ML19221B128
| ML19221B128 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-06.2.1., NUREG-75-87, NUREG-75-87-6.2.1., SRP-06.02.01.03, SRP-6.02.01.03, NUDOCS 7907120467 | |
| Download: ML19221B128 (4) | |
Text
NU REG-75/087
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'S U.S. NUCLEAR REGULATORY COMMISSION
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o Mli'"y#"] STANDARD REV EW PLAN a,
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 6.?. l.3 MASS AND ENERGY RELEASE ANALYSIS FOR POSTULATED LOSS-OF-COOLANT ACCIDENTS REVIEW RESPONSIBILITIES Primary - Contairrent Systems Branch (CSB)
Secondary - Core Perforr:ance Branch (CPB)
I.
AREAS OF REVIEW The CSB reviews analyses of the mass and energy released to the containment during loss-of-coolant accidents (LOCA) in conjunction with the review of the functional capability of the containment structure. While the CPB has the primary responsibility for the review of mass and energy release analyses, the CSB reviews this area as it relates to containr.ent func-tional design. Se review includts the following areas:
1.
The source' sf the energy a',;ured to be released to the containment.
2.
Tne applicant's 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 stean generator stored energy available for release, the r: ass and energy release rate calculations for the core reflood and post-reflood phases of the accident.
II.
ACCEPTAN E _ CRITERIA The following acceptance criteria apply to the mass and energy release analysis for postula-ted loss-of-coolant accidents:
1.
Sources of Erery The socrces of stored and generated energy that should be considered in analyses of loss-of-enolant accidents include: reactor power; decay heat; stored energy in the core; stored energy in the reactor coolant system retal, including the reactor vessel and reactor vessel internals; metal-vater reaction energy; and stored energy in the secondary system (PWR plants only), including the steam generator tubing and secondary water.
Calculations of the energy available for release from the above sources should be done in general accordance with the requirements of 10 CFR Part SO, Appendix K, paragraph I.A.
(Ref. 2). However, additional conservatism should be included to maximize the energy release to the containment during the blowdown and reflood phases of a LOCA.
USNRC STANDARD REVIEW PLAN Standard rewtew piene ore peepared for the guicence of the Othee of Nucteer Reactor Roguestson ettff responerbio for the rev4ew of opphcetsono to construct end opeeste nuclear power plente These dscumente ere mede eweiseb6e to the public as port of the Commisedon e po 6cy to Mform the nucleer industry and the s
eenerei pubhc of regwiesory procedures and pobcees Brendard review piene are not e=betutee for esgulatory guedes or the Commission e regulatione and J
comphence wth them 6e n3e requered The svendeed review plan sect 6ene are heyed to Rev.aion 2 of the Standard Formet and Content of Sefety Anerysee Reporte for Nucteer Power Plante Not Ni sectione of the Standard Formet have e correspondmg roweew plan Pubhehed o'endeed review piene wel be revised periodically, se appropriate to accommodate comments and to reflect new mformation and wapenence Commente and ouggestions for emprovement will be conoedered end should be sont to the U S Nucteer Reguietory Commession. Office of Nucleet Reector Reguletion meh6agton. D C 20M6 7907120461
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.
Calculations In general, calculations of the mass and energy release rates for a loss-of-coolant accident should be done in a manner that conservatively establishes the containment internal design pressure; i.e., maximizes the post-accident containment pressure. The criteria given below for each phase of the accident indicate the conservatism that should exist. These calculations should be done for a spectv1 of possible pipe breaks to assure that the worst case has been iden+' Tied.
This spectrum shc_'d include hot leg, cold leg (pump suction), and cold leg (pump discharge) pipe breaks with cross-sectional areas up to and including a double-ended pipe treak, and longi-tudinal splits in the largest pipes with break areas equal to the cross-sectional area of the pipe, a.
Initial Blowdown Phase The initial mass of water in the re3ctor coolant system should be based on the reactor coolant system volume calculated for the temperature and pressure con-ditions existing at 102; of full power (Ref. 2).
Mass release rates should be calculated using the Moody model (Ref. 22), or a model that is demonstrated to a equally conservative.
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
.f heat transfer from the secondary coolant to the steam generator tubes for PWR's should be based on natural convection heat transfer for tube surfaces irrersed in water and condensing heat transfer for the tube surfaces exposed to steam.
For Mark I EWR plants, the nass and energy releases from the rupture of a re-circulation line should be based on the General Electric blowdown nodel(Ref. 30).
For Mark II and III plants, nass and energy releases from the rupture of a re-circulation line or main steam line should be based on a blowdown codel which accounts for the short-term (less than one second) pressure respense time of the drywell.
The f ollowing computer codes and rnodels are acceptable for calculating mass and energy release rates during the blowdown phase, if they incorporate the foregoing criteria: CRAFT (Ref. 23), SATAN VI (Pef. 24), CE FLASH-4 (Ref. 25), and the GE blowdown model (Ref. 30). Other codes will be acce: table for containment analyses if they are approved by CPB and are determined to be conservative for containment analyses.
6.2.1.3-2
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b. PWR Core Reflood Phase (Cold Leg Breaks Only) Following initial blowdown of the reactor coolant system, the water remaining in the reactor vessel should be assumed to be saturated and at the level of the bot-tcm of 'h-3ctive core. Calculations of the core flooding rate should be based on the emergency cose cooling system operating condition that maximizes the containment pressure either during the core reflood phase or the post-reflood phase. Calculations of liquid entrainment; i.f, 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 FLEr4T experiments (Ref. 26). Liquid entrainment should be assumed to continue until the water level in the core is ' eet from the top of the core. An acceptable approach is to assume a carryout rate fraction (CRF) of 0.05 to the 18-inch core level, a linearly increasing CRF to 0.80 at the 24-inch level, and a constant CRF of 0.80 until the water level is 2 feet from the top of the core. Above this level, a CRF of 0.05 nay be used. The assumption of steam quenching should be justified by comparison with appli-cable experimental data. Liquid entrainment calculations should consider the effect on the carryout rate fraction of the increased core inlet water temperature caused by steam quenching. Steam leaving the steam generators sho~ld be assumed to be superheated to the temperature of the secondary coolant. c. PWR Post-Reflood Phase (Cold leg Breaks Or.ly) 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 experinental data, d. PWR Decay Heat Phase (Cold leg Breaks Only) The dissipation of cnre decay heat should be considered during this phase of the accident. III. REVIEW PROCEDURES The procedures described below are followed for the review of the mass and erergy release analysis for loss-of-coolant accidents. 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. 6.2.1.3-3 A _i &-: - 9
The CPB and the CSB compare the sources of energy considered in the loss-of-coolant analysis and the nethods and assumptions used to calculate the energy available for release from the various sources with the accepta ice criteria listed in Section II, above. The CPB deter-nines the acceptability of the analytical models and the assumptions used to calculate the rates of mass and energy relea se during the initial blowdown, core reflood, and post-reflood phases of a loss-of-coolant accident. The CSB also compares energy inventories at various times during a loss-of-coolant accident 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 acceptance of the methods and deterTnination of the degree of conservatism is the resronsibility of the CPB. In general, such efforts are accomplished through co-operation between branches. Mass and energy release data for computer codes that have not been previously reviewed and accepted can be compared to the results of computer codes which have been found acceptable, such as the S ATA-VI, CRAFT, and CE FLASH-4 computer codes, and the GE blowdown model. FLOOD 1 or FLOOD 2 computer codes (Refs.17 and 16) are used, as appropriate, to analyze the mass and er.ergy releases for the PWR core reflood phase. The acceptability of 7 ssumptions made in the analyses regarding steam quenching may be determined by corrparing the results of the analyses with applicable experimental data. IV. EVALUATION FINDINGS The conclusions reached on completion of the review of this section are presented in Standard Review Plan 6.2.1. O V. REFERENCES. The references for this plan are listed in Standard Review Plan 6.2 O }kf 2 6.2.1.3-4}}