ML20198G592
| ML20198G592 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 05/05/1975 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1064 NUDOCS 8605290689 | |
| Download: ML20198G592 (8) | |
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C V. A. Moore, Assistant Director for Light Water Reactors, Group 2 EL WFPSS 1 & 4 REVISION 2 TO DRAFT SAFETT EVALUATION REPORT Plant Name: WPPSS 1 & 4 i
Desket Nos.: 50-460 and 50-513 l
Licensing Stage: CP l
NSSS Supp11 erg Babcock & Wilcox i
Architect Engineer: United Engineers and Constructors l
Containment Type Dry l
Responsible Branch & Project Manager: LWR 2-3; T. Com Requested Completion Date: May 9, 1975 Applicant's Response Dates N/S Review Status: Complete Emelosed is Revision 2 to the Draf t Safety Evaluation Report prepared by the Contairuneet Systems Branch for the Washington Public Power Supply System, Units 1 & 4.
This revisica addresses the unresolved items identified in' Revision 1 of the Draft Safety Rvaluation Report, dated March 4, 1975. The applicant has provided additional information in subsequent amendments to the PreH=inary Safety Analysis Report which satisfactorily resolves the outstanding items.
i The CSB review of the WPPSS, thits 1 & 4 is now complete.
Original Signed by Robert L. Tedesco Robert L. Tedesco, Assistant Director for Containment Safety Division of Technical Review Enclosures i
As stated cc:
S. Hanauer W. Mcdonald l
F. Schroeder D. Risenhut NRR Reading File j
T. Cox CS Reading File C. Lainas CSB Reading File A. Schwencer Docket File J. Shapaker J. Kudrick y
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o REVISION 2 TO THE DRAFT SAFETY EVALUATION (CONTAIhifENT SYSTEMS)
WASHINGTON PUBLIC POWER SUPPLY SYSTEMS, UNITS 1 & 4 DOCKET NOS. 50-460 & 50-513 6.2 Containment Systems (No change) 6.2.1 Containment Functional Design The containment will consist of a steel-lined, reinforced concrete structure with a net free volume of 3,090,000 cubic feet. The con-tainment structure will house the nuclear stea'm supply system, in-cluding the reactor, steam generators, reactor coolant pumps, and pressurizer, as well as certain components of the plant engineered safety features systems. The containment is designed for an internal pressure of 52.0 psig and a temper.ature of 283.5'F.
The applicant has analyzed the containment pressure response for the postulated loss-of-coolant accident. Calculated mass and energy release rates from postulated pipe breaks to the containment were used as input data for the CONTRAST-S computer program, which performs transient thermodynamic calculations including the effects of contain-ment heat removal systems and structural heat sinks to calculate the containment pressure.
The applicant has described the methods used to determine the contain-ment design pressure in the Preliminary Safety Analysis Report (PSAR).
A spectrum of loss-of-coolant accident break locations and sizes in g
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primary system piping and steam line breaks were considered. The consequences of a postulated double.-ended pipe rupture at the pump suction of the reactor coolant system result in the highest containment I
pressure.
The single active failure applied to this design basis accident is the loss of one of the two emergency diesel generators in conjunction with the loss of offsite power. Minimum containment cooling, assumed in the analysis, included one containment spray train. Minimum safety injection of the ECCS is correspondingly assumed. Thus the containment heat removal systems available are treated on a conservative basis.
The mass and energy released to the containment is considered in terms of the blo'wdown and post-blowdown phases of a loss-of-coolant accident.
The blowdown phase (about 24 seconds) of the accident is that time interval immediately following the occurrence of the postulated accident during which most of the energy contained in the reactor coolant system, including the primary coolant, metal and core stored energy is released to the containment.
The post-blowdown phase (af ter about 24 seconds) consists of the refill, reflood and post-reflood periods. The refill period (no refill time assumed) is that time during which the lower reactor vessel plenum is refilled to the bottom of the core by the ECCS. The reflood period (until about 140 seconds) is that time during which the core is reflooded to its 10-foot elevation and when most of the stored energy is removed from the steam generator located in the broken reactor pipe loop.
The stored energy in the steam generator in the intact reactor coolant loop is primarily removed during the post-reflood O
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period (af ter about 140 seconds) of the accident.
The CRAFT computer program was used by the applicant to determine the mass and energy addition rates to the containment during the blowdown ph'ase and for part of the post-blowdown phase of the accident. To obtain a conservatively high energy release rate to the containment during the blowdown phase, the applicant assumed that the core would remain in nucleate boiling until the quality of the coolant was 1.0, so that the energy release rate from the core would be maximized.
Under this assumption, the core transfers more heat to the containment than would be predicted by,a calculation suitable for core heatup and an emergency core cooling performance evaluation. This additional energy release from the core increases the calculated containment pressure and therefore assures a margin of conservatism in the analysis. The CRAFT computer code _has been accepted by the NRC for calculating energy released during a loss-of-coolant accident.
The CRAFT program was also used by the applicant to predict mass and energy releases to,the containment during the co're reflood phase of the accident. The reflood phase is important when analyzing postulated pipe ruptures in the reactor coolant system cold legs since the steam and entrained liquid carried out of,the core for these break locations can pass through the steam generators which constitute an additional energy source. The steau and entrained water leaving the core and passing through the steam generators will be evaporated and/or superheated to the temperature of the steam generator secondary fluid. During core e-
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i reflood, the carry-out-race fraction determines the amount of stea'm and entrained water leaving the core and, therefore, the amount of energy that can be transferred from the steam generator is calculated based on a correlation inherent in CRAFT. The CRAFT reflood calculation for the design basis accident included average carry-out-rate fractions in excess of 0.8.
Results of the FLECHT experiments indicate that the carry out fraction of fluid leaving the core during reflood is 80%
of the incoming flow to the core which confirms the assumptions of CRAFT. The rate of energy released to the containment during this phase is proportional to the flow rate into the core, and thus through the steam generator.
We have compared the mass and energy release to the containment during the reflood phase of the accident, as calculated with our FLOOD computer code, with those values predicted by the applicant. The #
results of this comparison indicate equivalent predictions of energy release. Therefore we have accepted th,e applicant's computer model as a conservative method of computing core reflood for this plant.
After the core is completely covered with water, decay heat generation will produce boiling in the core and a two-phase mixture of steam and water will exist in the core. This mixture can enter the steam generator and provide an additional energy release to the containment.
The applicant's analytical model accounts for this additional frothing energy.
At 1000 seconds af ter the accident essentially all of the availabic sensible heat has been removed from the primary system and the steam m
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.. generators. At this time the applicant uses the PRIT computer code to calculate the long-term decay energy release to the containment.
The decay energy is conservatively determined by using 1.2 times the ANS-5 decay heat curve based on infinite reactor operation.
The applicant has calculated the containment pressure from the mass and energy data discussed above. The peak containment pressure of 42.3 psig was calculated assuming minimum operation of the safety injection and containment sprey systems; i.e., assuming a single-active failure I
f of one emergency diesel generator.
f In our evaluation we analyzed the containment pressure response for the postulated double-ended, cold leg, pump suction break using 1) the CONTEMPT-LT computer co?.e (References 1 and 2), 2) the mass and energy release to the containment provided by the applicant, including the additional energy f rom the steam generator during the reflood and post-reflood phases of*the accident, as described above, 3) the con-tainment structural heat sink and heat removal systems, and 4) con-servative condensing heat transfer coefficients to the structures inside containment.
Our calculations agree with the applicant's results.
The applicant has also analyzed the containment pressure response resulting from a postulated failure of a main steam line within the containment, including consideration of a possible single active i
failure in the feedwater isolation system. The applicant calculated a peak containment pressure of 23.2 psig, which is well below the design pressure of the containment. We have reviewed the applicant's analysis and conclude that it is acceptable for this plant.
The applicant has analyzed the pressure response resulting from postulated pipe breaks within the containment interior compartments, such as the reactor vessel cavity, the primary shield pipe penetration, the steam generator compartments, and the pressurizer compartment.
e The applicant used the COMPRESS computer program to calculate the peak compartment pressure differentials. The applicant has set compartment design pressure differentials using at least a 40% margin between the maximum differential pressure calculated and design values used in the structural loading equations. We agree with this margin.
The applicant has calculated pressure differential of 259 psi for a single-ended hot leg rupture in the reactor cavity,1637 psi for a double-ended cold leg rupture in the primary shield pipe penetration, 20.6 p,si for a double-ended hot leg in a steam generator compartment, and 8.3 psi for a double-ended surge line rupture in the pressurizer 1
compartment. We have performed confirmatory analyses using our computer program and predicted pressures lower than the applicant's results.
We have evaluated the containment system functional design in accordance with the General Design Criteria stated in 10 CFR Part 50 of the Commission's Regulations and, in particular, Criteria 16 and 50.
Uo conclude that the applicant's containment design pressure of 52 psig provides adequate margin (about 23%) when compared to the maximum calculated containment pressure of 42.3 psig and is therefore acceptable.
In addition, based ou our confirmatory calculations and the 40% margin specified for the subcompartment design pressure differentials we find the subcompartment design pressure differentials acceptable.
Ce there-fore conclude the containment functional design meets the requirements O
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