NUREG-0800 (Formerly NUREG-75/087)

U.S. NUCLEAR REGULATORY COMMISSION

STANDARD REVIEW PLAN

OFFICE OF NUCLEAR REACTOR REGULATION

Rev. 2 - January 2006

USNRC STANDARD REVIEW PLAN

Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission’s policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are not substitutes for regulatory guides or the Commission’s regulations and compliance with them is not required. The standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants. Not all sections of the Standard Format have a corresponding review plan.

Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new information and experience.

Comments and suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C. 20555.

6.2.1.3 MASS AND ENERGY RELEASE ANALYSIS FOR POSTULATED

LOSS-OF-COOLANT ACCIDENTS (LOCAs)

REVIEW RESPONSIBILITIES

Primary - Organization responsible for the review of containment systems |

Secondary - None

I. AREAS OF REVIEW

The analyses of the mass and energy release are reviewed to assure that the data used to

evaluate the containment and subcompartment functional design are acceptable for that

purpose. The review includes the following areas: 

1. The energy sources that are available for release to the containment. 

2. The mass and energy release rate calculations for the initial blowdown phase of the

accident. 

3. For pressurized water reactor (PWR) plants, because of the additional steam generator

stored energy available for release, the mass and energy release rate calculations for

the core reflood and post-reflood phases of the accident. 
Rev. 2 - January 2006 6.2.1.3-2

| 4. Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) 

|

| For design certification and combined license reviews, the applicant's proposed

| information on the ITAAC associated with the systems, structures, and components

| (SSCs) related to this SRP section is reviewed in accordance with SRP Section 14.3. It

| is recognized that the review of ITAAC is performed after review of the application

| against acceptance criteria contained in this SRP section. Furthermore, the ITAAC are

| reviewed to assure that all SSCs in this area of review are identified and addressed as

| appropriate in accordance with SRP 14.3.

Review Interfaces

The related SRP section is identified by functional relationship. 

Review of the acceptability of piping design criteria, selected break locations and break sizes

based on the provisions made to limit pipe motion, for breaks postulated to occur within

subcompartments is performed under SRP Section 3.6.2. 

The acceptance criteria and review procedures are contained in the referenced SRP section.

II. ACCEPTANCE CRITERIA

| The acceptance criteria given below apply to the mass and energy release analysis for

| postulated LOCAs. If the mass and energy analysis complies with the relevant requirements of

| General Design Criterion 50 and 10 Code of Federal Regulations (CFR) Part 50, Appendix K,

| paragraph I.A, then the staff considers the analysis acceptable. The relevant requirements are

as follows: 

A. General Design Criterion 50, as it relates to the containment and subcompartments

being designed with sufficient margin, requires that the containment and its associated

systems can accommodate, without exceeding the design leakage rate, and the

containment and subcompartment design can withstand the calculated pressure and

temperature conditions resulting from any LOCA. 

B. 10 CFR Part 50, Appendix K, as it relates to sources of energy during the LOCA,

provides requirements to assure that all the energy sources have been considered. 

In meeting the requirements of General Design Criterion 50 the following specific

criterion or criteria that pertain to the mass and energy analysis are used as included

below: 

1. Sources of Energy

The sources of stored and generated energy that should be considered in

analyses of LOCAs include: reactor power; decay heat; stored energy in the

core; stored energy in the reactor coolant system metal, including the reactor

vessel and reactor vessel internals; metal-water reaction energy; and stored

energy in the secondary system (PWR plants only), including the steam

generator tubing and secondary water. 
Rev. 2 - January 2006 6.2.1.3-3

Calculations of the energy available for release from the above sources should

be done in general accordance with the requirements of 10 CFR Part 50,

Appendix K, paragraph I.A. However, additional conservatism should be

included to maximize the energy release to the containment during the blowdown

and reflood phases of a LOCA. An example of this would be accomplished by |

maximizing the sensible heat stored in the reactor coolant system (RCS) and |

steam generator metal and increasing the RCS and steam generator secondary |

mass to account for uncertainties and thermal expansion. |

The requirements of paragraph I.B in Appendix K to 10 CFR Part 50, concerning

the prediction of fuel clad swelling and rupture should not be considered. This will

maximize the energy available for release from the core. 

2. Break Size and Location

a. The staff’s review of the applicant's choice of break locations and types is

discussed in SRP Section 3.6.2. 

b. Of several breaks postulated on the basis of a., above, the break selected

as the reference case for subcompartment analysis should yield the

highest mass and energy release rates, consistent with the criteria for

establishing the break location and area. 

c. Containment design basis calculations should be performed for a

spectrum of possible pipe break sizes and locations to assure that the

worst case has been identified. 

3. Calculations

In general, calculations of the mass and energy release rates for a LOCA should

be performed in a manner that conservatively establishes the containment

internal design pressure (i.e., maximizes the post-accident containment pressure

and the containment subcompartment response). The criteria given below for

each phase of the accident indicate the conservatism that should exist. 

a. Subcompartment Analysis

The analytical approach used to compute the mass and energy release

profile will be accepted if both the computer program and volume noding

of the piping system are similar to those of an approved emergency core

cooling system (ECCS) analysis. The computer programs that are

currently acceptable include SATAN-V (Reference 1), CRAFT-2

(Reference 2), CE FLASH-4 (Reference 3), and RELAP4 (Reference 4),

when a flow multiplier of 1.0 is used with the applicable choked flow

correlation. An alternate approach, which is also acceptable, is to

assume a constant blowdown profile using the initial conditions with an

acceptable choked flow correlation. 
Rev. 2 - January 2006 6.2.1.3-4

b. Initial Blowdown Phase Containment Design Basis

The initial mass of water in the reactor coolant system should be based

on the reactor coolant system volume calculated for the temperature and

| pressure conditions assuming that the reactor has been operating

| continuously at a power level at least 1.02 times the licensed power level

| (to allow for instrumentation error). An assumed power level lower than

| the level specified (but not less than the licensed power level) may be

| used provided the proposed alternative value has been demonstrated to

| account for uncertainties due to power level instrumentation error.

Mass release rates should be calculated using a model that has been

demonstrated to be conservative by comparison to experimental data. 

Calculations of heat transfer from surfaces exposed to the primary

coolant should be based on nucleate boiling heat transfer. For surfaces

exposed to steam, heat transfer calculations should be based on forced

convection. 

Calculations of heat transfer from the secondary coolant to the steam

generator tubes for PWRs should be based on natural convection heat

transfer for tube surfaces immersed in water and condensing heat

transfer for the tube surfaces exposed to steam. 

c. PWR Core Reflood Phase (Cold Leg Breaks Only)

| Following initial blowdown, which includes the period from the accident

| initiation (when the reactor is in a steady-state full power operation

| condition) to the time that the reactor coolant system broken loop

| pressure equalizes to the containment pressure, the water remaining in

the reactor vessel should be assumed to be saturated. Justification

| should be provided for the refill period, which is the time from the end of

| the blowdown to the time when the emergency core cooling system

| (ECCS) refills the vessel lower plenum. An acceptable approach is to

assume a water level at the bottom of the active core at the end of

blowdown so there is no refill time. 

Calculations of the core flooding rate should be based on the ECCS

| operating condition during the core reflood phase, which begins when the

| water starts to flood the core and continues until the core is completely

| quenched, or the post-reflood phase, which is the period after the core

| has been quenched and energy is released to the RCS primary system

| by the RCS metal, core decay heat, and the steam generators, that

| maximizes the containment pressure. 

Calculations of liquid entrainment, i.e., the carryout rate fraction, which is

the mass ratio of liquid exiting the core to the liquid entering the core,

| should be based on the PWR full length emergency cooling heat transfer

experiments (Reference 5). Liquid entrainment should be assumed to
Rev. 2 - January 2006 6.2.1.3-5

continue until the water level in the core is 61 cm (2 feet) from the top of |

the core. An acceptable approach is to assume a carryout rate fraction

(CRF) of 0.05 to the 46 cm (18-inch) core level, a linearly increasing CRF |

to 0.80 at the 61 cm (24-inch) level, and a constant CRF of 0.80 until the |

water level is 61 cm (2 feet) from the top of the core. Above this level, a |

CRF of 0.05 may be used. 

The assumption of steam quenching should be justified by comparison

with applicable experimental data. Liquid entrainment calculations should

consider the effect on the CRF of the increased core inlet water

temperature caused by steam quenching assumed to occur from mixing

with the ECCS water. 

Steam leaving the steam generators should be assumed to be

superheated to the temperature of the secondary coolant. 

d. PWR Post-Reflood Phase

All remaining stored energy in the primary and secondary systems should

be removed during the post-reflood phase. 

Steam quenching should be justified by comparison with applicable

experimental data. 

The results of post-reflood analytical models should be compared to

applicable experimental data. 

e. PWR Decay Heat Phase

The dissipation of core decay heat should be considered during this

phase of the accident. The fission product decay energy model is

acceptable if it is equal to or more conservative than the decay energy

model given in Branch Technical Position ASB 9-2 in SRP Section 9.2.5. 

Steam from decay heat boiling in the core should be assumed to flow to

the containment by the path which produces the minimum amount of

mixing with ECCS injection water. 

The following methods and computer models are acceptable for calculating the |

mass and energy releases for containment design basis calculations: |

|

Babcock and Wilcox / Framatome ANP: CRAFT (Reference 6), CRAFT-2 |

(Reference 2), RELAP5/MOD2-B&W, Revision 1 (Reference 7), and |

RELAP5/MOD2-B&W, Revision 4 (Reference 8). |

|

Combustion Engineering: CEFLASH-4A (Reference 3), and CESSAR System 80 |

(Reference 9 and 10). |

|
Rev. 2 - January 2006 6.2.1.3-6

| General Electric: M3CPT (Reference 11), NEDO-20533 (Reference 12), and

| SHEX (Reference 13, 14, and 15).

|

| Westinghouse: WCAP-8312 (Reference 16), SATAN-V (Reference 1), WCAP-

| 10325 (Reference 17), SATAN-VI (Reference 18), and WREFLOOD (Reference

| 19).

|

| Theses codes and methods have been referenced in licensee submittals and on

| a case by case basis have been found to be acceptable for these purposes.

|

| Other methods will be acceptable if they are found to be conservative for these

| calculations.

|

| C. 10 CFR 52.47(a)(1)(vi) provides the requirement for ITAAC for design certification 

| reviews

|

| D. 10 CFR 52.97(b)(1) provides the requirement for ITAAC for combined license reviews

|

| Technical Rationale

| The technical rationale for application of the above acceptance criteria to the mass and energy

| release analysis for postulated LOCAs is discussed in the following paragraphs.

|

| 1. GDC 50 requires the containment structure and associated heat removal system to be

| designed with margin to accommodate any LOCA such that the containment design leak

| rate is not exceeded. A LOCA potentially causes the greatest pressure surge and

| release of fission products when compared to any other accident. Since it is the most

| severe challenge expected, containment must be designed to definitively withstand this

| accident. Following GDC 50 will ensure that containment integrity is maintained under

| the most severe accident conditions thus precluding the release of radioactivity to the

| environment. 

| 2. Appendix K to 10 CFR 50 provides required and acceptable features of evaluation

| models used to analyze various circumstances applicable to the ECCS. Section I.A of

| Appendix K provides a comprehensive list of LOCA heat (energy) sources and the

| reactor operating history assumptions associated with those heat sources. Since the

| mass and energy release analysis for postulated LOCAs is used to design containment

| and containment subcompartments such that they will withstand the worst case LOCA, it

| is critical that all potential energy sources are taken into account. Following 10 CFR 50

| Appendix K will ensure that containment and containment subcompartments are

| designed to accommodate all energy sources for the worst case LOCA, thus precluding

| the potential release of radioactivity to the environment following such a LOCA.

III. REVIEW PROCEDURES

The procedures described below are followed for the review of the mass and energy release

analysis for LOCAs. The reviewer selects and emphasizes material from these procedures as

may be appropriate for a particular case. Portions of the review may be carried out on a generic

basis or by applying the results of previous reviews of similar plants. 
Rev. 2 - January 2006 6.2.1.3-7

The reviewer confirms, with the lead reviewer for SRP Section 3.6.2, the validity of the

applicant's analysis of pipe break size, type and locations for subcompartments containing high

energy lines by using elevation and plan drawings of the containment showing the routing of

lines containing high energy fluids. The reviewer determines that an appropriate reference case

for subcompartment analysis has been identified. In the event a pipe break other than a

double-ended pipe rupture is postulated by the applicant, the lead reviewer for SRP Section

3.6.2 will evaluate the applicant’s justification for assuming a limited displacement pipe break. 

The reviewer compares the sources of energy considered in the loss-of-coolant analysis and

the methods and assumptions used to calculate the energy available for release from the

various sources with the acceptance criteria listed in Section II, above. The reviewer

determines the acceptability of the analytical models and the assumptions used to calculate the

rates of mass and energy release during the initial blowdown, core reflood, and post-reflood

phases of a LOCA. The reviewer also compares energy inventories at various times during a

LOCA to ensure that the energy from the various sources has been accounted for and has been

transferred to the containment on an appropriate time scale. 

The reviewer reviews comparisons made by the applicant to experimental data and makes

comparisons to other available experimental data to determine the amount of conservatism in

the mass and energy release models. 

The reviewer may perform confirmatory analyses of the mass and energy profiles. The purpose

of the analysis is to confirm the predictions of the mass and energy release rates appearing in

the safety analysis report, and to confirm that an appropriate break location has been

considered in these analyses. 

For reviews under 10 CFR Part 52, the reviewer should follow the above procedures to verify |

that the design set forth in the safety analysis report, and as applicable, site interface |

requirements, and combined license action items, meet the acceptance criteria. Following this |

review, SRP Section 14.3 should be followed for the review of Tier I information for the design, |

including the site parameters, interface criteria, and ITAAC. |

IV. EVALUATION FINDINGS

The conclusions reached on completion of the review of this SRP section are presented in SRP

Section 6.2.1. 

V. IMPLEMENTATION

The following is intended to provide guidance to applicants and licensees regarding the NRC

staff's plan for using this SRP section. 

This SRP section will be used by the staff when performing safety evaluations of design |

certifications and license applications submitted by applicants pursuant to 10 CFR 50 or 10 CFR |

52. Except in those cases in which the applicant proposes an acceptable alternative method for |

complying with specified portions of the Commission's regulations, the method described herein

will be used by the staff in its evaluation of conformance with Commission regulations. 
Rev. 2 - January 2006 6.2.1.3-8

| The provisions of this SRP section apply to reviews of applications docketed six months or more

| after the date of issuance of this SRP section.

VI. REFERENCES

The references for this SRP section are listed in SRP Section 6.2.1.

| (As this is the first SRP section of 6.2.1 being updated, references, in sequence, are provided

| here for completeness)

|

| 1. WCAP-7750, “SATAN-V, A Computer Space Time Dependent Analysis of Loss of

| Coolant,” August 1971.

|

| 2. BAW-10092, “CRAFT-2 Fortran Program for Digital Simulation of a Multinode Reactor

| Plant During a Loss-of-Coolant Accident,” Babcock and Wilcox Company, December

| 1974.

|

| 3. CENPD-133 Supplement 5-A, “CEFLASH-4A — A Fortran77 Digital Computer Program

| for Reactor Blowdown Analysis,” dated June 1985.

|

| 4. W. H. Retting, G. A. Jayne, K. Y. Moore, C. E. Slater, and M. L. Uptmor, “RELAP3 - A

| Computer Program for Reactor Slowdown Analysts,” IN-1321, Idaho Nuclear

| Corporation, June 1970.

|

| 5. F. C. Cadek, et al., “PWR FLECHT (Full Length Emergency Cooling Heat Transfer),

| Final Report,” UCAP-7665, Westinghouse Electric Corporation, April 1971.

|

| 6. BAW-10030, “CRAFT - Description of Model for Equilibrium LOCA Analysis Program,”

| Babcock and Wilcox, Lynchburg, VA, October 1971. 

|

| 7. B&W-10164P, Revision 1, “RELAP5/MOD2-B&W - An Advanced Computer Program for

| Light Water Reactor LOCA and Non-LOCA Transient Analysis,” Babcock and Wilcox,

| October 1988.

|

| 8. BAW-10164P-A, Revision 4, “RELAP5-MOD2-B&W - An Advanced Computer Program

| for Light Water Reactor LOCA and Non-LOCA Transient Analysis,” November 2002.

|

| 9. Combustion Engineering letter DP-456, F. M. Stern to E. Case, dated August 19, 1974,

| Chapter 6, Appendix 6B to CESSAR System 80 PSAR.

|

| 10. “NRC Safety Evaluation Report - Standard Reference System, CESSAR System 80,”

| Combustion Engineering Inc., December 1975.

|

| 11. NEDO-10320, “The General Electric Pressure Suppression Containment Analytical

| Model,” General Electric Company, April 1971; Supplement 1, May 1971; Supplement 2,

| January 1973.

|

| 12. NEDO-20533, “The General Electric Mark III Pressure Suppression Containment

| System Analytical Model,” June 1974.
Rev. 2 - January 2006 6.2.1.3-9

|

13. “SHEX Model Description,” PDR ADOCK 050000321, 9807130260, 980706. |

(Proprietary, not available for public release) |

|

14. NEDE-30911, “SHEX-04 User's Manual,” Class II, General Electric Company, August |

1985. And “SHEX-04V User's Manual (Addendum to SHEX-04 User's Manual),” |

NEDE-30911-1, June 1994. |

|

15. “Use of SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for |

Containment Long-Term Pressure and Temperature Analysis,” NRC Letter from A. |

Thadani (NRC) to G. L. Sozzi (GE), July 13, 1993. |

|

16. WCAP-8264-P-A, “Westinghouse Mass and Energy Release Data for Containment |

Design,” Westinghouse Electric Corporation, March 1974 (Proprietary, not available for |

public release). WCAP-8312-A (Non-Proprietary). |

|

17. WCAP-10325-P-A “Westinghouse LOCA Mass and Energy Release Model For |

Containment Design - March 1979 Version,” May 5, 1983 (Proprietary, not available for |

public release), WCAP-10326-A, May 5, 1983 (Non-Proprietary). |

|

18. WCAP-8302, “SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of |

Loss-of-Coolant,” June 1974 (Proprietary, not available for public release). WCAP-8306 |

(Non-Proprietary). |

|

19. WCAP-8170, “Calculational Model for Core Reflooding after a Loss-of-Coolant Accident |

(WREFLOOD Code),” June 1974 (Proprietary, not available for public release). WCAP- |

8171 (Non-Proprietary). |
SRP Draft Section 6.2.1.3

Attachment- Description of Changes

Rev. 2 - January 2006 6.2.1.3-10

The following summarizes the changes from Revision 1, dated July 1981. Minor editorial

changes and formatting changes are not identified by side bars. 

1. Review Responsibilities: 

a. Editorial change made to reflect the current primary review function –

organization maintained separate from the SRP section.

2. Areas of Review:

a. Added "Review Interfaces" heading to Areas of Review. Reformatted existing

description to describe the relationship to SRP Section 3.6.2.

b. Added paragraph on ITAAC, SRP Section 14.3

3. Acceptance Criteria:

a. Section A, changed "subcompartment" was changed to "subcompartments" to

reflect the fact that a containment may have more than one subcompartment.

b. Section B.1, provided an example of additional conservatism.

c. Section B.3.b, modified to be consistent with measurement recapture under 10

CRF 50 Appendix K . Federal Register: June 1, 2000 (Volume 65, Number 106)

[Page 34913-34921].

d. Section B.3.c, added brief definitions for blowdown, refill, reflood and post-reflood

phases of LOCA and clarified that the treatment is to address the core flooding

rate for designs which are peak pressure limited during the reflood phase or

during the post-reflood phase. 

e. Section B.3, provided an updated list of acceptable models.

f. Added C and D to refer to ITAAC for design certification and combined license

reviews.

g. Technical rationale associated with GDC 50 and Appendix K to 10 CFR Part 50

was formulated and added to this section in accordance with the specified SPRUDP format. 

4. Review Procedures:

a. Added paragraph on implementing of 10 CFR 52. A standard paragraph was

added to address Review Procedures in design certification reviews.
Rev. 2 - January 2006 6.2.1.3-11

5. Implementation:

a. Editorial changes capturing applicability to Part 52 and time-frame in which SRP 

update goes into effect.

6. References:

a. Which were provided in Section 6.2.1 are listed in this section for completeness.

b. Added new computer code references as described above.