated with appropriate data and cooling water should be taken into account in calcu analyses. lating the core thermal hydraulics and the steam flow through the reactor coolant pipes during the time the

3.12.2.1 Model Evaluation Procedures for accumulators are discharging water. Best-estimate Post-Blowdown Thermal Hydraulics. A correlation models will be considered acceptable provided their or model to be used in ECCS evaluation to calculate technical basis is demonstrated with appropriate data level swell should be checked against an acceptable and analyses.

set of relevant data and should recognize the effects of depressurization, boil-off, power level, fluid 3.12.4 Post-Blowdown Heat Transfer for conditions, and system geometry. Pressurized Water Reactors The correlation proposed by Chexal, Horowitz, During refilling of the reactor vessel and ultimate and Lellouche (Ref. 47) provides acceptable results reflooding of the core, the heat transfer calculations when compared to experimental data reported in should be based on a best-estimate calculation of the References 43, 48, 49, and 50. fluid flow through the core, accounting for unique Uncertainties and bias of a correlation or model emergency core cooling systems. The calculations used to calculate level swell should be stated, as should also include the effects of any flow blockage should the range of applicability. calculated to occur as a result of cladding swelling or rupture. Heat transfer calculations that account for The primary coolant pumps should be assumed two-phase conditions in the core during refilling of to be operating in the expected manner, based on the the reactor vessel should be justified through com assumptions of Regulatory Position 3. 1, when calcu parisons with experimental data. Best-estimate mod lating the resistance offered by the pumps to fluid els will be considered acceptable provided their tech flow. Models will be considered acceptable provided nical basis is demonstrated through comparison with their technical basis is demonstrated through com appropriate data and analyses.

parison with appropriate data and analyses.

The FLECHT-SEASET tests (Refs. 40, 45, and The total fluid flow leaving the core exit (car 46) should be considered when establishing an ac ryover) should be calculated using a best-estimate ceptable set of relevant data. Reference 7 contains model that includes the effect of cross-flow on car extensive information regarding a large amount of ex ryover and core fluid distribution. Thermal-hydraulic perimental reflood heat transfer data. This informa phenomena associated with unique emergency core tion should also be considered when developing and cooling systems, such as upper plenum injection and assessing models. The results from the 2D/3D pro upper head injection, should be accounted for. The gram are particularly relevant.

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3.13 Convective Heat Transfer Coefficients for different from those phenomena that would occur Boiling Water Reactor Rods Under Spray during a large-break loss-of-coolant accident. The Cooling distribution of liquid throughout the reactor system, in addition to the total liquid inventory, is of in Models will be considered acceptable provided creased importance for the small-break loss-of their technical bases can be justified with appropriate coolant accident. A number of special factors must

- data and analyses. These models should contain the be given increased consideration in small-break loss following:

of-coolant accident calculations to correctly predict

1. Following the blowdown period, convective phenomena influenced by the liquid inventory heat transfer coefficients should be deter distribution.

mined based on the calculated fluid condi Break flow may be greatly influenced by the loca tions and heat transfer modes within the bun tion and specific geometry of the break. For a break dle and on the calculated rod temperatures. in a horizontal pipe containing stratified flow, the quality of the break flow will be a strong function of

2. During the period following the flashing of the assumed location of the break on the pipe (e.g.,

the lower plenum fluid, but prior to ECCS

initiation, heat transfer models should in top or bottom). Small-break loss-of-coolant accident clude cooling by steam flow or by a two calculations should, therefore, include various as phase mixture, if calculated to occur. sumed break locations in the spectrum of breaks ana lyzed. The assumed operating state of the reactor

3. Following initiation of ECCS flow, but prior coolant pump will also influence the distribution of to reflooding, heat transfer should be based liquid throughout the system and the amount of liquid on the actual calculated bundle fluid condi lost through the break.

tions and best-estimate heat transfer models The pump operation assumptions used in the cal that take into account rod-to-rod variations in heat transfer. culations should be the most likely, based on operat ing procedures, with appropriate consideration of the

4. After the two-phase reflood level reaches the uncertainty of the pump operation during an actual level under consideration, a best-estimate event. Level depression in the core region and subse heat transfer model should be used. This quent core heatup may be influenced by liquid model should include the effects of any flow holdup in the steam generator tubes, manometric ef blockage calculated to occur as a result of fects of liquid in the piping and loop seal region, and cladding swelling or rupture. liquid levels relative to vent paths for steam through upper plenum bypass flow paths and vent valves.

5. Thermal-hydraulic models that do not calcu Steam generator heat transfer under "reflux" or late multiple channel effects should be com

"boiler-condensor" modes of operation may also pared with applicable experimental data or strongly influence core inventory through level de more detailed calculations to ensure that all adequately pression and the effect on total system pressure and, important phenomena are calculated. thus, on ECCS flow. These phenomena should be carefully considered in the calculatio

n. Sensitivity

3.14 Boiling Water Reactor Channel Box Under studies of the importance of these effects should be Spray Cooling performed for use in the uncertainty evaluation.

Following the blowdown period, heat transfer Heat transfer from an uncovered core under from the channel box and wetting of the channel box high-pressure conditions typical during a small-break should be based on the calculated fluid conditions on loss-of-coolant accident may include contributions both sides of the channel box and should make use from both convective and radiation heat transfer to of best-estimate heat transfer and rewetting models the steam. Models will be considered acceptable pro that have been compared with applicable experimen vided their technical basis is demonstrated through tal data. comparison with appropriate data and analyses. Spe cific guidance regarding uncovered bundle heat trans

3.15 Special Considerations for a Small-Break fer is given in Regulatory Position 3.9.3.

Loss-of-Coolant Accident in Pressurized Water Reactors 3.16 Other Features of Best-Estimate Codes The slower small-break loss-of-coolant accident No list of best-estimate code features could be leads to fluid conditions characterized by separation all-inclusive, because the important features of a of the fluid phases versus the more homogeneous best-estimate code may vary depending on the tran fluid conditions that would result from rapid large sient to be calculated and the required accuracy of break loss-of-coolant accident transients. Phenomena the calculation. Because of this, no attempt has been

~ that would occur in a PWR during a small-break loss made to construct an exhaustive list of best-estimate of-coolant accident would, therefore, be significantly code features. Rather, features that were identified as

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important for inclusion in Appendix K were used as a tainty to ensure that meaningful comparisons are be basis for the above list. These features are not neces ing made.

sarily any more or less important than other code fea tures, but were highlighted because it is necessary to 4. ESTIMATION OF OVERALL

give specific examples of how current best-estimate CALCULATIONAL UNCERTAINTY

models may vary from methods used traditionally in evaluation model codes using the various Appendix K 4.1 General conservatisms. In addition, models have not been The term "uncertainty," when applied to best included for areas in which the best model would be estimate thermal-hydraulic transient codes, is used at highly dependent on the specific plant design or the two levels. At the lower or more detailed level, the specific transient under consideration. term refers to the degree to which an individual model, correlation, or method used within the code The NRC staff believes that good examples of represents the physical phenomenon it addresses.

best-estimate thermal-hydraulic transient codes are These individual uncertainties, when taken together, those developed by the NRC (e.g., TRAC-PWR, comprise the "code uncertainty."

TRAC-BWR, RELAPS, COBRA, and FRAP). Al though these codes are subject to further improve The combined uncertainty associated with indi ment, based on their ongoing use and assessment, vidual models (i.e., code uncertainty) within the best they currently provide reasonable best-estimate cal estimate codes does not account for all of the uncer culations of the loss-of-coolant accident in a full-scale tainty associated with the model's use. In addition to light-water reactor. This is substantiated through the the code uncertainty, various other sources of uncer code development and assessment literature gener tainty are introduced when attempting to use best ated by the NRC and its contractors over the past estimate codes to predict full-scale plant several years.

thermal-hydraulic response. These sources include uncertainty associated with the experimental data It is possible, however, to describe in general used in the code assessment process (including appli how other features of best-estimate codes should be cability of the data to full-scale reactors), the input constructed. Two basic criteria should be applied, boundary and initial conditions, and the fuel behav completeness and comparisons to experimental data. ior. Additional sources of uncertainty stem from the use of simplifying assumptions and approximations. A

3.16.1 Completeness careful statement of these assumptions and approxi Best-estimate codes should contain models in mations should be made, and the uncertainty associ sufficient detail to predict phenomena that are impor ated with them should be taken into account. There tant to demonstrate compliance with the acceptance fore, the "overall calculational uncertainty" is criteria specified in paragraph 50.46(b) of 10 CFR defined as the uncertainty arrived at when all the Part 50 (e.g., peak cladding temperature). Simplifi contributions from the sources identified above, in cations are acceptable as lopg as code uncertainties cluding the code uncertainty, are taken into account.

or biases do not become so large that they cast doubt A 95% probability level is considered acceptable on the actual behavior that would occur or on the to the NRC staff for comparison of best-estimate pre true effect of assumed initial and boundary condi dictions to the applicable limits of paragraph tions (e.g., equipment sizing, safety system settings). 50.46(b) of 10 CFR Part 50 to meet the requirement Comparisons of the overall calculations to integral ex of paragraph 50.46(a)(1)(i) to show that there is a periments should be performed to ensure that impor high probability that the criteria will not be exceeded.

tant phenomena can be predicted and to help in The basis for selecting the 95% probability level is making judgments on the effect of code simplifica primarily for consistency with standard engineering tions. Consideration should also be given to the un practice in regulatory matters involving thermal hy certainty and validity of the experiment to ensure that draulics. Many parameters, most notably the depar meaningful comparisons are being made. ture from nucleate boiling ratio (DNBR), have been found acceptable by the NRC staff in the past at the

3.16.2 Data Comparisons 95% probability level.

Individual best-estimate models should be com This 95% probability level would also be applied pared to applicable experimental data to ensure that to small-break loss-of-coolant accidents, which have realistic behavior is predicted and that relevant ex a higher probability than large breaks. The dominant perimental variables are included. Uncertainty analy factors influencing risk from small-break loss-of ses are required to ensure that a major bias does not coolant accidents include equipment availability and exist in the models and that the model uncertainty is operator actions. Calculational uncertainties are small enough to provide a realistic estimate of the ef much less important than factors such as operator fect of important experimental variables. Uncertainty recognition of the event, the availability of equip analyses should also consider experimental uncer- ment, and the correct use of this equipment.. The use

1.157-12

of a best-estimate calculation with reasonable and best-estimate capability of the code. For large-break quantifiable uncertainty is expected to provide a re loss-of-coolant accidents, the most important key pa duction in the overall risk from a small-break loss-of rameter is peak cladding temperature, which is ad coolant accident by providing more realistic calcula dressed by one of the criteria of paragraph 50.46(b)

tions with which to evaluate operator guidelines and and has a direct influence on the other criteria. In

-' determine the true effect of equipment availability. addition, a code uncertainty evaluation should be performed for other important parameters for the Regulatory Position 3 provides a description of transient of interest to evaluate compensating errors.

the features that should be included in the overall For small-break loss-of-coolant accidents, the clad code uncertainty evaluation that is called for in para ding temperature response is the most important graph 50.46(a) (1). This uncertainty evaluation parameter; however, the ability of the codes to pre should make use of probabilistic and statistical meth dict overall system mass and reactor vessel inventory ods to determine the code uncertainty. For a calcula distribution should also be statistically examined.

tion of this complexity, a completely rigorous mathe In evaluating the code uncertainty, it will be nec matical treatment is neither practical nor required. In essary to evaluate the code's predictive ability over many cases, approximations and assumptions may be several time intervals, since different processes and made to make the overall calculational uncertainty phenomena occur at different intervals. For example, evaluation possible. A careful statement of these as in large-break loss-of-coolant accident evaluations, sumptions and approximations should be made so separate code uncertainties may be required for the that the NRC staff may make a judgment as to the peak cladding temperature during the blowdown and validity of the uncertainty evaluation. The purpose of post-blowdown phases. Justification for treating these the uncertainty evaluation is to provide assurance uncertainties individually or methods for combining that for postulated loss-of-coolant accidents a given them should be provided.

plant will not, with a probability of 95% or more, ex ceed the applicable limits specified in paragraph The experimental information used to determine

50.46(b). code uncertainty will usually be obtained from facili ties that are much smaller than nuclear power reac

4.2 Code Uncertainty tors. Applicability of these results should be justified for larger scales. The effects of scale can be assessed This regulatory guide makes a distinction be through comparisons to available large-scale separate tween the terms "code uncertainty" and "overall cal effects tests and through comparison to integral tests culational uncertainty." The latter term is defined from various sized facilities. If there are scaling prob

"*-' above and includes the contributions to the uncer lems, particularly if predictions are nonconservative, tainty described in Regulatory Positions 4.2 and 4.3. the code should be improved for large-scale plants The components of the code uncertainty (i.e., the (i.e., nuclear reactors). Codes not having scaling ca contribution to the overall uncertainty from the mod pability will not be acceptable if their predictions are els and numerical methods used) are described in nonconservative.

this section.

4.3 Other Sources of Uncertainty Code uncertainty should be evaluated through di When a best-estimate methodology is used to rect data comparison with relevant integral systems predict reactor transients, sources of uncertainty and separate-effects experiments at different scales. other than the limitations in the individual models In this manner, an estimate of the uncertainty attrib and numerical methods (i.e., code uncertainty) are utable to the combined effect of the models and cor introduced. The following contributors to the overall relations within the code can be obtained for all calculational uncertainty should also be considered in scales and for different phenomena. Comparison to a the uncertainty analysis.

sufficient number of integral systems experiments, from different test facilities and different scales, 4.3.1 Initial and Boundary Conditions should be made to ensure that a reasonable estimate and Equipment Availability of code uncertainty and bias has been obtained. When a plant input model is prepared, certain When necessary, separate-effects experiments should relationships describing the plant boundary and initial be used to establish code uncertainty for specific phe conditions and the availability and performance of nomena (e.g., comparisons to Cylindrical Core Test equipment are defined. These include factors such as Facility data to ascertain code uncertainty in model initial power level, pump performance, valve activa ing upper plenum injection performance). Code com tion times, and control systems functioning. Uncer parisons should account for limitations of the meas tainties associated with the boundary and initial con urements and calibration errors. ditions and the characterization and performance of These comparisons should be performed for im equipment should be accounted for in the uncertainty portant key parameters to demonstrate the overall evaluation. It is also acceptable to limit the variables

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to be considered by setting their values to conserva and justification of the statistical distribution and in tive bounds. the estimation of its statistical parameter

s. If a normal

4.3.2 distribution is selected and justified, the probability Fuel Behavior limit can be conservatively calculated using two Variability of the results of plant transient calcu standard deviations. The added conservatism of the lations can result from uncertainties associated with two standard deviations compared to the 95th fuel behavior, which are not included in the compari percentile is used to account for uncertainty in the sons of code results with integral experiments since probability distribution. Other techniques that most integral tests use electrically heated rods. This account for the uncertainty in a more detailed uncertainty includes many effects such as fuel con manner may be used. These techniques may require ductivity, gap width, gap conductivity, and peaking the use of confidence levels, which are not required factors. These uncertainties should be quantified and by the above approach.

used in the determination of the overall calculational uncertainty. The evaluation of the peak cladding temperature at the 95% probability level need only be performed

4.3.3 Other Variables for the worst-case break identified by the break spec There may be individual models within the best trum analysis in order to demonstrate conformance estimate code whose effect may not have been evalu with paragraph 50.46(b). However, in order to use ated by the comparison to the integral systems data. this approach, justification must be provided that For example, since most integral systems experiments demonstrates that the overall calculational uncer use electrically heated rods, uncertainties associated tainty for the worst case bounds the uncertainty for with the prediction of core decay heat and cladding other breaks within the spectrum. It may be neces metal/water reaction have not been evaluated. In ad sary to perform separate uncertainty evaluations for dition, to demonstrate the overall adequacy of the large- and small-break loss-of-coolant accidents be predictive ability of the best-estimate code, it may be cause of the substantial difference in system thermal necessary to use empirically arrived at break hydraulic behavior.

discharge coefficients to obtain a reasonable break The revised paragraph 50.46(a) (1)(i) requires flow. The uncertainties in the individual models that that it be shown with a high probability that none of have not been evaluated by comparison to integral the criteria of paragraph 50.46(b) will be exceeded, systems data should be quantified and used in the and is not limited to the peak cladding temperature determination of overall code uncertainty. criterion. However, since the other criteria are strongly dependent on peak. cladding temperature,

4.4 Statistical Treatment of Overall explicit consideration of the probability of exceeding Calculational Uncertainty the other criteria may not be required if it can be The methodology used to obtain an estimate of demonstrated that meeting the temperature criterion the overall calculational uncertainty at the 95% prob at the 95% probability level ensures with an equal or ability limit should be provided and justified. If linear greater probability that the other criteria will not be independence is assumed, suitable justification exceeded.

should be provided. The influence of the individual parameters on code uncertainty should be examined 4.5 NRC Approach to LOCA Uncertainty by making comparisons to relevant experimental Evaluation data. Justification should be provided for the as Chapter 4 of the "Compendium of ECCS

sumed distribution of the parameter and the range Research for Realistic LOCA Analysis" (Ref. 7)

considered. presents a methodology that has been used for evaluating the overall calculational uncertainty in In reality, the true statistical distribution for the peak cladding temperature predictions for key parameters (e.g., peak cladding temperature) is best-estimate thermal-hydraulic transient codes that unknown. The choice of a statistical distribution the NRC has developed.

should be verified using applicable engineering data and information. The statistical parameters appropri ate for that distribution should be estimated using

D. IMPLEMENTATION

available data and results of engineering analyses. The purpose of this section is to provide informa Supporting documentation should be provided for tion to applicants and licensees regarding the NRC

this selection process. These estimated values are as staff's plans for using this regulatory guide.

sumed to be the true values of the statistical param Licensees and applicants may propose means eters of the distribution. With these assumptions, an other than those specified by the provisions of the upper one-sided probability limit can be calculated at Regulatory Position of this guide for meeting applica the 95% level. As the probability limit approaches ble regulations. This guide has been approved for use

2200'F, more care must be taken in the selection by the NRC staff as an acceptable means of

1.157-14

complying with the Commission's regulations and for allow the use of realistic models as an alter evaluating submittals in the following categories: native to the features of Appendix K of

10 CFR Part 50.

1. Construction permit applicants that choose to make use of the provisions of § 50.46 that allow the use of realistic models as an alter 3. Operating reactor licensees will not be native to the features of Appendix K of evaluated against the provisions of this guide

10 CFR Part 50. except for new submittals that make use of the provisions of § 50.46 that allow the use

2. Operating license applicants that choose to of realistic models as an alternative to the make use of the provisions of § 50.46 that features of Appendix K of 10 CFR Part 50.

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2. Idaho National Engineering Laboratory, Research Institute, Palo Alto, CA, April 1981.

"TRAC-BD1/MOD1: An Advanced Best Esti mate Computer Program for Boiling Water Re 13. D. Abdollahian et al., "Critical Flow Data Re actor Transient Analysis," NUREG/CR-3633, 4 view and Analysis," Report NP-2192, Electric Vols. (EGG-2294), April 1984. Power Research Institute, Palo Alto, CA, Janu ary 1982.

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"RELAP5/MOD2 Code Manual," Vols. 1 & 2, 14. USNRC, "The Marviken Full Scale Critical NUREG/CR-4312, August 1985. (Available in Flow Tests, Summary Report," (Joint Reactor the NRC Public Document Room.) Safety Experiments in the Marviken Power Sta tion, Sweden), NUREG/CR-2671, May 1982.

4. Pacific Northwest Laboratory, "COBRA/TRAC

- A Thermal-Hydraulics Code for Transient 15. M. Reocreux, "Contribution to the Study of Analysis of Nuclear Reactor Vessels and Pri Two-Phase Steam-Water Critical Flow," Ph.D.

mary Coolant Systems," NUREG/CR-3046, 5 Thesis, L'Universite Scientifique Medicale de Vols. (PNL-4385), March 1983. Grenoble, 1974. (English translation available from NTIS, LIB/Trarns-576.)

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Code for the Transient Analysis of Oxide Fuel Study of Nonequilibrium Flashing of Water in a Rods," NUREG/CR-2148 (EG&G, EGG-2104), Converging-Diverging Nozzle," NUREG/

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CR-1864, Vols. 1-2 (Brookhaven National Laboratory, BNL-NUREG-51317), March

6. G. A. Berna et al., "FRAPCON-2: A Com 1982.

puter Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel 17. G. L. Sozzi and W. A. Sutherland, "Critical Rods," NUREG/CR-1845, January 1981. Flow of Saturated and Subcooled Water at High Pressure," General Electric Company, GE Re

7. "Compendium of ECCS Research for Realistic port NEDO-13418, 1975. (Available in the LOCA Analysis," NUREG-1230, December NRC Public Document Room.)

1988.

18. R. A. Edwards and T. P. O'Brien, "Studies of

8. D. Lanning and M. Cunningham, "Trends in Phenomena Connected with the Depressuriza Thermal Calculations for Light Water Reactor tion of Water Reactois," Nuclear Energy (Jour Fuel (1971-1981)," in Ninth Water Reactor nal of the British Nuclear Energy Society), Vol.

Safety Research Information Meeting, USNRC, 9, No. 2, April 1970.

NUREG/CP-0024, Vol. 3, March 1982.

19. Commissariat a L'Energie Atomique, C. Jean

9. Idaho National Engineering Laboratory, dey et al., "Auto vaporisation d'ecoulements

"MATPRO Version 11 (Revision 2): A Hand eau/vapeur," Report TT, No. 163, Centre book of Materials Properties for Use in the d'Etudes Nucleaires de Grenoble, Dept. des Analysis of Light-Water Reactor Fuel Rod Be Reacteurs a Eau, Service des Transferts Ther havior," NUREG/CR-0497, Rev. 2, August miques, Grenoble, France, July 1981. (Copies

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10. American Nuclear Society, "American Na Cedex, France.)

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1979. (ANS, 555 North Kensington Avenue, La Flows in a Short Super Moby Dick Pipe," Rap Grange Park, Illinois 60525.) port TT/SETRE/7 1, Centre d'Etudes Nucleaires de Grenoble, Grenoble, France, September

11. J. V. Cathcart et al., "Zirconium Metal-Water 1983. NRC Translation 1401 available from the Oxidation Kinetics: IV Reaction Rate Studies," NRC Public Document Room (52 FR 6334), ac Oak Ridge National Laboratory, ORNL/ cession number 8704060298.

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21. J. L. Anderson and W. A. Owca, "Data Report 31. D. G. Schumacher et al., "BWR Refill-Reflood for the TPFL Tee/Critical Flow Experiments," Program Task 4.4 - CCFL/Refill System Ef NUREG/CR-4164 (EG&G Idaho, Inc., fects Tests (300 Sector). SSTF Systems Re EGG-2377), November 1985. sponse Test Results," NUREG/CR-2568 (Gen eral Electric Company, GEAP-22046, EPRI

S 22. J. Reimann and M. Khan, "Flow Through a NP-2374), April 1983.

Small Break at the Bottom of a Large Pipe with Stratified Flow," Nuclear Science and Engineer 32. J. A. Findlay, "BWR Refill-Reflood Program ing, Vol. 88, pp. 297-310, November 1984. Task 4.4 - CCFL/Refill System Effects Tests

(300 Sector). Evaluation of ECCS Mixing Phe

23. V. E. Schrock et al., "Steam-Water Critical nomena," NUREG/CR-2786 (General Electric Flow Through Small Pipes from Stratified Up Company, GEAP-22150, EPRI NP-2542), May

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Tien, V. P. Carey, and J. K. Ferrell, Editors; 33. G. P. Gaspari, C. Lombardi, G. Peterlongo, Vol. 5, pp. 2307-2311; Hemisphere Publishing "Pressure Drops in Steam-Water Mixtures.

Corp., 242 Cherry St., Philadelphia, PA Round Tubes Vertical Upflow," Centro Infor

19106, 1986. mazioni Studi Esperienze, Milan, Italy, CISE-R83, 1964. (Available from NTIS.)

24. V. E. Schrock et al., "Small Break Critical Dis charge - Roles of Vapor and Liquid Entrain 34. A. Alessadrini, G. Peterlongo, R. Ravetta, ment in Stratified Two-Phase Region Upstream "Large Scale Experiments on Heat Transfer and of the Break," NUREG/CR-4761 (Lawrence Hydrodynamics with Steam-Water Mixtures.

Berkeley Laboratory, LBL-22024), December Critical Heat Flux and Pressure Drop Measure

1986. ments in Round Vertical Tubes at the Pressure of 51 kg/cm 2 abs," Centro Informazioni Studi

25. W. D. Beckner and J. N. Reyes, Research In Esperienze, Milan, Italy, CISE-R86, 1963.

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1981. (Available in the NRC Public Document 35. E. Janssen and J. A. Kervinen, "Two-Phase Room.) Pressure Drop Across Contractions and Expan sions: Water-Steam Mixtures at 600 to 1400

26. W. D. Beckner, J. N. Reyes, R. Anderson, psia," AEC R&D Report GEAP-4622, 1964.

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36. E. Janssen and J. A. Kervinen, "Two-Phase

27. C. J. Crowley et al., "1/5-Scale Countercurrent Pressure Drop in Straight Pipes and Channels:

Flow Data Presentation and Discussion," Water-Steam Mixtures at 600 to 1400 psia,"

NUREG/CR-2106 (Creare Incorporated, Creare AEC R&D Report GEAP-4616, 1964. (Avail TN-333), November 1981. able in the NRC Public Document Room.)

37. R. T. Lahey, B. S. Shiralkar, D. W. Radcliffe,

28. K. H. Sun, "Flooding Correlations for BWR "Two-Phase Flow and Heat Transfer in Multi Bundle Upper Tieplate and Bottom Side-Entry rod Geometries: Subchannel and Pressure Orifices," in Multi-Phase Transport: Funda Drop Measurements in a Nine-Rod Bundle for mentals, Reactor Safety, Applications, Vol. 1, Diabatic and Adiabatic Conditions," AEC R&D

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38. G. L. Yoder et al., "Dispersed Flow Film Boil

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1.157-18

REGULATORY ANALYSIS

A separate regulatory analysis has not been pre Revision of the ECCS Rule and Supporting- Regula

> pared in support of this regulatory guide. The regula tory Guide," is available in the NRC Public Docu tory analysis that supports the rulemaking effort also ment Room, 2120 L Street NW., Washington, DC,

covers this regulatory guide. "Regulatory Analysis for under Regulatory Guide 1.157 (52 FR 6334).

1.157-19

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