ML19221B120
| ML19221B120 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-06.2.11, NUREG-75-87, NUREG-75-87-6.2.11, SRP-06.02.01.01, SRP-6.02.01.01, NUDOCS 7907120454 | |
| Download: ML19221B120 (5) | |
Text
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4 U.S. NUCLEAR REGULATORY COMMISSION M; im.
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STANDARD REV EW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION
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SECTICN 6.2.1.1.A PwR DRY CGNTAINMENTS, INCLUDING SUBATMOSPHERIC CONTAINPENTS REVIEW RfSPONSIBILITIES Primary - Containment Systems Branch (CSB)
Secondary - Instrumentation and Control Systems Branch (ICSB) 1.
AREAS OF REVIEW f or pressurizod water reactor (par) plants with dry containments, the CSB review covers the following areas-1.
The temperature and pressute cunditions in the containment due to a spectrum (includ-ir 7 break size and location) of postulated loss of conlant accidents (i.e.,
reactor coclant system pipe breaks) and secondary system steam and feedwater line breaks.
2.
The maximum expected external pressure to which the containment may be subjected.
3.
The minimum containment pressure used in analyses of emergency core cooling system capability.
4.
The effectiveness of static and active heat removal mechanisms 5.
The pressure conditions within st.bcompartments and acting on system components and supports due to high energy line breaks.
6.
The instrumentation provided to monitor and record containment conditions during and following an accident.
7.
The prcposed technical specifications at the operating license stage of review pertaining to the surveillance requirements for spring or weight loaded check valves used in subatmospheric containments, and vacuum relief devices.
At the request of the CSB, the ICSP wi'l evaluate; (1) the ef fectiveness of the administra-tive controls and the instrumentation and control provisions to prevent inadvertant peration of the containment heat removal systems or system trains; and (2) the functional capability of the post-accident monitoring instrumentation and recording equipment.
USNRC STANDARD REVIEW PLAN
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II.
ACCEPTANCE CRITERI A The following acceptance criteria complement General Design Criterion 50 and apply to thr design a7d f unctional capability of PWR dry containments:
1.
For plants at the construction permit (CP) stage of review, the containment design pressure should provide at least a 10% margin above the accepted peak calculated cnntainment pressure following a loss of coolant accident, or a steam or feedwater 1ine break.
2.
For plants at the operating license (OL) stage of review, the peak caiculated containment pressure following a loss of-coolant accident, or a steam or feedwater line break, should be less than the containment design pressure.
In general, the peak calculated containment pressure should be approximately the same as at the construction permit stage of review.
However, revised or upgraded analytical models or minor changes in the as-built design of the plant may result in a decrease in the margin.
3.
Ine containment pressure should be reduced to less than 50't of the peak calculated pressure for the design basis loss-of coolant accident within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the postulated accident. This is in conformance with the requirement of General Design Criteria 38 " Containment Heat Removal Systems," that the containment pressure and temperature following a loss of-coolant accident be rapidly reduced. If analysis shows that the calculated containment pressure may not be reduced to 57t of the peak calculated pressure within 24 huars the AAB should be notified.
4.
for subatmospherit containments, the containment pressure should be reduced to below atmospheric pressure within one hnur after the postulated accident, and the sututmospheric condition maintained f c r at least 30 days f
5.
The loss-of-coolant accident a'lalysis should be based on the assumption of loss of offsite p w r and the most severe single active failure in the emergency power system (e.g., a diesel generator failure), the containment heat removal systems (e.g., a fan, pump, or valve failure), or the core cooling systems (e.g., a pump ur valve failure).
The selection made should result in the h ghest calculated contain-ment pressure.
6.
The containra r* response analysis for postulated seconf3ry system pipe ruptures should be based on the most severe single active failure in the containment heat removal systems (e.g., a fan, pump, or valve failure) or the secondary system isolation provisions (e.g., main steam isolation valve failure or feedwater line isolatio1 valve failure). The analysis should also be based on a spectrum of pipe Lreak sizes and reactor power levels. The accident conditions selected should result in the highest calculated containment pressure or temperature depending on the purpose of the analysis.
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7.
The minimum calculated coltainment pressure should not be less than that used in the analysis of the emergency core cooling system capability (See SRP Section 6.2.1.5, " Minimum Contai...,ent Pressure Analysis for Emergency Core Cooling System Performance Capability Studies").
8.
Provi s ions shoult' be made to protect the containment structure against possible damage from externt ' pressure conditions that may result, for example, from in-advertent operation f containment heat removal systems. The provisions made shculd include conser.ative structural design to assure that the containment struc-or inter-ture is capable of wil lstandii.g the maximum expected external pressure; locks in the plant protection system and administrative controls to preclude in-advertent operation of the systems; or for steel conta nment vessels, vacuum relief i
devices proviaed in accordance with the requirements of the ASME Boiler and Pres-sure Vessel Code,Section III, Division 1, Subsection NE (Ref. 3), and applicable requireme:.ts of General Design Criteria 54 and 56.
J.
If the primary containment is designeu to withstand the m nimum expected external pressure, the external design pressure of the containment should provide an adequate margin above the maximum expected external pressure to account for uncertainties in the analysis of the postulated event.
10.
Ccitainment internal structures and system components (e g.,
reactor vessel, pres-surizer, steam generators) and supports should be designed te withstand the differen-tial pressure loadings that may be imposed as a result of pipe breaks within the containment subcompartments (See SRP Section 6.2.1.2, "Subcompartment Analysis").
11.
Instrumentation capable of operating in the post accident environment should be provided to monitor the containment atmosphere pressure and temperature and the sump water temperature following an accident. The instrumentation should have adequate range, accuracy, and response to assure that the above parameters can be tracked and recorded throuq1out the course of an accident. Reguitory Guide 1.97, "In3trumentation For Light Water Cooled Nuclear Power Plants to Assess Plant Condi-tions Durine and following An Accident," should be followed.
For those areas of review identified in subsection I of this SRP section as being the responsibility of other branches. The acceptance criteria and their methods of applica-tion are contained in the SRP sections corresponding to those brancht III. REVIEW PRDCEDURES The following procedures are for the review of PWR dry containments. The reviewer l
selects and emphasizes material from these procedures as may be appropriate for a par-ticular case. Portions of the review may be carried out ]n a generic basis for aspects of functional design common to a class of dry containments or by adopting the results of previous reviews of plants with essentially the same containment functional design.
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6.2.1.1.A-3 Rev. I
Upon request from the primary reviewer, the secondary review branches will provide input for the areas of review stated in subsection I of this plan. The primary reviewnr obtains and uses such input as required to assure that this review procedure is complete.
The C58 reviews the containment response analyses to determine the acceptability of the calculated containment design pressure and temperature, and in addition, the containment depressurization tine.
The AAB must be notified if the containment depressurization time does not meet the acceptance criterion). The CSB reviews the assumptions made in the analyses to maximize the calculated containment pressure a1d t.mperature.
The CSB determiaes the conservatism of the respective containment response analyses by comparing the analytical models, and the assumptions made, with the acceptance criteria in subsection II, and by performing appropriate confirmatory analyses. It is not necessary to perform accident pressure calculations for every plant.
The CSB will ascertain, however, that the adequacy of the applicant's calculatior.al nedel has been demonstrated.
The CSB determines that the applicant has identified the pipe break (s) resulting in the highest containment pressure and temperature. Hot leg, cold leg (rump suction), and cold leg (pump discharge) pipe breaks of the reactor coolant system, and secondary system steam and feedwater line breaks, should be m lyzed by the applicant. The CSB reviews the assumptions used to determine that the maiyses are acceptably conservative.
The CSB performs confirmatory containment response analyses when necessary using the CONTEMPT-LT computer code (See References 7, 8, and 9 f or a description of this code).
l The purpose of these analyses is to confirm the applicant's predictions of the response of the containment to loss of-coolant accidents and main steam and feedwater line breaks.
In general, only the limiting pipe breaks, i. e., the pipe breaks which establish the containment design pressure and containment depressurization time, are analyzed. However, if in the judgment of the C5B the worst break has not been identified, other pipe breaks will be ana'yzed.
The CSB reviews analyses of the external pressure of the containment structure caused by pressure and temperature changes inside the containment due to inadvertent operation of containment heat removal systems. The CSB determines whether the most severe condition has been identified, and whether the analysis was done in a conservative manner.
The CSB evaluates the acceptability of the provisions made in the plant design to mitigate or withstand the consequences of the above postulated events, and evaluates in conjunc-tion with the ICSB, the administrative controls and instrumentation and control provi-sions to preclude these events.
The CSB determines whether instrumentation capable of withstanding the post accident environment, and recording equipment, and has been provided to monitor and record the course of an accident within the containment. The CSB also determines whether the instrumentation and recording equipment can accomplish the objectives stated in Regula-tory Guide 1.97.
This review is coordinated with the'ICSB.
The ICSB, under SRP section 7.3, has review responsibility for the acceptability of, and the qualification tost O
Rev. 1 6.2.1.1.A-4
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program for the sensing and actuation instrumentation of the plant protection system and the post accident monitoring instrumentation and recording equipment.
IV.
EVALUATION FINDINGS The conclusions reached on completion of the review of this section are presented in SRP Section 6.2.T.
V.
REFERENCES The references for this SRP section are listed in SRP Section 6.2.1.
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