ML20059F906

Ack Receipt of Re Rept Technical Aspects of Alwr Emergency Planning

ML20059F906
Person / Time
Issue date:	01/11/1994
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Mcdonald R AFFILIATION NOT ASSIGNED
References
PROJECT-669A NUDOCS 9401140218
Download: ML20059F906 (3)

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January 11, 1994 Project No. 669 Mr. R. P. Mcdonald, RP - ARC Bin 854 Southern Company Services Room 518 P. O. 2625 Birmingham, Alabama 35202

SUBJECT:

THE ELECTRIC POWER RESEARCH INSTITUTE'S (EPRI's) REPORT ON EMERGENCY PLANNING SUBMITTED BY LETTER DATED DECEMBER 31, 1993 This is to acknowledge receipt of your letter dated December 31, 1993, which submitted the EPRI report, " Technical Aspects of Advanced Light Water Reactor Emergency Planning." Your letter stated that this report was submitted to complement the criteria and methodology contained in your May 3, 1993, submittal on emergency planning by providing additional technical basis information and supporting detail. The staff has reviewed your May 3, 1993, submittal, which was incorporated in Revision 5 to the passive Requirements Document. The staff's review is documented in Section 2.6 of Chapter 1 of the staff's final safety evaluation report on the EPRI passive Requirements Document.

We understand that the Nuclear Management and Resources Council will have the lead for coordinating any future industry action on your proposed policy issue concerning simplification of offsite emergency planning.

If you have any questions about the staff's review, contact the project manager, J. H. Wilson, at (301) 504-1108.

Sincerely, (Original signed by)

Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation cc: See next page DISTRIBUTIO :

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Mr. R. P. Mcdonald Project No. 669 EPRI cc: Mr. John Trotter Nuclear Power Division Electric Power Research Institute Post Office Box 10412 Palo Alto, California 94303 Mr. Brian A. McIntyre, Manager Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, Pennsylvania 15230 Mr. Joseph Quirk GE Nuclear Energy Mail Code 782 General Electric Company 175 Curtner Avenue San Jose, California 95125 Hr. Stan Ritterbusch Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095

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ORIGINAL DUE DT: 01/21/94 TICKET NO: 0940002 DOC DT: 12/31/93 NRR RCVD-DATE: 01/04/94 R. PATRICK MCDONALD EPRI TO:

MURLEY FOR SIGNATURE OF :

• • YEL CRUTCHFIELD

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ROUTING:

TECHNICAL ASPECTS OF ADVANCED LIGHT WATER REACTOR MURLEY EMERGENCY PLANNING MIRAGLIA CRUTCHFIELD CALLAN MAILROOM 12G18 ASSIGNED TO:

CONTACT:

DAR CRUTCHFIELD SPECIAL INSTRUCTIONS OR REMARKS:

• • COORDINATE WITH ADT**

Please review'the due date immediately:

If the due date does not allow adequate time, please contact Celeste Smyre on 504-1229, or E-mail (CDS) to request / revise a due date.

The NRR mailroom is located in room 12G18.

Please do not hand carry concurrence packages to the Directors office without first going through the NRR mailroom.

(ckf) p YY WWCo O k EPRI Electnc Power Researon institute Leadershp in Science and Technology December 31, 1993 Dr. Thomas hiurley Director, Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 Dear Dr. hiurley; Enclosed for NRC review is the report, " Technical Aspects of Advanced Light Water Reactor Emergency Planning", dated December 1993.

On many occasions, we have communicated to the NRC our very strong interest in establishing for the ALWR a sound technical basis for updating emergency planning requirements and our conviction that the ALWR designs being developed are consistent with that objective.

On hiay 3,1993 we submitted to the NRC the ALWR emergency planning design criteria and a revision to Volume III of the Utility associated methodology, as Requirements Document (URD).

The enclosed report complements those criteria and methodology with technical basis information and supporting detail.

As a companion effort to the development of this technical basis, we have also been considering ths policy aspects of updated emergency planning. Our ideas on that topic have been assembled in preliminary form and will be provided to NUh1 ARC shortly.

From that point, NUhiARC will have the lead for coordinating industry review, deliberation and subsequent action, as deemed appropriate, for the policy aspects.

We request NRC's review of the technical report as a way to secure the necessary confidence among plant design teams and prospective ALWR owners and licensees that properly designed plants for the next generation will, in fact, have the opportunity to utilize an updated, technically sound emergency planning approach.

As soon as your staff conducts its initial familiarizatic,n review of the report, we suggest a meeting to discuss scope, timing and output product from your detailed review.

As always, we appreciate NRC's active involvement in the industry's ALWR initiatives. Please call if you have any question about this submittal.

Verv Trulv Yours O A 4 - [h, Qf.

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• Telex: 82977 EPRI UF e Fax: (415) 85s-2954 Washington oMee 2000 L Street. NW swte 805, Wasnington. DC 2c336. USA e (202) 872-9222 e Far (202) 2964040

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James Wilson, NRC/ Project 669 John Taylor, EPRI Joe Santucci, EPRI Alan Nelson, NUMARC John Trotter, EPRI David Leaver, Polestar 6

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l TECHNICAL ASPECTS OF ADVANCED LIGHT WATER REACTOR EMERGENCY PLANNING Prepared by the Electric Power Research Institute on behalf of the Advanced Light Water Reactor Program December,1993 k

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TABLE OF CONTENTS Executive Summary and Conclusions iii 1.0 Introduction and Background 1-1

1.1 Purpose and Scope

1-1 1.2 Applicability 1-1 1.3 Technical Reasons to Update the Technical Basis for ALWR l-2 Emergency Planning 1.3.1 Greatly improved Severe Accident Technology 1-2 1.3.2 Superior ALWR Design Capabilities 1-2 2.0 ALWR Utility Requirements - The Technical Foundation 2-1 for ALWR Emergency Planning 2.1 ALWR Design Philosophy and Requirements for Core Damage 2-1 Prevention 2.2 ALWR Emergency Planning Design Criteria and Methodology 2-2 2.2.1 Summary of Criteria and Methodology 2-3 2.2.2 Integral Nature of Criteria and Methodology 2-4 2.3 Containtnent Performance Requirements 2-5 2.3.1 Deterministic Perspective for Severe Accident Requirements 2-5 2.3.2 Probabilistic Perspective for Severe Accident Requirements 2-10 2.3.3 ALWR Performance for Accidents Comprising Existing 2-11 Emergency Planning Basis 2.4 Offsite Dose Requirements 2-14 2.5 Supplementary PRA Evaluation 2 16 3.0 Preliminary Assessment of Passive ALWR Design 3-1 Conformance With Requirements 3.1 Containment Performance Criterion 3-1 3.1.1 Plant Design Characteristics and Features to Address 3-2 Containment Challenges 3.1.2 Containment Evaluation Against ASME Limits 3-2 3.1.3 Assessment of Uncontrolled Release 3-12 3.2 Dose Criterion 3-12 3.3 Supporting PRA Require. ment 3-13 3.4 Conclusions Regarding Passive Plant Conformance to ALWR 3-15 Requirements 4.0 Conclusions 4-1 5.0 References 5-1

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i Tables F

Table 2-1 Potential Severe Accident Containment Challenges 2-6 Table 2-2 Accident Sequence Types Which Tend to Dominate Risk 2-12 for Existing Emergency Planning Basis Table 2 3 Comparison Between WASH-1400 and ALWR 2-13 Requirements Table 2-4 PWR Release Fractions to Primary Containment 2-15 Atmosphere Table 2-5

-BWR Release Fractions to Primary Containment 2-15 Atmosphere Table 3-1 Summary of AP600 Severe Accident Sequence 3-4 Conditions Table 3-2 Summary of SBWR Severe Accident Sequence 3-8 Conditions Appendixes

. Appendix A ALWR Emergency Planning Criteria and Methodology and Updated Containment Performance Requirements Appendix B Summary of ALWR Requirements to Address Severe Accident Containment Challenges Appendix C ALWR Design Characteristics and Features Which Address Dominant WASH 1400 and Subsequent PRA Accident Sequences and Failure Modes Appendix D Assessment of AP600 Design Conformance with ALWR Containment Requirements Appendix E Assessment of SBWR Design Conformance with ALWR Containment Requirements I

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EXECUTIVE

SUMMARY

AND CONCLUSIONS Since 1985, U.S. utilities have been working, through the Advanced Light Water Reactor (ALWR) Program, to develop a technical foundation for a new generation of nuclear power plants. The new plant designs are building on the extensive experience base of existing LWRs in the U.S. and around the world, and will improve upon these existing plant designs in many imponant respects. One aspect of potential improvement is in the area of emergency planning.

This report is intended to establish a thorough and solid technical basis, for use by industry and NRC decision makers,in considering updated emergency planning for ALWRs.

This report provides an integrated treatment of the factors to be considered in developing an updated technical basis for emergency planning for the ALWR. These factors include the technical reasons to update emergency planning for the ALWR, the Utility Requirements Document (URD) emergency planning design criteria and methodology, and the ability of the passive plant designs to meet the design criteria.

The focus of ALWR Program emergency planning work to date is the passive plant. For this reason, the report addresses the Passive ALWR. In general, however, the technical basis for emergency planning, as outlined in the report, could apply to any ALWR standard plant design.

On that basis, the conclusions in the report should not be considered as being limited to passive plants, since they could be adopted for Evolutionary ALWRs as well.

Technical Reasons to Update ALWR Emergency Planning The primary reasons for updating the technical basis for ALWR emergency planning are the-l significant advances in severe accident technology and in plant design capability over the past 15 years. The emerging ALWR designs have superior core damage prevention and severe accident mitigation capability, and the current technical understanding of severe accident risk-is greatly improved compared to that available when the existing emergency planning requirements were estaMished in the late 1970s. Therefore it is appropriate and timely to update the ALWR emergency planning technical basis to ensure that it reflects technical reality for ALWRs.

Design Criteria and Methodology Technical design criteria and associated methodology have been defined for ALWR emergency planning in the areas of containment perfonnance and offsite dose. The complete set of criteria iii

and methodology are specified in Volume III, Chapter 1 of the URD and may be summarized as follows:

Containment Performance Criterion and Methodolocy Plant design characteristics and features shall be provided to preclude core.

damage sequences which could bypass containment and to withstand core damage sequence loads. Containment loads representing those associated with low pressure core damage sequences shall not exceed ASME Service Level C/ Unity Factored Load limits. Accident sequences will be shown not to result in loads exceeding those limits for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, beyond approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, there shall be no uncontrolled release.

The methodology for demonstrating the containment performance criterion includes incorporating design characteristics and features specified in the URD to address severe accident challenges, and use of best estimate evaluations of loads associated with core damage sequences.

Dose Criterion and Methodolouv

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Dose at 0.5 mile from the reactor from a physically-based source term shall not exceed 1 rem for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The methodology for demonstrating the dose criterion includes the use of median dose (i.e., median meteorology) and use of effective dose equivalent with a 50 year commitment.

The criteria and methodology are intended to be applied together and are primarily deterministic.

For a particular ALWR design, it is intended that the criteria and methodology eventually be demonstrated as part of design certification. A supplemental PR A evaluation (10-5 core damage frequency and 10-(', I rem at 0.5 mile)is also required by the URD in support of the two criteria, 1

As part of the PR A evaluation, it is also to be demonstrated that the prompt accident quantitative health objective of the NRC Safety Goal Policy is met with no credit for offsite evacuation prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This reliance on deterministic criteria with PRA as a supplement is consistent with the NRC Severe Accident Policy.

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a Passive ALWR Design Conformance

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Using Standard Safety Analysis Report information, an evaluation was made of the two passive plant designs being developed for design certification against the above URD criteria in order to establish that there will in fact be actual standard design certification applicants which have the capability to pursue ALWR emergency planning. The assessment indicates that both designs, the AP600 and the SBWR will be able to meet the emergency planning criteria with margin.

Conclusions The overall conclusion from the work performed to date on the technical aspects of ALWR emergency planning is that the likelihood and consequences of a severe accident for an ALWR are fundamentally different from that assumed in the technical basis for existing emergency.

planning requirements.15 years ago. Specific conclusions are as follows:

The updated emergency planning technical basis should be utilizedfor the ALWR, The primary reason for this is that the ALWR plant design capability, along with the greatly improved technical understanding of severe accident risk which has evolved over the last 15 years, resuh in significantly reduced ALWR radiological risk.

A strong technical basisfor updated emergency planning exists in the URD. A set of deterministic criteria in the areas of severe accident containment performance anj offsite dose, supplemented by PRA goals, have been developed for ALWR emergency planning and included in Volume III of the _URD. For standard plant designs which demonstrate that these criteria are met, even in the extremely unlikely event of a severe accident the containment has been designed to maintain integrity and thus any radioactivity release will be very slow and small. A period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more exists before reaching offsite -

. dose levels at which the U.S. EPA recommends that actions be taken to protect members of the public.

ALWR designs have excellent potential to meet the design criteria. A preliminary assessment of AP600 and SBWR conformance with the ALWR emergency planning design criteria has been performed and indicates that the designs will l

meet the criteria. The Plant Designers have committed to provide demonstrations as part of design certification that their respective designs meet the criteria.

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4' Section

1.0 INTRODUCTION AND BACKGROUND

1.1 PURPOSE AND SCOPE Since 1985, U.S. utilities have been working, through the Advanced Light Water Reactor (ALWR) Program, to develop a technical foundation for a new generation of nuclear power plants. The new plant designs are building on the extensive experience base of existing LWRs in the U.S. and around the world, and will improve upon these existing plant designs in many important respects. One aspect of potential improvement is in the area of emergency planning.

This report is intended to establish a thorough and solid technical basis, for use by industry and NRC decision makers,in considering updated emergency planning for ALWRs.

4 The objective of the report, " Technical Aspects of Advanced Light Water Reactor Emergency Planning " is to provide an integrated in atment of the factors to be considered in developing an updated technical basis for emergency planning for the ALWR. These factors include the reasons to update the technical basis of ALWR emergency planning, the ALWR Utility Requirements Document (URD) emergency planning design criteria, and the ability of the passive plant' designs to meet the design criteria. The report supporte the conclusion that the likelihood and consequences of a severe accident for an ALWR are fundamentally different than that which is the basis for existing emergency planning requirements.

1.2 APPL lCABILITY The focus of this report is the Passive ALWR. For that reason, Volume III of the ALWR URD[1] specifies emergency planning design criteria for the Passive ALWR. In general, however, the technical basis for emergency planning, as outlined in the following sections, could apply to any ALWR standard plant design. On that basis, the conclusions herein should not be considered as being limited to passive plants, since they could be adopted for Evolutionary ALWRs as well.

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L3 TECHNICAL REASONS TO UPDATE ALWR EMERGENCY PLANNING The primary reason for updating the technical basis for ALWR emergency planning is that, as discussed in this document, the likelihood and consequences of a severe accident for an ALWR are fundamentally different from that assumed in the basis for existing emergency planning requirements. The emerging ALWR designs have superior core damage prevention and severe accident mitigation capability, and the current technical understanding of severe accident risk is greatly improved compared to that available when the existing emergency planning requirements were established nearly 15 years ago. Therefore it is appropriate and timely to update the ALWR emergency planning technical basis to ensure that it reDects technical reality for ALWRs. This is discussed further below.

1.3.1 Greativ improved Severe Accident Technology Existing emergency planning requirements are based on the understanding of severe accidents which was available in the mid to late 1970s. The technical basis for existing emergency-planning is primarily contained in NUREG 0396/ EPA-520/1-78-016[2] published in December, 1978 which in turn utilized severe accident sequence evaluations from WASH 1400[3], the 1975 Reactor Safety Study which was the first comprehensive probabilistic risk assessment (PRA).

The key NRC emergency planning implementation guidance document is NUREG 0654[4]

which is a joint NRC and Federal Emergency Management Agency (FEM A) report published in November,1980. shortly after the Three Mile Island Unit 2 (TMI-2) accident.

Since the time of NRC's promulgation of the emergency planning guidance, a great deal has been learned about severe accident phenomenology and how LWRs respond to severe accidents. In parallel with promulgating the emergency planning guidance, NRC, DOE, various industry organizations, and a number of research organizations worldwide initiated extensive research programs to investigate severe accidents and plant response under. severe accident conditions. A number of these research programs have been completed in recent years, with major advances in understanding of severe accident phenomena. This work has significantly increased the capability to predict LWR severe accident effects, and supports the ability of LWRs to withstand severe accidents to a much greater extent than believed in the 1960s and 1970s.

In August,1985, the NRC Severe Accident Policy [5] was issued. The policy concluded that generic changes to address severe accidents in existing reactors were not warranted, that individual plant examinations should be conducted to look for site or design speciGc risks that 1-2

did warrant attention, and that the design of future reactors should address severe accidents as an integral part of the design process.

J In addition, the NRC developed new PRA tools, culminating in the issuance of NUREG 1150[6),

the 1989 update and replacement of WASH 1400. The technical groundwork of NUREG 1150 together with more recent experimental data and analyses is providing the basis for the ongoing NRC effort to update the design basis source term for ALWRs.

i A comparison of the 1975 WASH 1400 study results (on which NUREG 0396 was based) with NUREG 1150 shows that the accident frequencies and source terms for current plants were

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originally overstated by one to two orders of magnitude [7]. It is also recognized by the authors of the study that the WASH 1400 source term was quite conservative [8]. As a result, the risk j

posed by nuclear plants, even of conventional design, is now understood to be much less than j

these very conservative values which were thought possible when today's emergency planning requirements were formulated. While this report does not address the technical aspects of emergency planning for current plants, it is appropriate to incorporate an updated technical basis into emergency planning requirements for the next generation of plants in order to avoid perpetuating this overstatement of the technical factors of risk.

i 1.3.2 Sunctior ALWR Desien Canabilities All of the above severe accident experience is being brought to bear on the ALWR design. The NRC Severe Accident Policy statement that future reactors address severe accidents as an integral part of the design process is being implemented by the ALWR designers, resulting in a high degree of severe accident protection, including both core damage prevention and accident i

mitigation. Highly effective core damage prevention is a central objective of the ALWR design process and has resulted in design features such as increased margin to core safety limits, use of state-of-the-art man-machine interface systems (MMIS) which greatly simplify the job of the plant operator, greatly decreased dependence on operator action after an accident, and, for the f

passive plants, safety systems which do not require ac power and service water. Based on ALWR design requirements and plant specific PRAs, AL,WR core damage frequencies are expected to be well below 10-5 er year.

p Accident mitigation features have also been heavily emphasized in the ALWR design to provide high assurance of containment integrity and low offsite dose even in the highly unlikely event of-a severe accident. Key accident mitigation provisions include a strong containment with 1-3 4

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significant margin for severe accident loads, features to prevent containment bypass, and -

extremely reliable containment heat removal.

As a supporting requirement for updated emergency planning, the URD specifies a mean frequency of less than 10-6 er year for 1 rem p

dose at 0.5 miles from the reactor, and preliminary assessments indicate that the requirement will be met, with margin.

In addition to the plant designer efforts to incorporate severe accident experience, as part of the ALWR regulatory review process the NRC is developing severe accident requirements. These requirements are being implemented through a number of policy papers and the safety evaluation reports for the standard plant designs. Thus, through the plant designer effort to address severe accidents proactively as part of the design process together with subsequent regulatory review, ALWRs are achieving an unprecedented level of assured severe accident performance capability.

Summarizing, the assured severe accident performance capability of ALWR designs is -

fundamentally different from the limited capability which was assumed in promulgating the existing emergency planning requirements. The key differences involve greatly improved core damage prevention, design features to preclude early containment failure, the adoption of a newly validated source term methodology, and the regulatory assurance of containment performance during severe accidents. These elements combine to provide an extremely low -

likelihood of core damage, and effective mitigation of potential releases even if core damage should occur, greatly reducing the need for offsite protective action. Thus, it is reasonable and prudent to reflect this design capability in the emergency planning requirements for the ALWR.

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P Section 2.0 ALWR UTILITY REQUIREMENTS - TIIE TECHNICAL FOUNDATION FOR ALWR EMERGENCY PLANNING The URD sets policy, principles, and specific design requirements to produce ALWR designs which are reliable, economical, and very safe. With respect to severe accident mitigation (and therefore emergency planning) the URD establishes specific criteria, and associated methodology for demonstrating that the criteria have been met, in the areas of containment performance and offsite dose. In addition, a supplemental PRA evaluation is required by the URD in support of the demonstration of the criteria. Together, these form the technical foundation for emergency planning for the ALWR.

2.1 ALWR DESIGN PillLOSOPilY AND REQUIREMENTS FOR CORE DAMAGE PREVENTION The URD provides for a comprehensive and balanced approach to safety. Highest priority is assigned to the prevention of core damage accidents, both through measures to ensure high accident resistance (e.g., through reduction in safety system challenges) and excellent safety systems to prevent initiating events from progressing to the point of core damage. Excellent mitigation capability is also incorporated in ALWR designs as a defense-in-depth measure to reduce even further the likelihood and consequences of serious accidents.

While the emergency planning requirements focus on containment and accident mitigation capability, it is noted that highly effective core damage prevention is key to overall plant safety and for that reason forms an important part of the technical foundation for ALWR emergency planning. Core damage prevention of the ALWR is rooted in the URD emphasis on simplicity, engineering margin, and human factors throughout the design process.

Examples of requirements in *nese areas include:

Natural circulation decay heat removal from the core

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No recirculation pumps or piping in the BWR Canned rotor pumps, thus eliminating pump seal loss of coolant accident (LOCA),in

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the PWR No loop seals and a minimal number of welds in PWR primary system piping 2-1

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Increased thermal margin in the fuel (15% above regulatory limits)

PWR primary system hot leg temperature of 60(PF or less to reduce steam generator tube corrosion Improved resistance to embrittlement in the reactor vessel Increased reactor coolant system (RCS) coolant inventory which delays core

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uncovery in the event of an accident Decreased dependence on operator action after an accident Improved control room which makes the plant easier and safer to operate Improved accessibility for maintenance Decreased dependence of safety systems on support systems In addition, there are requirements specifically directed toward avoiding core uncovery during shutdown conditions. The ALWR Program reviewed existing shutdown risk issues and the Volume Ill URD provisions to address these issues [9]. Additional requirements were defined as a result of this review. With proper plant specific implementation of these requirements and appropriate administrative controls and procedures provided by the Plant Owner and operator. -

core uncovery during shutdown conditions will not be a credible event.

Finally, accident management requirements exist to prevent as well as limit the extent of core damage. Equipment and procedures for accident management are being considered as part of the plant design process, thus increasing the likelihood of successful recovery actions.

In summary, while the remainder of this report focuses on containment and accident mitigation matters, the ALWR emphasis on core damage prevention and the resulting extremely low probability of an accident are important factors in the consideration of emergency planning requirements.

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- 2.2 ALWR EMERGENCY PLANNING DESIGN CRITERIA AND I

METIIODOLOGY Technical design criteria and associNed methodology have been defined for ALWR emergency planning in the areas of containment perfonnance and offsite dose. The complete set of criteria and methodology are specified in Volume 111. Chapter 1 of the URD[1] and are reproduced in Appendix A of the report.

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2.2.1 Summary of Criteria and Methodoinev A summary of the criteria and the associated methodology is as follows:.

Containment Performance Criterion Plant design characteristics and features shall be provided to preclude core damage sequences which could bypass containment and to withstand core damage sequence loads. Containment loads representing those associated with low pressure core damage sequences shall not exceed ASME Service Level C/ Unity Factored Load limits. Accident sequences will be shown not to result in loads exceeding those limits for approximately 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: beyond approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, there shall be no uncontrolled release.

The methodology for demonstrating the containment performance criterion includes the following:

Incorporate the design characteristics and features specified in the URD to address severe accident challenges.

Demonstrate using best estimate severe accident methods that the loads associated with core damage sequences are no more limiting than the peak LOCA plus hydrogen loads.

Prctection of the ;ontainment for overpressurization beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shall be provided. Overpressure protection may be provided by the size and strength of the containment. On the order of two to three days is judged to be adequate time for actions by the plant staff to bring the accident under control.

Dose Criterion -

Dose at 0.5 mile from the reactor

• from a physically-based source term shall not exceed I rem for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• It is intended that the dose criterion be stated as I rem at 0.5 mile from the reactor (vs. I rem at the site boundary as stated in reference [1].) This will be corrected in the next revision to reference [1].

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The methodology for demonstrating the ' dose criterion includes the use of a probabilistic dose method (e.g., CRAC2 or M ACCS), use of median dose (i.e.,

median meteorology), and use of effective dose equivalent with a 50 year commitment.

The criteria and methodology are primarily deterministic and, for each specific ALWR design, are eventually intended to be reDected in design certification. A supplemental PRA evaluation is also required by the URD in support of the two criteria. This reliance on deterministic criteria with PRA as a supplement is consistent with the NRC Severe Accident Policy [5]. The supporting requirements for the containment performance criterion, the dose criterion, and the supplemental PR A evaluation are described in more detail below in Sections 2.3,2.4, and 2.5, o

respectively.

4 2.2.2 Intecral Nature of Criteria and Methodolorv i

The ALWR emergency planning design eriteria are intended to be applied together with the methodology specified in the URD. Thus, for example, it would be inappropriate to require l

plants to meet I rem at 0.5 mile with a dose evaluation methodology which is more conservative than that in Volume 111. Chapter 1, Section 2.6.5. Application of the criteria with the specified methodology is considered to provide adequa'e margin based on the following:

t The bounding nature of the core damage progression and associated radioactive release specified in the URD methodology, given any credible severe accident.

The very low likelihood of any severe accident in an ALWR. Given this extremely low likelihood, conservatism beyond that noted above is considered unwarranted.

The margin in the I rem,24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dose requirement. The 1 rem is at the lower end of the U.S. Environmental Protection Agency (EPA) range of 1 to 5 rem for evacuationi10], and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides significant margin to perform offsite protective measures.

Additional detail is provided in Sections 2.3 and 2.4 below.

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B 2.3 CONTAINMENT PERFORM ANCE REQUIREMENTS The licensing design basis for the ALWR containment is the traditional set of deterministic loads and load combinations compared against ASME Section 111 limits. Loads associated with events including loss of coolant accidents and the safe shutdown earthquake are combined in the design of the plant. Further, the licensing design basis includes loads associated with generation of l

hydrogen in accordance with 10CFR50.34(fM11].

i in addition to the licensing design basis, the URD includes the safety margin basis which contains requirements that provide margin beyond the licensing design basis. The safety margin basis specifies severe accident requirements in support of the emergency planning containment performance design criterion defined above. These requirements have been developed from a i

deterministic perspective. A probabilistic perspective has also been applied to provide added confidence in the completeness of the deterministic requirements and to make use of the significant body of PRA information. Each of these perspectives is discussed below.

2.3.1 Deterministic Persocetive for Severe Accident Reauirements The severe accident requirements in support of the containment performance design criterion were developed in two steps. In the first step, a set of design characteristics and features was defined to address severe accident containment challenges. A comprehensive set of potential severe accident challenges was identified based on systematic consideration of past PRAs, operating experience, severe accident research and unique design aspects of the ALWR Table 2-1 contains a list of these potential challenges. There are 23 challenges in the table. The first 13 challenges represent events which could occur independent of or precede core damage, such as bypass accidents. The remaining 10 challenges could occur as a result of a severe accident, such as containment pressure loads from a core damage event.

1 A systematic evaluation of the URD was performed to assess the degree to which each of the 23 potential challenges was addressed in the requirements [12]. This systematic evaluation contains a challenge by challenge assessment of the requirements for both the passive PWR and the

)

passive BWR. Appendix B provides a summary of the design characteristics and features specified in the URD to address each challenge. It is concluded from this systematic evaluation that potential challenges, regardless of the extremely low likelihood of the challenge, have been systematically and explicitly addressed in the URD.

i 2-5

i Table 21 -

Potential Severe Accident Containment Challenges r

Cil AI.lENGES/ Fall.,URE MODES TilAT ARE INDEPENDENT OF OR COINCIDENT WITil A SEVERE ACCIDENT 1.

Containment Isolation s

2.

Interfacing System LOCA 3.

Blowdown Forces 4.

Pipe Whip and Jet Impingement 5.

Steam Generator Tube Rupture (PWR) 6.

Anticipated Transient Without Scram (ATWS) 7.

Suppression Pool Bypass (BWR).

8.

Reactor Pressure Vessel (RPV) Failure 9.

Internal Vacuum 10.

Internal (Plant) Missiles 11.

Tomado and Tornado Missiles 12.

Man-Made Site Proximity Hazards 13.

Seismic CIIAl.LENGES/ Fall,URE MODES POTENTIALLY RESULTING FROM A SEVERE ACCIDENT 14.

High Pressure Melt Ejection (HPME) 15.

Hydrogen Detonation / Deflagration 16.

In-vessel Debris-Water Interaction 17.

Ex-vessel Debris-Water Interaction 18.

Noncondensable Gas Generation During Core-Concrete Interaction 19.

Containment Basemat Erosion or Reactor Pressure Vessel Support Degradation During Core-Concrete Interaction 20.

Core Debris in Containment Sump 21.

Core Debris Contact with Containment Shell Liner 22.

Decay Heat Generation 23.

Steam Generator Tube Rupture (SGTR) from Natural Circulation of Hot Gases (PWR) 2-6 y

e m

, ~

~

i In the second step, the results of this systematic evaluation were applied to establish the types of severe accident sequences for which containment response should be evaluated against the Service Level C/ Unity Factored Load limits as specified in the containment performance design

'l criterion. This is necessary since a number of accident sequence types are potentially precluded or otherwise impacted by design.

On the basis of existing plant PRAs, generically applicable severe accident research results, and preliminary passive plant design information, assessments indicated that for the first group of 13 challenges (i.e., containment bypass type challenges), as well as for high pressure melt ejection, hydrogen detonation, steam explosion, basemat erosion or pressure vessel support degradation, core debris contact with shell liner, and steam generator tube rupture from hot gases, the severe accident requirements will provide high assurance of containment integrity [12,13]. This set of challenges includes those which could pose an early threat to containment integrity. The assessments considered the engineered capabilities of the containment systems, i.e., utilize proven technology, function in the environment which the systems will experience, perform functions ' reliably (e.g., incorporate redundancy or passivity), avoid the need for rapid or complex operator actions, minimize dependence on support systems, and be sufficiently independent from the systems whose failure could lead to core damage in the first place so as to avoid significant vulnerability to common cause failure.

An additional factor relative to containment challenges is that, even if it was assumed that j

containment systems do not perform as designed, the plant operators have the ability to perform accident management actions to assure containment integrity. An example in this regard is containment isolation. Accident management procedures have been developed and implemented to address containment isolation as follows[14.15]:

]

1 Confirmation of containment isolation. In the event of a containment isolation signal, emergency operations and/or alarm response procedures call for the operator to confirm that containment isolation valves have closed using valve position indications in the control room. For the ALWR, on the order of hours are expected to be available for the operator to perform any necessary valve closures before significant

.i release of radioactivity into the containment.

Continuous survey of radiation in key plant areas, providing indication of the existence and location of non-isolated or leaking lines. Monitoring systems have 2-7

~_.~. - -

i

-1

\\

been designed for areas such as building ventilation stacks, sampling lines, and sumps j

such that if excessive leakage begins to occur, it can be detected immediately, I

In case of leakage, complementary confirmation of containment isolation including local verifications and/or operator actions when necessary.

1 Generally,it is considered that a relatively small, well-trained team of plant personnel can be effective in accident management for containment isolation as well.as other containment challenges. As noted in Section 2.1 above, the ALWR URD specifies that accident management equipment and procedures be developed as part of the design process.

On the basis that challenges which pose an early threat to containment integrity are being addressed by. well-engineered containment systems, and considering the extremely low likelihood of core damage in the first place as well as the capability of accident management to address problems, accident sequences involving early containment failure are not considered credible in ALWRs.

The remaining challenges (i.e., hydrogen plus LOCA loads, pressurization from debris-water interactions, the potential for core concrete interaction, and decay heat loads) should be considered in establishing the accident sequences for which containment response should be evaluated. In considering these remaining challenges, the effect of plant design characteristics and features on the containment loads should be included. For example, passive containment heat removal does not depend upon any electrical or mechanical equipment which must function i

in a severe accident environment. Thus it is reasonable to assume that passive containment heat removal functions as designed during the accident.

Thus, on the basis of the deterministic perspective, the types of severe accident sequences for which containment response should be evaluated against the Service Level C/ Unity Factored Load limits are as follows:

Core Damage i

Rapid core damage progression, i.e., bernning at approximately one hour after the initiating event, and or curring over a tin:e frame of a few hours 1

Large scale core melt and associated pas and aerosol release Steam release out of phase wi+ ne.osol release 2-8

Consideration of in-vessel core damage and the potential for ex-vessel core damage Reactor Coolant System Condition Limited aerosol plateout in the RCS A vapor pathway exists in the RCS (i.e., from the core to the containment 1

atmosphere)

RCS is depressurized to about 1(X) psig or less Containment Condition Containment is isolated and otherwise intact at the time of core damage (i.e., no containment bypass has occurred)

Water exists in the reactor cavity / lower drywell prior to or immediately upon reactor vessel lower head penetration Containment systems are functioning as designed (heat removal, fission product removal, hydrogen control, pH control)

Containment leaking at design basis leak rate (or leak rate proportional to pressure)

Secondary Building Condition Containment leakage released into secondary building volume Building volume mixing and exchange with the environment is based upon plant design characteristics (e.g., safety envelope leakage)

Building volume bypass pathways taken into account' As noted in Appendix A, the above severe accident sequence types are specified in URD Chapter

5. Section 2.6, Criteria and Methodology for ALWR Emergency Planning. The loads associated -

with these severe accident characteristics must not exceed specified ASME limits for.

approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ASME Service Level C/ Unity Factored Load limits were specified in order to provide high confidence that containment leakage would, at most, be a linear extrapolation of design basis leakage. This is based on several factors including:

Service Level C assures stress levels below yield in steel containrnents, and unity-factored load assures limits on linear deformation in conemte containments; leaks are not expected in membranes with such small deformations.

A review of experimental and analytical evidence [16] which indicates that there is essentially no increase in penetration leakage under severe accident conditions up to Service Level C/ Unity Factored Loads.

2 -

Nuclear plant containment leak test data indicating that, for pressure increases up to design pressure, leak rate does not exceed a value proportional to the pressure [16].

An additional point is the fact that the fission product mass is almost exclusively paniculate [17]

and as noted in reference [17], aerosol plugging of leak paths is expected which should significantly reduce the actual mass leaked during an accident compared to that assumed in design basis leakage.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is consistent with the I rem 24-hour limit specified in the dose criterion and allows appropriate time for ad-hoc public protective actions.

No uncontrolled release beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has been specified to provide additional margin for emergency planning, While approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is considered more than adequate for ad hoc evacuation, it is desirable to avoid long-term overpressure failure.

2.3.2 Probabilistic Persocctive for Severe Accident Reauirements PRA has been applied to confirm that the appropriate severe accident sequence characteristics are being considered in the evaluation of containment response against the Service level C/ Unity Factored Load limits. From a probabilistic perspective, the URD requires that functional sequence types with frequency greater than approximately 10-7 per year be evaluated for containment response. Lower frequency functional sequence types are to be reported for discussion (i.e., identification of design characteristics and features which are credited in reaching this low frequency), but are not required to be evaluated for containment response. This 10-7 per year frequency threshold for sequence types to be evaluated for containment response is consistent with the NUREG 1420118] limit for insignificant risks and with previous regulatory guidance (e.g., Standard Review Plan guidance to evaluate potential accidents from hazards in the vicinity of the plant site which exceed approximately 10-7 per year.) Also, consideration of functional sequence types greater than approximately 10-7 er year helps provide assurance that p

the cumulative effects of such sequence types will not exceed the 10-6 er year probability goal p

for offsite consequences, As described in Section 3'below, review of the passive plant designs indicates that accident sequences which are of the order of 10-7 er year or greater involve core damage into an intact p

containment with the reactor coolant system at least partially depressuri7ed and containment 2-10 1

4 systems functioning as designed. That is, the characteristics of these sequences from the PRA are similar to the characteristics defined from a deterministic perspective.

2,3.3 ALWR Performance for Accidents Comorisine Existine Emereency Plannine Basis Given the above ALWR design requirements, it is useful at this point to examine the accident types and failure modes which dominated the risk in the existing emergency planning basis and the manner in which these sequence types and failure modes are addressed by the ALWR design.

At the time of the development of the existing emergency planning basis, defined in NUREG 0396[2], WASH 1400[3] provided the most detailed perspective on the types of accident scenarios which made up the collection of " Class 9" events. Accident scenario types and c

containment failure modes which dominated the risk in WASH 1400 are summarized in Table 2-2, and it is these events which formed the basis for existing emergency planning requirements.

Also included in Table 2-2 are important challenges identified as a result of PRA ' work subsequent to WASH 1400. More recently, improved understanding of severe accident behavior i

as well as modifications to plants and procedures have changed the characteristics of accident scenarios which dominate risk compared to WASH 1400. This applies to a significant extent in 4

existing plants and to an even greater extent in ALWRs. ALWR design requirements directly address those events which dominated the risk in WASH 1400 and subsequent PRAs. Appendix C describes the Passive ALWR design characteristics and features that have been provided to preclude or accommodate the accident sequence types and failure modes listed in Table 2-2 as contributors to core damage and containment failure.

I It is apparent from this comparison that the Table 2-2 WASH 1400 issues which dominated the risk and formed the basis for existing emergency planning, as well as subsequently identified containment challenges (shown in Table 2-2 with a footnote), have been addressed explicitly in the ' ALWR requirements. Therefore, the characterization of risk for ALWRs will differ significantly from a WASH 1400 type characterization, or even from the characterization in subsequent PR As. Table 2-3 provides clear illustration of this difference in risk characterization.

It is the ALWR risk characteiization, which reflects the above design characteristics and features and the improved phenomenological understanding of severe accidents, that should be used in

.l formulating ALWR emergency planning regulatory requirements.

2-11

Tabis 2-2 Accident Sequence Types Which Tend to Dominate Risk for Existing Emergency Planning Basis DOMINANT ACCIDENT SEQUENCES LEADING TO CORE DAMAGE

• PWRs BWRs LOCAs (large or small)

LOCAs (large or small)

Loss ofinjection (AD. SD)

Loss ofinjection (AE, SE)

Loss of recirculation (AH, SH)

Vessel Rupture (R)

Vessel Rupture (R)

Interfacing LOCA (V)

Transients Transients Loss of secondary heat removal (TML)

+ Loss of containment heat removal (TW)

Station blackout (TMLB')

Loss of all injection (TQUV)

ATWS (TKQ)

ATWS (TC)

Shutdown Conditions **

Shutdown Conditions **

i POTENTIAL CONTAINMENT FAILURE MODES

• PWRs BWRs Overpressure (S)

Overpressure (D)

In-Vessel Steam Explosion (a)

In-vessel Steam Explosion (u)

Hydrogen Combustion (S)

Containment isolation ( )

Containment isolation (S.c)

Liner Melt-Through**

Basemat Penetration (c)

Ex-Vessel Steam Explosion **

Direct Containment Heating **

Overtemperature*

• Steam Generator Tube Rupture **

Notes:

Characters in parentheses are sequence and failure mode designators from WASH 140()

lssues which were identified in PRA work subsequent to WASH 14(Kl.

2-12 l

l i

Table 2 3 Comparison Between WASII-1400 and ALWR Requirements 5

Mean Frequency of Mean Frequency of Mean Core Exceeding Exceeding Damage Frecuency 1 Rem Promnt Effects Dose WASH-14(X)

(doses at 10 miles from reactor)

- 1.5 x 10-4 /yr

-4 x 10-5 /yr(I)

~4 x 10-6/yr (2)

ALWR Requirements (doses at 0.5 miles

< 10-5 /yr

< 10-6 /yr (3) from reactor)

Passive Plant (doses at 0.5 miles from reactor) 6/yr(4)

<10~7 /yr (4)

<10'8 /yr(4) l Notes:

(1) Based on mean core damage frequency of -1.5 x10~4/yr (i.e.,3 x the WASH-1400[3]

median value of 5 x10 5) and, from Figure I-11 of NUREG-0396[2], -0.3 conditional probability of exceeding I rem at 10 miles.

(2) Based on mean core damage frequency of -1.5 x10'4/yr and, from Figure 111 of NUREG-0396,-0.03 conditional probability of exceeding prompt effects dose at 10 miles.

4 (3) Functional sequence types which could threaten containment must be less than ~10 /yr.

(4) Preliminary estimates based on initial AP600 and SBWR PRA work.

l 2-13

2.4 OFFSITE DOSE REQUIREMENTS As part of the technical foundation for emergency planning in ALWRs, an offsite dose limit is required. A maximum dose of I rem at 0.5 mile from the reactor for a period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the beginning of fission product release to containment has been specified on the basis of EPA guidancello) for actions to protect the public in the early phase of a nuclear incident. The approximately 24-hour period is considered to provide significant margin for accident detection, notification of the public in the community around the site, and offsite protection measures such as ad hoc evacuation.

The methodology for demonstrating the 1 rem dose criterion is based on deterministic analyses.

The source term to be utilized by the design certification applicant as part of the demonstration is a physically-based source term. A physically-based source term is proposed for design basis applications for the ALWR as well as for emergency planning use. It specifies fission product -

release timing and magnitude to containment. chemical form of the fission products, fission product removal from containment, and fission product holdup in the secondary building. The physically-based source term is based on fission product release and removal phenomena from actual ALWR core damage sequences which, although extremely low in probability, are considered credible for purposes of defining the source term. The physically-based source term has been defined so as to envelope potential source terms from such sec,uences i.e., sequences having the characteristics defined above in Section 2.3. Thus, the physically-based source term provides signi6 cant margin beyond the actual fission product release which would be expected if a core melt accident were assumed to occur at an ALWR. The physically based source terms which were developed by the ALWR Program in early 1992 for the passive PWR and BWR are given in Tables 2-4 and 2-5[17]. Additional ALWR Program work, mainly on fission product removal, was submitted to NRC in 1993 (for example see reference [19}). NRC is presently working on an updated design basis source terml20] which is similar to the ALWR physically-based source term. The source term to be used by design certification applicants will reDect the resolution of differences between the NRC and ALWR source term, which is being addressed as of this writing. Major differences are not expected.

The methodology specified for the dose evaluation is similar in concept to what is typically done in Level 3 PR A evaluations, e.g., a CRAC2 or MACCS calculation. Median meteorological conditions are specified on the basis that the ALWR physically-based source term has significant 2-14

Table 2-4 PWR Release Fractions to Primary Containment Atmosphere

• 0-l hr.

1-5 hr.

5 hr.* *

• 5-24 hr.

Coolant Early Ex-Late Nuclide Activity In-Vessel Vessel In-Vessel Total Nobles 0.80 0.20 1.0 I

0.38 0.17 0.55 Cs 0.30 0.18 0.48 Te 0.08 0.03 0.11 Sr. Ba 0.004 0.004 Ru 0.(X)4 0.004 Remainder 0.(XXX)4 0.(XXK)4 Table 2-5 BWR Release Fractions to Primary Containment Atmosphere

• 0- 1 h r.

1-3 hr.

3 hr.* *

• 3-24 hr.

Coolant Early Ex-Late Nuclide Activity In-Vessel Vessel In-Vessel Total 0.80 0.20 1.0 Nobles I

0.30 0.20 0.50 Cs 0.23 0.18 0.41 0.03 0.09 Te 0.06 Sr. Ba 0.003 0.003 Ru 0.003 0.(X)3 Remainder OJXXX)3 0.00003 Notes:

All numbers are fraction of origin:d core fission prixluct inventory.

Coolant activity makes a negligible contribution to the source term from a core damage event and so is not -

included here.

All nobles released either early or late in-vessel. Remaining fission products retained in yuenched debris or scrubbed through overlying water pool in reactor cavity (PWR) or drywell (BWR).

2-15 t

N 4

t margin to that expected from any credible ALWR core damage sequence source term as noted above. Thus the combination of median meteorology and the physically-based source term-bounds most core melt sequences. The site meteorology which has been specified for design certification applicant dose calculations is that which is in the URD Key Assumptions and Groundrules for PRA. This site was selected to have a Chi /Q greater than 80 to 90 percent of U.S. operating nuclear plant sites to provide siting Dexibility for the ALWR. Committed effective dose equivalent (CEDE) is to be used (as opposed to the older whole body concept) on the basis of the recent EPA report [10] and revised 10CFR20[21].

2.5 SUPPLEMENTARY PRA EVALUATION As described in Sections 2.3 and 2.4, the two ALWR emergency planning criteria, containment performance and offsite dose, stress a deterministic approach. To complement the deterministic approach associated with the criteria, a supporting PRA evaluation has also been specified. The PRA is required to demonstrate that core damage frequency is less than 10-5 er year and that p

the cumulative frequency for sequences that result in greater than I rem for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 0.5 mile from the reactor is less than 104 As part of the PRA evaluation, it is also to be per year.

demonstrated that the prompt accident quantitative health objective of the NRC Safety Goal Policy [22] is met with no credit for offsite evacuation prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The PR A goals are not emergency planning criteria, nor is it intended that the goals be made part of design certification or any other rulemaking. Rather the PRA is intended to demonstrate the integrated effectiveness of the two emergency planning criteria and to serve as a tool for the Plant Designer for refining and optimizing the design. Also, the PRA will provide additional confidence to the NRC in the overall safety of the design and in the margin to NRC guidelines on core damage frequency and large release. Finally, the NRC Safety Goal Policy quantitative health objective provision demonstrates that an acceptable level of radiological risk to the public, as defined by the NRC Safety Goal Policy, can be achieved with ad hoc evacuation which can be accomplished with significant margin within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

As noted in Section 2.2 above. this approach of deterministic criteria, with PRA used as a supporting evaluation. is consistent with the industry interpretation of the NRC Severe Accident Policy [5] which states that safety acceptability should be based on an approach which stresses deterministic engineering analysis and judgment, complemented by PRA.

2-16

Section 3.0 PRELIN11 NARY ASSESSN1ENT OF PASSIVE ALWR DESIGN CONFORNIANCE WITH REQUIREN1ENTS Two passive plant designs, the Westinghouse AP600 and the General Electric SBWR, have been submitted to NRC for design certi6 cation under 10CFR52. A preliminary assessment of these standard passive plant designs has been conducted to determine the degree to which they meet the ALWR emergency planning design criteria. The assessment is based on a review of the AP600 Standard Safety Analysis Report (SSAR)[23] which was completed in June,1992, and the SBWR SSARl24] which was completed in February,1993.

While this preliminary assessment has been conducted for the passive plants, both ABB-CE and General Electric have committed to perform similar assessments for their evolutionary designs, System 80+ and ABWR, which are presently in the design certification process.

3.1 CONTAINNIENT PERFORN1 ANCE CRITERION The containment performance criterion for emergency planning appears in Chapter 1 of the URD and is repeated in Appendix A and discussed in Section 2 above.

The steps used for the preliminary assessment of compliance with the criterion were as follows:

f 1.

Confirm that the design meets the requirements of the URD, Chapter 5, Section 6.6.2.1 by performing a comparison between the passive plant design characteristics and features and the requirements identified in Reference 14 and summarized in Appendix B.

2.

Confirm that containment loads representing those from core damage sequences do not exceed ASME limits specified in the URD Chapter 5.

Section 6.6.2.2 for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under realistic severe accident assumptions.

3.

Confinn that no uncontrolled release will occur beyond approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3-1

~

3.1.1 Plant Desien Characteristics and Features to Address Containment Challenges The preliminary assessment was performed by reviewing the respective SSARs to confirm, for each containment challenge, the existence of specific design features or characteristics to fulfill the key URD requirements associated with the challenge. The list of challenges and associated requirements as summarized in Appendix B was used for this review. A requirement was considered met when an explicit reference to the system, feature, or characteristic was made in the SSAR.

Table D-1 in Appendix D summarizes the results of the preliminary assessment for AP600. This table lists the challenges and associated requirements from Appendix B, and identifies in i

brackets the sections of the AP6(X) SSAR which address each requirement. With the exception of the items identified in Table D-2, specific SSAR design features or capabilities have been identified in msponse to the requirements.

t Table E-1 in Appendix E summarizes the results of the preliminary assessment for SBWR, This table also lists the challenges and associated requirements from Appendix B, and identifies in brackets the sections of the SBWR SSAR which address each requirement. With the exception of the items identified in Table E-2, specific SSAR design features or capabilities have been identified in response to the requirements.

On the basis of the preliminary assessment, it is expected that the AP6(X) and SBWR will be I

able to demonstrate that the requirements of Chapter 5, Section 6.6.2.1 of the URD are met.

While there are several exceptions which require additional action to resolve, these exceptions are not major and are expected to have little,if any, impact on the design, 3.1.2 Comainment Evaluation Against ASME Limits As discussed in Section 2, for plant designs which meet all of the URD provisions related to f

containment challenges, the severe accident sequences for which containment performance' should be evaluated are low pressure core melts into an intact containment with the RCS at low pressure and containment systems functioning as designed.

A preliminary assessment of AP6(H) containment performance against ASME limits has been performed by evaluating a low pressure core' melt sequence from the AP6(X) PRA. In this Base i

3-2 1

1

)

i Case sequence presented in the PR A, the accident is caused by a 4 inch LOCA, with successful j

depressurization but failure of the internal refueling water storage tank (IRWST) to inject due to check valve failure. The core is uncovered at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the vessel fails at approximately H

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The debris is cooled in the reactor cavity to less than 8(X) F but temporarily reheats to 1340 F after the water present in the reactor cavity is boiled off. Condensation from the passive containment cooling system (PCCS) eventually results in the IRWST water overflowing into the '

reactor cavity, cooling the debris. Ilydrogen produced from metal oxidation is controlled by igniters. The containment peak pressure and temperature are 47 psia and 368 F respectively, well under the design pressure of 60 psia. The conditions corresponding to Service Level C have been determined in the SSAR to be 104 psia at 400 F, and the ultimate capacity has been determined to be 135 psia at 4(KFF. Thus, the AP600 containment design provides substantial margin to loads which would be expected should a severe accident occur.

In addition, variations on the Base Case sequence as well as other sensitivity sequences were analyzed. The variations on the Base Case sequence were taken from dominant accident scenarios determined in the Level 1 PRA. These additional analyses address the sensitivity of the results to ex-vessel debris coolability, containment pressurization due to core concrete interaction, hydrogen igniter operation, creep rupture of reactor coolant piping system, and v

availability of PCCS water. A summary of the sequences analyzed and the corresponding containment pressures and temperatures are presented in Table 31.

i Based on the results in Table 3-1, sequences involving low pressure core melt into an intact containment with containment systems functioning as designed meet the Service Level C limit with significant margin. Even the sensitivity sequences in Table 3-liin which containment systems are assumed to have degraded performance, meet the 24-hour Service Level C criterion.

Three of the sequences analyzed in the AP600 PRA are associated with the containment bypass l

and isolation failure release type. Passive design capability to preclude or accommodate these types of events has been provided. On the basis of this design capability, this release type is not considered credible. Further,its PRA frequency is roughly an order of magnitude less than the URD 10-7 per year threshold. It is also noted that the three sequences presented in Table 3-1

- represent a bound of eight PRA sequences which have a range of release timing. The majority of these eight sequences has release beginning after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with about half having release after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The frequency of release before 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is about 8 x 10-9 er year.

p 1

From this preliminary assessment, there is confidence that the AP600 will be able to meet the

]

ASME Service Level C limits.

b 3-3

~

Table 31

SUMMARY

OF AP600 SEVERE ACCIDENT SEQUENCE CONDITIONS

.' CONTAINMENT :

t SITE BOUNDARY :

1 REPRESENTATIV E -

- MAXIM UM 1

REM ARKS --

RELEASE TYPE :

FREQUENCY;-

DOSE LEVELS FOR s

-PRESSURE AND.-

-OF RELEASE 1 JRELEASE TYPE.!

SEQUENCES

' TEMPERATURE

TYPE (Per.

? d4 HRS Amt Ct)RE ::

DDI.7fA.. er' Year)

-- ndtscti MEDIAN y= c=*w m m e

-.pgg g {

Dase Case DCI: 1.oss of Ctmlant Accident 47.1 psia Release associated with 2.5 s 10 0.07 -

(IIX'A) with In-Containment Ref ueling Water 368'E the leakage fnnn an Storage Tank Check Valve Failure intact containment that is not pressuri. red atuve the design pressure.

VRPl: Vessel Rupture 45 psia 296'F SLP: Small 1.OCA with Passive Residual liest 26.1 psia Removal (RllR), Core Makeup Tanks (CM I )

215'F Fail. Automatic Depressueintion System (ADS)

Not Actuated MLP: Medium LOCA. Passive RilR I; ails.

36 3 psia. 296*F ADS Fads IGN: Igniter i ailure.

47 psia < pike Peak pressure from hydrogen BC1 + 67'Tr of cladding is reacted in vessel and 800*F spike burn does not exceed design hydrogen igniters are turned off.

29 psia.260*F pressure.

at 24 hrs CC: Passive Containment Cooling System 68 psia spike Assuming constant rate of Release associated with 76s1040 o pg (PCCS) Water Failure.

at 12 hrs pressurization by non-leakage from a DCl + failure of PCCS water en outside of 80 psia condensable Fases generated contairunent w hich is '

shell three out of four CMT and accumulaters 296*F st 48 hrs from CCI. containment failure overpressurized by available.

is espected greater than 4 2 noncondensible gas days.

g,y i

1 k

-i 1

a.-

, ua m:

- c-

... ~

. ~..

Table 3-1

SUMMARY

OF APMMI SEVERE ACCIDENT SEQUENCE CONDITIONS (Cont'd)

CONTAINMENT -

SITE IlOUNDARY-REPRESENTATIVE.

' %f AXIMU31 RESIA R KS RELEASE TYPE FREQUENCY

. DOSE LEVELS FOR

~ SEQUENCES PRESSURE AND ~

.0F RELEASE RELEASE TYPE TEM PI;R ATUR E TYPE (Per c4 nus AFTra coat 7,M%",. e ece Year)

. DAktACE) AlEDlAN tyy omm m rm e -

.. pgg g i)KP: PCCS I ailure. CoolaNe Debris.

75 4 psia Containment has reached a Release associated with 5.6 x 10 h 0.12 CC + four out of four CM T and accumulators 314F steady state at 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />. not leakage from an intxt available.

expected to fail.

containment that has been piessunted above design but below Service I evel C n essure.

LIWl: 1.oss of l'eedwater and Containment 25 7 psu Release associated with 2 x 10-8

>l Isolation. Passive RilR CMT and Al)S I ail 196*l:

the leakage from a contairunent that is bypassed or has not been isoiated SG~I R: Steam Generator Tube Rupttue 235 piia Core not predicted to become (SGTR), Steam Generator Safety Valve Stuck 188*F uncovered until after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Open.

Passive RIIR actuated on high temperature signal in hot leg rather than a low steam generator level.

ADS fails.

SG2: SGTR + Passive RilR failure, ADS 50 psia spike fails.

1140'l: spike 22 psia 260T at 32 hrs

4 Table 3-1 SUhlNIAlW Ol' AI'600 SEVERE ACCIDENT SEQUENCE CONDITIONS (Cont'd)

. CONTAINMENT SITE BOUNDARY

. REPRESENTATIVE -

%I A XI\\lN31 REAIA R KS

' RF. LEASE TYPE '

FREQUENCY:'

REl. EASE TYPE DOSE I.EVEI.S FOR SEQdENCES '

- PRESSURE. AND.

0F REllASE'

- TESIPER ATUR E.

TYPE (Per -

1. 24 nus AFTra cour U. U cOnenm e ecs

- YearI~

DOSE (Rem)

~

rmtsco MEDIAN 7 **"$t "? m '

SENSI~IIVITY ANAI.YSES PERFORalEI) INDEPENDEN'II.Y OF PROIIAltll.lTY OF OCCURRENCE CR; Creep Rupture Siie Sensitivity.

50 psia spike Contaisiment pressure reaches 5 x 10-9 Italf square-foot creep rupture in RCS assumed.

1250*F spike equilibrium below design 26 psia.

pressure.

215*F at 28 hrs I)R Y: Passive Core Cooling System I ailure.

45 psia spike Assuming constant rate of

< 10*

332*F spike core <encrete interaction the Case BC1 assuming failure of all passive core 32.7 psia-basemat fails at 8.R days:

cooling system water sources.

260*F overyvessurization occurs at at 25 hrs 4.9 days.

CilF: Debns Coolability Sensitivity.

42 psia spike Basemat fails due to CCI at Not Calculated 550*F spike 26.5 days.

Case UCI assuming the debris not coolable even 34 8 psia.

Ovegvessuniation occurs at though cavity is ikxxted 280"F 16.7 days.

at 251.rs

t A preliminary assessment of SBWR containment performance against ASME limits has also been performed by evaluating low pressure core melt sequences from the SBWR PRA. The two base sequences LPL-SN and LPE-SN are similar in. nature. The initiating _ event is an inadvertently open relief valve which depressurizes the reactor. This initiating event is very similar to a LOCA and is used to determine the consequences from a LOCA. The reactor scrams and the Main Steam Isolation Valves (MSIVs) close. The feedwater pumps trip and the automatic depressurization system (ADS) opens the remaining safety relief valves (SRVs) and the depressurization valves (DPVs). All high and low pressure injection systems are assumed to fail. No credit is taken for operation of the Isolation Condenser (IC) units. At approximately 50 minutes into the event, core uncovery occurs which eventually leads to reactor vessel lower head penetrations failure at about 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Corium is deposited on the lower drywell Door which causes the flooder to open due to high local temperature. The debris is quenched and core concrete interaction does not occur. Steam generation in the lower drywell leads to further increase in the containment pressure until the PCCS heat removal capacity equals the decay heat generated by the core debris. The long-term containment pressure is about 0.56 MPa (80 psia) which is below the wetwell vent pressure setroint of 0.93 MPa (135 psia). The containment temperature is approximately 530K (495 F). The conditions corresponding to ASME Unit Factored Load and Service Level C have been determined in the SSAR to be 118 psia at 500 F, and the ultimate capacity has been determined to be 215 psia at 500 F. Thus, the SBWR containment design provides substantial margin to loads which would be expected should a severe accident occur. The LPL SN sequence is the same as LPE-SN except that one gravity drain cooling system pool injects water into the vessel delaying reactor vessel failure by about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In both cases, nonnal containment leakage is the only mode of fission product release.

In addition, other sequences were analyzed to address the sensitivity of the results to vessel rupture, high pressure core melt, limited debris coolability, failure of the flooder, failure of containment heat removal and dominant release path. Seventeen additional sequences were evaluated. Table 3-2 presents a summary of the nineteen sequences and the corresponding containment pressures and temperatures.

Based on the SBWR SSAR analyses, sequences involving low pressure core melt into an intact containment with containment systems functioning as designed meet the Service Level C limit with significant margin.

Some of the sequences analyzed in the SBWR PRA include system failures beyond the failures in the two basic sequences, e.g., high pressure melt ejection and containment bypass. Passive 3-7

1 Table 3-2

SUMMARY

OF SilWR SEVERE ACCIDENT SEQUENCE CONDITIONS CONTAINNIENT REPRESENTATIVE 3t A XIhlUNI R ESf A R KS REl. EASE TYPE FREQUENCY SEQUENCF.S PRESSURE AND GF RELEASE TEMPERATURE TYPE (f*er N.NANeisElN.

. Yea r) l,0," * '

SE(JUENCES WITil VESSEL, FAII,URE A l' I.OW PRESSURE LPE-SN: Inadvettently open relief valve (IORV),

0 56 h1Pa (80 psia)

Release associated with 7 x 104 htSIVs close, feedwater pumps trip. ADS opens, 530K (495F) the leakage imm an intact high and low pressure injection fail. no credit taken containment that is not for IC. Dnwell spray s fail, flomiers operate after pressurized above Service vessel failure I.evel C I PL-SN: Same as I.PE-SN etcept that one GDCS 014) h1Pa (87 psia)

Release associated with 6.4 x 104 pool injects into vessel, delaying vessel failure by

~530K(495 F) the leakage from an intact i

apprnximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

containment that is not pressurized above Service level C ITE-SCV: Same as 1 PE-SN assuming that the 0 91 h1Pa (135 psia) llem of concrete Sembbed release from 1.1 x 104 debris is not coolable.

before venting ablation in lower wetw cli vent 600K (620F) dr> well after 80 at 28.7 hrs hours LPE-SCD: Same as LPE-SCV assuming vent is 10 hlPa (145 psia) @

Release through a fai!cd 1 6 X 10-9 not opened at 28.7 hrs.

the time of head failure dqwell head 600K (620F) at 31.2 hrs 4

IPL-SCV Same as I.PL-SN assuming that the 0 93 h!Pa (135 psia)

! 46m of concrete Sembbed release from 1.1 X 10 debris is not cootable.

before venting ablati(m in lower wetwell vent

-600K (620F) diywell after 80 at 36 6 hrs inurs

Table 3 2

SUMMARY

OF SHWR SEVERE ACCIDENT SEQUENCE CONDITIONS (Cont'd)

CONTAIN% LENT REPHESENTATIVE M AXIMU31 RE AI A RKS RELEASE TYPE FREQUENCY SEQUENCES :

PRESSURE AND OF RELEASE TE3IPERATURE TYPE (rcr

's".=M7o7 "uU $5.

l'e n r) 535Elua"*

SEQUENCES WITil VESSI{I, Fall,URE AT I OW PRESSURE - (Cont'd) 1.PL-SD: Same as LPE-SN except that one htSIV Containment

< 10- 10 fails to cime Bvrass LPE-SWV: Same as I.PE-SN assuming failure of Scrubled reicaw from 2.3 x 10' 3 0 containment heat removal (both PCCS and wetwell vent at 40 2 hrs Suppress n Pool Coolinen SE(JUENCES Willi VESSEL. Fall.URE AT INTERMEDIATE PRFSSURE MPE-SN. Same as LPli-SN except that DPVs f ail 0.73 MPa (106 psia)

Re! case associated with

<1040 in ogwn leading to a medium pressure core melt

-470K (40011 the leakage from an intact containment that is not pressurized above Seivice LevelC MPL-SN: Same as I.PL-SN escept that DPVs fail 0 89 MPa (129 psia)

Release awociated with

< 10- 10 to open leading to a medium pressure core melt.

-470K (400F) the leakage from an intact containment that is pressurized below wetwell vent pressure MPE-SCV: Same as LPli-SCV execpt that DPVs 1.65m of concrete Scrubbed release from

< 10- !"

fail to open leading to a medium pressure core melt.

ablation in lower wetwell vent dry we11 after eighty at 38.9 hrs hours

(

MPL-SCV: Same as LPl.- SCV except that DPVs Scrubbed rekase frmi

<1040 i

fail to open leading to a medium pressure core melt.

wetwell vent at 39 7 hrs MPL-SCD: Same as I.PL-SCV except that DPVs Release through a failed H.I x 10- 4 0 fail to open leading to a medium pressure core melt.

drywell head at 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> and vent is not opened

Table 3-2

SUMMARY

OF SilWR SEVERE ACCIDENT SEQUENCE CONDITIONS (Cont *d)

CONTAINNIENT

' REPRESENTATIVE SIAXIMUSI RESIA R KS REl, EASE TYPE FREQUENCY SEQUENCES PRESSURE AND OF RELEASE TEMPERATURE TYPE (Per NNSo$s"u~nWn.

Yenr)

IlE*

• SEQUENCES WITil VESSEI, Fall,URE AT IllGli PRESSURE llP-N: less of site (uwer, MSIV close and reactor OM MPa (92 psia)

Release asmciated with 7.7 x 10-9 scrams, feedwater pump trip. high pressure injection 450K (3500) the leakage from an intact fails, ADS fails, but SRVs cycle at setpoint containment that is not pessure. IC inoperable. Ihlers and dqwell sprays pressurited above Service operate after vessel failure.

I.e vel C llP-SG: Same as llP-N except that dqwell sprays 0 68 MPa (99 psia)

Leakage iluuugh the 5.5 x 10*#

fail.

800K (98017) drywell head due to high temperature seal degradation at <24 hrs IIP-SIG Same as IIP-N escept that flooders and 0.57 MPa (83 psia)

Some CCI occurs Leakage through the

<10- 8 0 drywell sptays both fail

-800K (980F) due to ihnders drywell head due to high failure. 0 05m of temperature seat concrete ablation in degradation @ 27.7 hrs lower drywell before debris is quenched.

IIP-SCG: Same as llP-N except that drywell sprays 0.72 MPa (105 psia) 013m of concrete I.eakage through the 4 3 x 10 fail and debris is assumed not to be coolable.

-800K (980F) ablation in lower drywell head due to high -

e drywc!) after eighty temperature seal htmrs degradation at <24 hrs

Table 3-2

SUMMARY

Ol'SRWR SEVERE ACCil)ENT SEQUENCE CONDITIONS (Cent'd1 CONTAINMENT REPRESENTATIVE MAXIMUhl RE31 A R KS RELEASE TYPE FREQUENCY OF SEQUENCES PRESSURE AND RELEASE TYPE TESIPERATURE frer Year)

"s Y E /sT [$ E Y5555::U"'

VESSEI, RUPTURE SEQUENCES V R-SN: large 1 OCA in RPV lower heait, reactor 036 MPa (53 psia)

Release associated with 3.9 x 104 scrams, ADS fails, all modes of injection fail, 500K (440 F) the leakage from an fkoders operate after core relocates in lower drywell.

O 80 hrs intact containment that is not pressuriicd above desien pressure VR-SX: Same as VR-SN except that containment Containment fails at

< 10- 30 is assumed to fail u hen core debris is expelled frinn 500K 1440F) @ 80 hrs

< 24 hrs the R PV.

VR-SCV: Same as VR-SN except that debris is 18m of concrete Scrubbed release from 6.3 x 10- 80 assumed tot to be coolable.

ablauon in lower wetuell vent at 34 hrs dry well after eighty hours NOTES - MAAP-SIMVR SEQUENCE NAMING CONVENTION:

First two or three characters (Hase Sequence):

1 PC ~ Iisw Pressure' Core ' Melt w'ith loss of Short4enn Coolant Makeup MPI, Medium Pressure Core Melt (depressurization through SRVs only) with

~

I.Pl. Low Pressure Core Melt with less of leng-Term Coolant Makeup I.oss of Ixng-Term Coolant Makeup M PE Medium hessure Core Melt (depressurization thinugh SRVs only) with VR Vessel Rupture less of Short-Term Coolant Makeup IIP liigh Pressure Core Melt Characters in lietween First Two and I,ast Characters (Failures):

C~ ~ 'ijmit'ed ikbrh Coolniihy' ~ ~ ' ~

W Failure of Containment Ileat Removal tBoth PCCS and Suppression Pool F

Failure of the Flooder Cooling)

S Failure of the Drywell Sprays to Operate I,ast Character (Dominant Release Path):

N ~ ~ N6tmal Contain~rnent liak' age ~- ~

D Drywell llead Failure

~

V Suppression Chamber Vent X

Early Containment Failure G

Leakage hugh Drywellllead Seal D

Containment Bypass

design capability to preclude or accommodate these types of events has been provided. On' the basis of this design capability these release types are not considered civdible. Their frequencies are confirmed to be an order of magnitude or more below the 10-7 er year URD thirshold.

p 3.1.3 Assessment of Uncontrolled Release For AP600 core damage sequences with adequate cavity flooding and debris coolability, no containment overpressure is expected. Even for sensitivity sequences that are assumed to lead to overpressurization by noncondensable pases or to basemat penetration, failure is predicted to occur much later than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the onset of core damage, Three additional sensitivity cases (CR, DRY, and CHF) were analyzed in this regard, even though these cases have negligible frequency of occurrence. They are presented at the bottom of Table 3-1.

Similarly, for SBWR core damage sequences with adequate cavity flooding and debris 1

coolability, no containment overpressure is expected. For sensitivity sequences in which the ex-

)

i vessel debris is assumed to be non-coolable, overpressure is predicted to ts : reached at about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, at which time overpressure protection from the suppression poca vapor space (i.e., a scrubbed release) could be utilized if necessary.

3.2 DOSE CRITERION The dose criterion limits the dose at 0.5 mile from the reactor from a physically-based source term to less than I rem for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the start of release of fission products into the containment.

Dose evaluations have been performed in the AP600 PRA. The Base Case sequence described in-the PRA closely approximates the URD physically based source term with 1009 noble gas release and 619 volatile fission product release. The containment leak rate is taken as the AP600 design leakage of 0.12 volume 9/ day. The containment leaks from the penetration area to the middle annulus between the primary and secondary containment shell which results in _

holdup of fission products and a reduction in offsite dose of about a factor of 20. The dose evaluation was performed using the MACCS code assuming that the release occurs at ground level and that 59 of the iodine release to containment is volatile and does not deposit. The median dose after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the start of release of fission products from the fuel is 0.07 rem-CEDE, well under the 1 rem level.

3-12 P

,v

l As stated in the AP600 PRA, variations on the Base Case and sensitivity sequences with isolated containment have fission product releases to the containment that are bounded by the URD physically-based source term. The release type associated with containment bypass and isolation failure sequences has dose greater than I rem, but as noted above and discussed in Section 2, such sequences are not considered credible as the passive plant has been designed to preclude such challenges. Offsite doses and frequencies of release for AP600 are presented in Table 3-1.

This table sumrnarizes the approach followed in the PRA in which four release types have been identified and quantified in terms of frequency of occurrence.

No evaluation of AP600 against the 5 rem thyroid limit specified in the URD emergency' planning design criteria and methodology was included in the PRA. However, on the basis of ALWR Program evaluations, the 5 rem thyroid limit can be met by AP600.' Also, experience indicates that given the 0.07 rem CEDE result, the thyroid dose will be under 5 rem.

Westinghouse has committed to provide the thyroid evaluation, and the ALWR Program will P

track this item.

Dose evaluations against the emergency planning dose criterion were not included in the SBWR SSAR. However, on the basis of ALWR Pmgram evaluations, the SBWR is capable of meeting both the i rem CEDE and the 5 rem thyroid dose for a physically-based source term. This is not unexpected since, as discussed above, the SBWR maintains containment load below appropriate ASME limits for credible accident sequences (i.e., low pressure core melts with containment intact) which should lead to low offsite doses. General Electric has committed to provide the dose evaluations for SBWR, and the ALWR Program will track this item.

3.3 SUPPORTING PRA REQUIREMENT The supporting PRA requirement is to demonstrate that the core damage frequency is less than 10-5 er year, that the cumulative frequency for sequences resulting in a dose at 0.5 mile greater p

than I rem for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is less than 10-6 er year, and that the prompt accident qualitative health p

objective of the NRC Safety Goal Policy is met with no credit for offsite evacuation prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A PRA was performed for the AP600 in accordance with Volume III, Chapter 1, Appendix.A of the URD. The total mean frequency of core damage was estimated to be 3.3 x 10-7 per year for internal events at power. For external events the core damage frequency for fires and intemal 3-13

f Goods was estimated to be less than 10-7 per year. Other external events are site specific, but on the basis of design characteristics and features provided to address such events the contribution of these events to core damage frequency is also expected to be negligible. For shutdown conditions the core damage frequency was estimated to be less than 10-7 per year. Thus, the j

total core damage frequency is expected to have significant margin to the 10-5 per year URD poal.

The AP600 complementary cumulative distribution function (CCDF) for offsite dose for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has been developed in the PRA. The cumulative frequency for sequences resulting in greater than 1 rem is approximately 3x10-8 per year, thus providing significant margin to the URD 10-6, I rem goal.

A PRA was also performed for the SBWR as required by Volume III, Chapter 1, Appendix A of -

the URD. The total mean frequency of core damage was estimated to be 1.8x10-7 er year for p

internal events at power. For external events the core damage frequency for fires and intemal Goods was estimated to be less than 10-6 per year. Other external events are site specific, but on the basis of design characteristics and features provided to address such events the contribution j

of these events to core damage frequency is also expected to be negligible. For shutdown conditions the core damage frequency was estimated to be less than 10-7 per year. Thus, similar to the AP600, the total core damage frequency for SBWR is expected to have significant margin I

to the 10-5 er year URD poal p

The SBWR CCDF for sequences resulting in greater than I rem over the course of the accident is approximately 2x 10-8 per year, thus providing signiGeant margin to the 10 6, I rem requirement.

i The SBWR SSAR indicates that the prompt accident quantitative health objectives of the NRC Safety Goal Policy are met with several orders of magnitude margin. No evaluation of AP600 against these objectives has been provided as yet. However, on the basis of ALWR Program evaluations, this objective can be met for AP600 with no credit for evacuation. Westinghouse has committed to demonstrate that the quantitative health objective is met, and the ~ ALWR Program will track this item.

i 3-14

3.4 CONCLUSION

S REGARDING PASSIVE PLANT CONFORSI ANCE TO ALWR REQUIRES 1ENTS Based on this preliminary assessment it is expected that the passive plant designs will be able to meet the emergency planning design criteria. Additional conformance assessment work may be

' appropriate as the design evolves and to assure that the containment systems being provided are well-engineered as described in Section 2.3.1. It is recognized that the URD, as well as the plant specific designs, have continued to evolve since the SSARs were issued. This design evolution is not expected to impact the conclusions of this assessment, and in fact may further enhance plant performance. In any case, the Plant Designers are responsible to demonstrate that their -

certified designs meet the emergency planning design criteria.

4 3-15

b Section

4.0 CONCLUSION

S The overall conclusion from the work performed to date on the technical aspects of ALWR emergency planning is that the likelihood and consequences of a severe accident for an ALWR are fundamentally different from that assumed in the technical basis for existing emergency planning requirements 15 years ago. Specific conclusions are as follows:

The updated emergency planning technical basis should be utilized for the ALWR. The primary reason for this is that the ALWR plant design capability, along with the greatly improved technical understanding of severe accident risk which has evolved over the last 15 years, result in significantly reduced ALWR s adiological risk.

A strong technical basis for updated emergency planning exists in the URD.

A set of deterministic criteria in the areas of severe accident containment performance and offsite dose, supplemented by PRA goals, have been developed for ALWR emergency planning and included in Volume III of the URD. For standard plant designs which demonstrate that these criteria are met, even in the extremely unlikely event of a severe accident the containment has been designed to maintain integrity and thus any radioactivity release will be very slow and small. A period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more exists before reaching offsite dose levels at which the U.S. EPA recommends that actions be taken to protect members of the public.

ALWR designs have excellent potential to meet the design criteria. A preliminary assessment of AP600 and SBWR conformance with the ALWR emergency planing design ciiteria has been performed and indicates that the designs will meet the criteria. The Plant Designers have committed to provide demonsustions as part of design certification that their respective designs meet the criteria.

q l

4-1 l

Section

5.0 REFERENCES

1.

" Advanced Light Water Reactor Utility Requirements Document," Electric Power Research Institute, Palo Alto, California. Volume I, March,1990; Volume II, Rev.

6, December,1993; Volume III, Rev. 6, December,1993.

2.

" Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants,"

NTREG-0396/ EPA 520/1-78-016, December 1978.

3.

" Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH 1400, October 1975.

4.

" Criteria for Preparation and Evaluation of Radiological E.mergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG 0654 / FEM A-REP-1, Rev.1, November 1980.

5.

"NRC Policy on Future Reactor Designs," NUREG-1070, July,1985.

6.

" Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"

NTREG-1150, December,1990.

7.

" Report to the Congress from the Presidential Commission on Catastrophic Nuclear Accidents," Volume 1, August 1990.

8.

N. C. Rasmussen,"Three Mile Island, Chernobyl; What Happened? What Did -

Not?" Presented at the OECD TMI-2 VIP Meeting, Boston, Massachusetts, October 20-22,1993.

9.

" Passive Plant Requirements Related to Shutdown," Prepared by the Advanced Reactor Severe Accident Program in Support of the Electric Power Research Institute, April,1992.

10.

" Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," U.S. Environmental Protection Agency, Office of Radiation Programs, Washington, D.C.,1991.

11.

" Additional TMI-Related Requirements," Code of Federal Regulations, Title 10, Part 50.34(f).

5-1

12.

D.E. Leaver, et al, "ALWR Utility Requirements Document Containment Performance Requirements /' Accepted for publication in Nuclear Engineering and Design, Reference Code NED 527.

13.

J.C. DeVine, et al,"The Passive ALWR Approach to Assuring Containment Integrity," SMIRT 12,1993 Post Conference Seminar: Containment of Nuclear Reactors, Karlsruhe, August 24,1993 14.

N. G. Trikouros, GPU Nuclear, personal communication to D.E. Leaver, Polestar, December,1993.

15.

G. Serviere, EDF-SEPTEN, letter to D.E. Leaver, Polestar, December 22,1993.

16.

" Review of Containment Shell and Penetration Leak Rate Data for Loading -

Beyond Design Basis," Prepared by the Advanced Reactor Severe Accident-Program in Support of the Electric Power Research Institute, December,1993.

17.

D. E. Leaver, et al., " Passive ALWR Source Term," DOE /ID-10321,~ U.S.

Department of Energy, Idaho Falls,ID, February,1991.

18.

H. J. C. Kouts et al., "Special Committee Review of the Nuclear Regulatory Commission's Severe Accident Risks Report (NUREG 1150)," NUREG 1420, August,1990.~

l' 19.

" Passive ALWR Containment Natural Aerosol Removal," Prepared by the Advanced Reactor Severe Accident Program in Support of the Electric Power Research Institute, Forwarded to NRC by EPRI letter dated April 30,1993.

20.

L. Soffer, et al.," Accident Source Terms for Light-Water Nuclear Power Plants,"

NUREG 1465, Draft Report for Comment, June 1992.

21.

" Standards for Protection Against Radiation," Code of Federal Regulations, Title 10, Revised Part 20 (20.1001 - 20.2401), mandatory as of January 1,1993.

22.

Nuclear Regulatory Commission,"10CFR Part 50 Safety Goals for the Operations-of Nuclear Power Plants; Policy Statement," Federal Register, Vol. 51, No.149, August 4,1986.

23.

Westinghouse Electric Corporation,"AP600 Standard Safety Analysis Report,"

DE-AC03-90SF18495, June 26,1992.

24.

General Electric Nuclear Energy,"SBWR Standard Safety Analysis Reports,"

25A5113, Rev. A, February,1993.

(

i 5-2

.~.

25.

" Meeting Summary," CSNI Specialists Meeting on Fuel-Coolant Interactions (University of California at Santa Barbara, January,1993), NEA Letter EN/S/532, March 1,1993.

26.

M. M. Pilch et al., "The Probability of Containment Failure by Direct Containment Heating in Zion," NUREG/CR 6075, June,1993.

27.

" Guidelines for Combining Natural and External Man-Made Hazards at Power Reactor Sites," ANSI /ANS Standard 2.12.

5-3

APPENDIX A ALWR Emergency Planning Criteria and Methodology and Updated Containment Performance Requirements (Reproduced from Reference 1) e

Section A.1 ALWR Emergency Planning Criteria and Methodology (Volume III, Chapter 1, new Section 2.6) 4

VOLUME Ill, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale kev.

2.5.3.4.7 Engineering As-built Walkdown Engineering As-built Waikdown 3

A detailed plant walkdown shall be performed after each An essential part of an SMA is the engineering walkdown to 3

ALWR plant is constructed to complete the SMA process. The look for potential undesirable seismic conditions in the com-selected primary and alternate success paths shall be walked pleted plant which cannot be identified during the design down using the guidance given in EPRI Report NP6041 to process. The SMA walkdown is performed to verify that the verify that the assumptions made in the SMA are valid. If any calculated rnargins have been achieved. During the equipment in the success paths is determined to have an ac.

walkdown, the review team will look for obvious deficiencies tual HCLPF less than the SME. it shall be evaluated to deter.

in the success path components selected for review and will mine thct the HCLPF will exceed the SSE by a suitable margin be cognizant of potential systems Interaction issues which or shall be strengthened. The walkdown process shall include cannot be identified during the design process. The designer review of construction drawings and documents.

should anticipate all concems that will be addressed during the walkdown.

2.6 CRITERIA AND METHODOLOGY FOR ALWR EMERGENCY CRITERIA AND METHODOLOGY FOR ALWR EMERGENCY 5

PLANNING PLANNING The Passive ALWR shall be designed to allow simplification Technical criteria and methodology are provided so as to 5

and standardization of emergency planning. The Plant Desig.

specify what a Plant Designer seeking approval of ALWR ner shall perform an evaluation of the plant design against two emergency planning for a particular plant design must ALWR emergency planning technical criteria prescribed below demonstrate during design certification. It is intended that for containment performance and site boundary dose. The these criteria and methodology form the technical basis for methodology which is specified for demonstrating the criteria any necessary regulatory action (e.g., a generic emergency shall be utilized in this evaluation.

planning rule in paraffel with Passive ALWR design certifica-tion rulemaking). The criteria and methodology are intended to be used in an integrated manner and the criteria should not be applied without utilizing the methodology specified in this section-Page 1.2-27

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS

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Paragraph No.

Requirement Rationale Rev.

2.S CRITERIA AND METHODOLOGY FOR ALWR EMERGENCY CRITERIA AND METHODOLOGY FOR ALWR EMERGENCY 5

PLANNING (CONTINUED)

PLANNING (CONTINUED)

The Plant Designer shall also perform a supplemental PRA The criteria and methodology for containment pedormance 5

evaluation in support of the evaluation against the two ALWR and dose evaluation are primar3y deterministic. The PRA emergency planning criteria.

evaluation is not a criterion itself but rather is intended to complement the two criteria. This is consistent with the NRC Severe Accident Policy which states that safety acceptability should be based on an approach wh.ch stresses determinis-tic engineering analysis, complemented by PRA.

The requirements in this section are generally unique to emer-5 gency planning although the containment performance criterion draws heaviy on containment performance require-ments in other locations of the Utilty Requirements Docu-ment. The requirements which are unique to emergency plan-ning apply only to plants which are seeking approval of ALWR emergency planning and not to other plants.

2.6.1 Containment Performance Criterion Containment Performance Criterion 5

For ALWR emergency planning, the plant shall be provided While ALWR accident prevention design features make the 5

with the capabDity to address severe accident containment possiblity of core damage extremely remote, specifying the challenges, including design features and characteristics to capabRity to address severe accident containment challen-preclude core damage sequences which could bypass contain-ges, including avoiding containment bypass and withstanding ment, and to withstand loads representing those associated loads which are expected to envelope best estimate pressure with core damage sequences. The methodology in Section and temperatures associated with severe accident conditions, 2.6.4 below shall be used to evaluate that capabHity.

provides confidence that the containment can withstand a severe accident.

ASME limits specified in Chapter 5, Section 6.6.2.2 should not Meeting ASME limits for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides low 5

be exceeded for a period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the leakage for the period corresponding to the site boundary start of release of fission products from the fuel.

dose criterion.

Page 1.2-28

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VOLUME Ill, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Rev.

2.6.1 Containment Performance Criterion (Continued)

Containment Performance Criterion (Continued) 5 Beyond approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, means for preventing uncon-Even if a core damage event should occur, the ALWR Pro-5 trolled fission product release from containment shall be gram considers that it is very likely that the ALWR contain-provided in accordance with Chapter 5, Section 6.6.2.5.

ment would be able to meet appropriate ASME limits for an indefinite time period, I e., no containment overpressure would occur. This is based on LWR accident management capabilities and the TMI-2 accident experience which suggest that it is likely that core damage events will be recovered in-vessel, and on ALWR reactor cavity design features (e.g.,

debris spreading area, flooding of debris) which are designed to quench the ex-vessel debris. Nevertheless, for defense-in-depth purposes, a requirement has been specified for no un-controlled release beyond approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide protection against long-term containment overpressure failure. Radioactive decay and removal of fission products in containment is such that a release at 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, or even ear-lier depending on the plant design, would result in no acute health effects at the site boundary. Thus, the aporoximately 24-hour period provides significant margin to thamime at which the acute heafth effects dose threshold cou!d be ex-ceeded.

Page 1.2-29

VOLUME 111, CHAPTER.1: OVERALL REQUIREMENTS Paragraoh No.

Requirement Rationale key.

2.6.2 Site Boundary Dose Criterion Site Boundary Dose Criterion 5

Dose at the site boundary shall be evaluated per the methodol-The 1 rem value is the Protective Action Guide (PAG) dose 5

ogy in Section 2.G.5 below and shall be shown not to exceed level which is specified by the Environmental Protection Agen-t rem for a period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the start of cy in a 1991 report as guidance for actions to protect the release of fission products from the fuel.

public in the early phase of a nuclear incident.

As noted in NUREG-1338, based on experience for non-5 radiological emergencies, ad hoc evacuations take from two to eight hours, including time to notify the public. Not ex-ceeding the PAG for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would provide significant margin for ALWR accident detection, notification, and ad hoc evacuation.

2.6.3 Supplemental PRA Evaluation Supplemental PRA Evaluation 5

A PRA evaluation shall be performed per the methodology in The requirement to perform the supplemental PRA evaluation 5

Section 2.6.6 below to demonstrate that the following goals and the associated goals are intended to demonstrate the in-are met:

tegrated effectiveness of the two emergency planning criteria (Sections 2.6.1 and 2.6.2 above). The supplemental PRA A core damage frequency s 10-5/yr; also serves as a tool for the Plant Designer for refining and optimizing the design. Finally, the supplemental PRA will pro-4 A cumulative frequency < 10 /yr for sequences resulting in vide confidence to the NRC in the overa!! safety of the plant greater than 1 rem over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the site boundary.

and in the margin to NRC guidelines on core damage fre-quency and large release. Given the guidance in the NRC Severe Accident Policy Statement, it is not intended that the PRA goals be made part of design certification or of any rufemaking on emergency planning.

In addition, it shall be demonstrated that ALWR designs are This requirement demonstrates that an acceptable level of 5

consistent with the prompt accident quantitative health objec-radiological risk to the public, as dafined by the prompt acci-tive of the NRC Safety Goal Policy with no credit for evacua-dent quantitative health objective of the NRC Safety Goal tion prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Policy, can be achieved with ad hoc evacuation, which as noted in Section 2.6.2, can be accomplished with sign! recant margin within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Page 1.2-30.

VOLUME Ill, CHAPTER 1: OVERALL REQUIREMENTS Paragrar>h No.

Requirement Rationale Rei 2.6A Methodology for Demonstrating Containment Performance Methodology for Demonstrating Containment Perfor-5 Criterion mance Criterion The Plant Designer shall demonstrate that the pressure and Chapter 5. Section 6 6.2.2, requires that the peak LOCA plus 5

temperature loads associated with core damage sequences hydrogen loads not exceed applicable ASME limits. The are no more limiting than the peak LOCA plus hydrogen loads loads associated with core damage sequences mu-st there-of Chapter 5. Section 6 6 2.2. For plant designs meeting the fore be no more limiting than the LOCA plus hydrogen loads.

requirements of Chapter 5. Section 6 6 2.1, the characteristics of the core damage sequences shall be as follows:

Containment is isolated and otherwise intact (i e., no bypass Consistent with Chapter S. Section 6 6 2 and the report. Pas-S has occurred);

sive ALWR Requirements to Prevent Containment Failure.

(DOE /tD-10291), December,1991, design characteristics and Reactor coolant system is depressurized to < 100 psig; features are to be provided which address severe accident challenges, including bypass and loads from core damage se-Ample water is in the reactor cavity / lower drywell prior to or quences. An exhaustive set of severe accident cha!!enges, immediately upon vessel penetration for cooling ex-vessel regardless of the probability of occurrence of the challenge, core debris; have been addressed based on systematic consideration of past PRAs, operating experience, severe accident research.

Passive containment heat removal is adequate, and unique design aspects of the ALWR. The conclusion from the technical work in support of this requirement is that BWR containments are inerted, and hydrogen control if core damage should occur, it wil be into an intact contain-system is functioning ment with the RCS at low pressure and with containment sys-tems functioning as designed.

Dest estimate severe accident methods sha!! be utilized in Best estimate methods are appropriate for the severe acci-5 evaluating the loads. Accepted industry computer codes such dent evaluation since the evaluation relates to matters as MAAP sha!! be applied beyond the design basis, l.e., the ALWR Safety Margin Basis, and since the ALWR plant features for addressing severe acci-dent cha!!enges significantly reduce the uncertainty in severe accident phenomena.

Page 1.2-31

VOLUME Ill, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Rev.

2.6.5 Methodology for Demonstrating Site Boundary Dose Methodology for Demonstrating Site Boundary Dose 5

Criterion Criterion The demonstration that the site boundary dose criterion is met The physically-based source term is based on release and 5

shall utRize a physically-based source term as defined in Chap-removal phenomena from actual core damage sequences ter 5. Section 2.4.1, including fission product release into an in-and is expected to envelope potential source terms from the tact containment. and fission product removal from the con.

probabilistica!!y significant sequences. The intact contalry tainment and the secondary building as applicable in the ment is based on ALWR containment performance require-design.

ments which have been specified such that severe accident challenges to containment are effectively precluded or can The methodology for the PAG dose evaluation shall consist of be accommodated, thus providing integrity of the contain-the following.

ment.

2.6.5.1 Approach Approach 5

A probabalstic dose (PD) method (e.g. CRAC2 or MACCS)

A PD inethod is chosen for consistency with the basis for ex-5 shail be used.

Isting emergency planning and the fact that PD methods have provision for the particulate component of the source term and thus are an appropriate method for calculating PAG comparison doses. The use of CRAC2, MACCS, or another similar code is consistent with current level 3 PRA evaluations and ALWR PRA Key Assumptions and Groundrules (KAG).

2.6.5.2 Meteorological Database Meteorological Database 5

The meteorological database shall be that provided in Annex This meteorological database is that provided la the PRA 5

B to Appendix A to Chapter 1 of the URD.

KAG. It is an actual site meteorological database for which the RG t.145 two-hour Exclusion Area Boundary X/O is es-timated to be greater than the X/O for 80 to 90 percent of U S. operating sites.

Page 1.2-32

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Kev.

2.6.5.3 Direction-Dependent vs. Direction-Independent Direction-Dependent vs. Direction-Independent 5

The dose calculation shall be direction independent.

The calculations supporting existing emergency planning are 5

direction-Independent i e., the frequency of exceeding given dose levels is provided independent of direction. The NRC safety goals use a direction-independent approach as well.

The use of a direction-independent approach is also consis-tent with the methods to be used in preparing the com-piementary cumulative distribution function (CCDF) for the ex-ceedance frequency of off-site doses at the site boundary re-quired by the PRA KAG.

2.6.5.4 Statistical Measure of Dose to be Compared to PAG Values Statistical Measure of Dose to be Compared to PAG Values 5

The dose to be compared to the PAG values for ALWR emer-Existing emergency planning used the PD method and.

5 gency planning shall be the median dose.

based on WASH-1400 source terms and frequencies, estab-lished that "most* core melt accidents would not exceed the PAG. There were two sources of variability in the supporting calculations which determined the meaning of "most"in this analysis: the source term itself (magnitude, timing, and eleva-tion / plume energy) and the meteorology. The ALWR physical-ly based source term already has significant margin com-pared to "most* core melt source terms since for "most-Pas-sive ALWR core melt accidents, the containment is expected to remain intact and the physically-based source term is bounding. Thus the comparison to the PAG value for ALWR emergency planning is based on the 50th percentile (i e.,

median) dose since "most* core melt accidents wotjid result in doses equal to or less than the median value calculated using the PD method involving weather as the only other source of variability.

Page 1.2-33

a VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Rev.

2.6.5.5 Whole Body Dose vs. Effective Dose Equivalent Whole Body Dose vs. Effective Dose Equivalent 5

The effective dose equivalent (EDE) shall be used.

The October 1991 revision to Manual of Protect /vs Action 5.

Guides and Protective Actions for Nuclear Incidents (PAG Manual) ca!!s for the use of EDE as the basis for determining off-site doses in relation to the 1 rem PAG. MACCS already 4

employs this concept, as does the current 10CFR20.

2.6.5.6 Comparison to Thyroid Dose PAG Comparison to Thyroid Dose PAG 5

The thyroid dose shall not exceed 5 rem.

Since the October 1991 revision of the PAG Manual con-5 tinues to consider the thyroid PAG, it 's appropriate to meet that guideline as a condition for ALWR emergency planning.

2.6.5.7 inclusion of Organic lodide in the PAG Calculation inclusion of Organic lodide in the PAG Calculation 5

In calculating doses for comparison with the PAG values to The I and Hi are quite reactive and are likely to undergo 5

justify ALWR emergency planning. the contribution from or-natural deposition as rapidly (or more rapidly) than the par-ganic iodide can be neglected.

ticulate. Given that pH is controlled as specified in the Utility Requirements Document, the dose contribution from organic iodide is very small (a few percent of thyroid dose) and thus can be omitted from the dose calculation.

2.6.5.8 Dose Commitment Dose Commitment 5

A dose commitment of 50 years shall be included.

In the October 1991 revision of the PAG Manual, it is required 5

that the EDE be a committed value or CEDE where the com-mitment is assumed to the *1tfetime". It is judged that a 50-year commitment is adequate on a generic basis to fulfill that requirement; it is also the duration used in the cunent 10CFR20.

This differs from the PRA as specified in the KAG where the 5

intent is to compare calculated doses to the 25 rem threshold for acute health effects (based on the current 25 rem whole txxty requirement in 10CFR100).

Page 1.2-34

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale llev.

2.6.5.9 Radionuclides to be included Radionuclides to be included 5

The radionuclides kjentified in Table 11-2 of the CRAC2 User's There are 54 radionuclides identified in this list. In MACCS 5

Guide (NUREG/CR-2326) shall be the minimum list of there are six adddional radionucindes: St-92, Y-92, Y-93, radionuclides included in the calculation of doses for the pur-Ba-139. La-141, and La-142. These are not critical for tfm pose of meeting the limits for ALWR emergency plana.ag.

PAG comparison calculation; the impact of the Sr, Y, Ba and La isotopes already included in the CRAC2 list is much greater, given their relative quantities, half-lives arxf dose con-version factors; therefore, the CRAC2 list is acceptable.

2.6.5.10 Dose Conversion Factors Dose Conversion Factors 5

Extemal dose conversion factors (plume and ground ex-Federal Guidance Report No.11 is the document referenced 5

posure) shall be based on Kocher, D.C., " Dose Rate Conver-by the October 1991 revision of the PAG Manual. However, slon Factors for External Exposure to Photons, and Electron in this guide, external dose conversion factors are provided Radiation from Radionuclides Occurring in Routine Releases only for noble gases. The extemal dose conversion factors from Nuclear Fuel Cycle Factitles " Health Phys., Volume 38, used in MACCS for NUREG-1150 calculations are referenced pp. 543-621 (1980). Inhalation dose conversion factors shall in NUREG/CR-4551 to the specified Health Physics article.

be based on Federal Guidance Report No.11

• Limiting These are judged to be acceptable for the use described Values of Radionuclide intake and Air Concentration and Dose herein. The inhalation dose conversion factors provided in Conversion Factors for Inhalation. Submersion and Ingestion.-

the guide are for a 50-year *1!fetime' commitment, consistent Office of Radiation Programs USEPA (1988).

with 2.6.5 8 above.

Page 1.2-35

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No Requirement Rationale Rev.

2.6.5.11 Plume Modeling Plume Modeling 5

The model used to treat dispersion in the calculation of doses The plume modeling in MACCS differs somewhat from that in 5

for the purpose of meeting the limits for ALWR emergency CRAC2. The differences have been resolved as follows:

planning shall be a straightline Gaussian plume. Plume center-To demonstrate that the PAGs wdl not be exceeded within line doses shall be reported The values of oy and o that are r

used to characterize the Gaussian plume expansion shall be the exclusion area boundary (EAB) radius, the peak based on Pasquil-Gifford curves If the analytical model used centerline value is the value that should be reported.

In the analysis employs a uniform approximation of the expan.

To obtain this value, the CRAC2 results ;nust be multiplied sion in the crosswind (y) direction (e.g, CRAC2), the final by a factor of 1.2. In addition, to compensate for the initially result shall be increased by an appropriate factor to provide more disperse plume in CRAC2 (which results from setting cente line doses. In the case of CRAC2 (which employs a 3 the initial oy equal to building width /3 instead of building oy " top hat" approximation of the cross-wind Gaussian distribu.

width /4 3), it is necessary to set the CRAC2 building width at tion), the factor shall be 1.2.

the input level to 70% of its actual value.

The initial oy shall be the building width divided by 4.3 if some 5

other factor is used to determine the initial oy (e g., a factor of 3 in CRAC2), and the buiding width specifiction shall be changed at the input level to compensate (e g., the building width for CRAC2 shall be input as 70% of its actual value).

i i

Page 1.2-36

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale key.

2.6.5.11 Plume ModeIIng (Continued)

Plume Modeling (Cont!nued) 5 In CHAC2. the expansion in the z direction (vertical) is 5

Tho correlation for dispersion in the vertical direction (z) shall be the form or = axb + c where x is the distance the plume controlled by an expression for or as a function of D

has traveled. The values for a b and c shall be the fixed plume travel, x. The expression has the fc m or = ax values in CRAC2. In the event a simpler form has been

+ c with the constants fixed in the coding. In MACCS, D

employed for calculational ease (e g, or = ax in MACCS),

a different correlation which does not use an additive the coefficients shall be set to provide the same value of o at constant ("c" term) has been employed, but only for the r

a site boundary of 0.5 mie and at a low popu!ation zone (LPZ) purpose of convenience. For specific radialintervals of radius of two maes as would be calculated using the fixed interest, values of a and b can be defined to give the values for a, b and c in CRAC2. Those values are as follows:

same values of X/O as CRAC2 at the two specific radial distances that define the interval. This is what has been Stability a

b done in this methodology specification. The 0.5-mie site boundary and 2-mile LPZ were chosen singly as A

2.47E-4 2.118 typical radial distances.

B 0.078 1.085 C

0.144 0 911 D

0.368 0.6764 E

0.2517 0.6720 F

0.184 06546 Page 1.2-37

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Rev.

2.6.5.11 Plume Modeling (Continued)

Plume Modeling (Continued) 5 For long release times (greater than a few minutes).

5 The time base for plume meander for long duration releases shall be the fixed value in CRAC2. three minutes.

plume meander becomes an important factor in deter-mining peak centerline doses. In CRAC2. the time base for plume meander was fixed at 3 minutes; in MACCS it is a user input with 10 minutes having been used in NUREG 1150 and appearing in the standard problem input tite. The data base supporting the modeling of plume meander includes averaging times (i e, the time base) of approximately 3 to 10 minutes. Since the im-portant parameter for plume meander is the ratio of release duration to the time base and since the release duration being used in the PAG assessment is to hours, per 2 6 5.14. duration to time base is better ap-proximated by using the low end of the averaging range (i e., the fixed CRAC2 value of 3 minutes) than the high end.

2.6.5.12 ' Release Height and Energy of Release Release Height and Energy el Release 5

The release height and energy of release assigned to the Current severe accident analysis practice is to use release 5

physically-based source term shall correspond to a cold, height and energy values that are consistent with the contain-ground 4evel release for the purpose of calculating the dose.

ment failure size / location or leak rate and associated ther-modynamic conditions. However, for the ALWR physically-based source term, containment is intact, releases are not credited through a stack, and best estimate meteorology is used. Thus a cold, ground level release is appropriate.

Page 1.2-38

VOLUME Ill, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No.

Requirement Rationale Re'v.

2.6.5.13 Duration of Exposure to Ground Contamination Duration of Exposure to Ground Contamination 5

The duration of exposure to ground contamination shall be 24 The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> perkxl ratovides margin for ALWR accident detec-5 hours from the start of release of fission products from the fuel.

tion, notification, ard ad hoc evacuation The 24-hour period is also consistent with the existing emergency planning basis.

2.6.5.14 Duration of Release and Number of Plume Segments Duration of Release and Number of Plume Segments 5

The release duration to be used in ca';,ulating doses for the The CRAC2 code has a limit on release duration of to hours 5

Passive ALWR physica!!y-based source term shall be 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and can employ only a single plume. The MACCS code wi!!

If a single plume segment is used or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if multiple accept a release duration greater than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and can plume segments are used.

employ multiple plumes (I e., different source terms in succes-sion), this capability being most useful when the character of the release to the environment abruptly changes in the course of an accklent. This is not the case for the Passive ALWR physically-based source term, where the difference in dose between a 10-hour release duration and a 24-hour release duration is only a few percent.

2.6.5.15 Shielding Factors Shielding Factors 5

Shielding factors shall be 0.75 for plume exposure and 0 33 The values given are those from NUREG 0396. Section F. no 5

for exposure to ground contamination.

immediate protective actions" and are consistent with the

" normal activity" requirement of the PRA KAG I

s Page 1.2-39

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS

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Paragraph No.

Requirement Rationale Rev.

2.6.5.16 Breathing Rate and Inhalation Protection factors Breathing Rate and Inhalation Protection Factors 5

1. 3 The breathing rate shall be 3 3 x 10 m /sec. For codes with The breathing rate identified in the October 1991 revision of 5

provision for an inhalation protection " actor, this value shall be the PAG Manual is the value specitied. In the MACCS code, set at 0.4. For codes without an inhalation protection factor, there is provision to reduce the inhalation dose by a factor to the breathing rate shall be reduced by a factor of 2.5.

account for differences between the plume concentration and the concentration actually being breathed. NUREG!CR-4551 (one of the supporting documents for NUREG-1150) sug-gests an annual average value of 0.4 for normal activity (0 2 for active sheltering). The use of a " normal activity" inhalation protection factor is consistent with the requirements of the PRA KAG.

2.6.5.17 Dry Deposition Velocity Dry Deposition Velocity 5

The dry deposition velocity shall be 1.0 cm/sec for iodine and These values are those of the October 1991 revision of the 5

0.1 cm/sec for other particulates.

PAG Manual. Cunent severe accident analysis practice is to use values of 1.0 cm/sec (NUREG-0396tCRAC2) to 0.3 cm/sec (NUREG-1150/MACCS); the PRA KAG does not estab-lish a requirement for dry deposition velocity.

Page 1.240

VOLUME 111, CHAPTER 1: OVERALL REQUIREMENTS Paragraph No Requirement Rationale Rev. -

2.6,6 Methodology for Performing Supplemental PRA Methodology for Performing Supplemental PRA 5

The suppfemental PRA shaQ be performed in accordance with The KAG is the ALWR methodology for PRA evaluations. The 5

the Volume til, Chapter 1. Appendix A. PRA Key Assumptions KAG specifies that the PRA address intemal events plus exter-and GroundnAes (KAG) with the exception that the off-site nal events with the exception of seismic risk which is to be dose exceedance limit is 1 rem, per Section 2.6.3 above.

addressed by the seismic margin approach per Chapter 1, Section 2.5.3.4, of the URD.

The required demonstration on the NRC Safety Goal Policy The numbers specified for risk comparisons are based upon 5

shall use the following methoddogy:

recent data from the National Safety Counca (Accident Facts, National Safety Counca,1988). The quantitatke objective for The ALWR reference site parameters in Annex B to the KAG latent cancer risks, which is also part of the NRC Safety Goal shall be used.

Policy, is not included in this required demonstration of Safety Goal compliance because, as noted in NUREG-1150, No evacuation shall be assumed prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

emergency response in close-in regions does not contribute Subsequent to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the evacuation parameters substantially to differences in latent cancer risk. It is ex-of the KAG, Annex B, shall be used.

pected, however that ALWRs would have no difficulty in meeting the latent cancer risk quantitatNe otiective.

To demonstrate the NRC Safety Goal Policy quantitatNe objective for risk to an average individual (less than 0.1%

of the risk from all other accidents). ALWR accident risk shall be less than 4x10# per person per year.

Page 1.2-41 l

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h P

t e

Section A.2 L

i Updated Containment Performance Requirements (Volume III, Chapter 5, revised Section 6.6.2) b k

4 i

.i; 6

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VOLUME 111, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No 9equirement Rationale R'ev 6.6.2 Containment Performance Containment Performance 0

The ALWR containment performance requirement shall consist The elements below comprise a deterministic approach to 5

of a number of elements as speertied below. The in:tial ele-containment performance. The deterministic approach is ment shall include a matrix of plant design characteristics and complemented by the PRA requirements, including meeting features to address a comprehensive set of containment chal.

ALWR PRA goals. T* is deterministic approach, comple-lenges from severe accidents This matrix approach, together mented by PRA, is consistent with NRC Severe Accident with the other elements of containment performance shall pro.

Policy Statement guidance and provides the set of contain-vide high assurance of containment integrity and low off site ment performance requirements that are considered neces-dose in the event of a severe accident.

sary to address severe accidents The combined set of deter-ministic and PRA requirements satisfies the Commission response to SECY-90016 for a deterministic alternative which provides at least comparable mitigation capability to the conditional containment failure probabaity (CCF) of 0.1 but does not discourage improvements in core danuge prevention.

6.6.2.1 Plant Features to Address Containment Chattenges Plant Features to Address Containment Challenges 5

The plant shall include design characteristics and features to Design characteristics and features to address a comprehen-5 address a comprehensive set of severe accident chaftenges to sive set of severe accident containment challenges are neces-the containment. Design characteristics and features stu!Iin-sary to provide severe accident protection for the ALWR con-ciude:

sistent with the NRC Severe Accident Policy, ALWR safety policy, and to meet the ALWR requirements. A complete set of design characteristics and features and the adequacy of these characteristics and features is documented in the repor1. Passive ALWR Requirements to Prevent Containment failure (DOE /tD-10291). December 1991. In the report, an ex-hausthre set of severe accident challenges, regardless of Page 5 6-38

VOLUME lit, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rationale Rev 6.6.2.1 Plant Features to Address Containment Challenges Plant Features to Address Containment Challenges 5

(Continued)

(Continued) probability, have been addressed based on systematic con-5 sideration of past PRAs, operating experience, severe acci-dent research, and unique design aspects of the ALWR. The report concludes that the severe accident cha!!enges have been effectively precluded or can be accomrnodated by the ALWR design characteristics and features specified in the Re-quirements Document.

Features to provide reliable shutdown of the reactor by ReliatWe reactivity control, through rod insenion and the 5

rod insertion, e g., Chapter 4. Section 5.3 (BWR) and capability to accommodate failure to scram in the form Chapter 4. Section 6 2 (PWR) as well as dNerse reactivity of diverse means of reactivity insertion, limits the challen-control capability in the form of SLC, Section 4 5 (BWR) ges associated with ATWS.

and PSIS, Section 5 2 (PWR).

Features to reliably depressurize the RCS, e g. Sections A reliable depressurization system minimizes the prob-5 4.4 (BWR) and 5.4 (PWR).

ability of high pressure core melts with subsequent potential for direct containment heating. Cavity con-figuration also limits the magnitude of containment pres-sure rise.

Features to limit the generation of non-condensible gases Containment integrity could be chattenged in the long 5

as a result of corium-concrete interaction, e g., Section term as a result of pressure buildup from production of 6 6.3.

non-condensible gases following cortum-concrete inter-action. Preventing or limiting this event enhances con-tainment performance.

Features that provide passive containment cooling for Long term containment cooling is required to maintain 5

l l

decay heat removal, e.g, Sections 4.3 (BWR) and 8 3 containment pressure within design limits.

(PWR).

I l

Page 5 6-39 l

^

VOLUME lil, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No Requirement Rationale Revl 6.6.2.1 Plant Features to Address Containment Challenges Plant Feature: a iddress Containment Challenges 5

(Continued)

(Continued)

Features to handle the pressure and temperature resuit-Features that control combustion and prevent detona-5 ing from generation of combustible gases, e.g, Section tion of hydrogen eliminate this threat to containment in-6 5.

tegrity following a severe accident.

Features to assure containment integrity including isola-Challenges to containment Integrity which result from 5-tion and precluding steam generator tube rupture and failures which occur independent of or coincident with other containment bypass scenarios, e g, Chapter 3 Sec-core damage (e.g, containment bypass events) must be tion 2, and Chapter 5, Sections 4.3,6 2, and 7.2, for the avoided.

BWR and Chapter 3. Sections 2 and 4, and Chapter 5, Sections 5.3 and 6.2, for the PWR.

6.6.2.2 Containment Performance Structural Evaluation Containment Performance Structural Evaluation 0

The PfarA Designer shall demonstrate that the containment sys-The ASME Section til Code referenced structurat integrity 4

tem pressure boundary, when sub[ected to the pressure and criteria satisfy the intended minimum requirements of temperature loads from LOCA plus hydrogen described 10CFR50.34(f)(3)(v). Also, any gross distortions ard sub-below, combined with the appropriate dead loads, meets the sequent large strains in pressure boundary material due to following ASME Code, Section Ill criteria:

potential shell buckling modes are precluded. The LDB re-quirements (Section 2.4 2) are expected to be limiting for in-erled containments while the SMB requirements are expected to be limiting for containments which are not inerted.

For Class MC free standing steel vessels and for the steel 5

portions of Class CC reinforced concrete vessels which are not backed up by concrete, the following require-ments shall apply-Paragraph NE-3221, Service Level C Limits on 0

strws intensity values.

Page 5.6-40

4 VOLUME Ill, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No Requirement Rationale Rev.

6.6.2.2 Containment Performance Structural Evaluation Containment Performance Structural Evaluation 0

(Continued)

(Continued)

Fo. regions of ellipsoktal or totispherical shell sur-Compressive stress in ellipsoidal or torispherical shell 5.

faces of containment, the allowatie compressive heads due to internal pressure loading is a localized stress due to internal pressure shall not exceed 60 stress field which does not represent a challenge to percent of the value of critical buckling stress deter-overall containment stab 8ity; thus is lower factor of mined by one of the methods given in ASME Sub-safety against buckling than otherwise permitted by paragraph 3222.1(a)

Subparagraph 3222.2 is appropriate in these regions.

The value of 60 percent of the critical buckling stress results in a safety factor of 1.67, which is consistent with the requirements of Code Case N-284 for local buckling For the steelliner portions of Class CC vessels which are 0

backed by concrete, the factored load limits on liner strains estatAished in Subarticle CC4720 shall apply.

For those portions of other ASME Code class com-O ponents which also constitute a portion of the contain-ment systems pressure boundary, the corresponding ASME Section til Service Level C Limits shall apply.

Page 5.641

VOLUME lil, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rationale ilev 6.6.2.2.1 Inerted Pressure Suppression Containments inerted Pressure Suppression Containments

~5 The analysis of LOCA plus hydrogen loads shall assume:

The assumptions maximize the pressure and temperature 5

loads in the containment in the performance of the Pool tsmperature equal to the peak temperature 10CFR50 34(f)(3)(v) analysis.

associated with the DBA LOCA within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the accident initiation.

All nitrogen in the drywell is located in the wetwell altspace.

The total hydrogen equivalent to 100% active fuel cladding metal water reaction is located in the wetwell airspace.

6.6.2.2.2 Non-inerted Containments Non-Inerted Containments 5

The analysis of LOCA plus hydrogen loads shall assume:

The Licensing Design Basis analysis required by 5

10CFR50.34(f)(3)(v) would credit a hydrogen control system Peak pressure associated with the DBA LOCA; as hydrogen is generated. The Safety Margin Basis analysis requirement contained in this section postulates the peak Accumulation of hydrogen associated with 75% active DBA pressure and a realistic upper bound to total hydrogen fuel cladding metal wates reaction; concentration, i e., that associated with 75% active clad oxida-tion. before crediting a hydrogen control system or Ignition Adiabatic isochoric complete combustion of this sources. This yields a higher peak pressure than that re-accumulated quantity of hydrogen.

quired by 10CFR50.34(f)(3)(v).

If containment is found to be steam inerted at the peak DBA Buming is assumed to occur at the highest potential contain-pressure, then combustion shall be assumed to occur at the ment pressure if inerting In!tla!!y precludes combustion.

time steam condensation reduces the mole fraction of steam to combustible levels (~ 50% mole fraction steam).

Page 5.6-42

~

VOI.UME lil, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rationale Rev.

6.6.2 3 Severe Accident Sequence Selection for Reporting Severe Accident Sequence Selection f+; Reporting 0

Containment Response Containment Response The Plant Designer shall report containment performance The primary means of addressing severe accktent contain-5 during severe accidents. Analysis of severe accident sequen-ment cha!Ienges is the deterministic matrix of design charac-ces shall be performed to confirm that the containment teristics and features of Section 6 6.2.1 and the deterministic provides substantial margin with respect to severe accident analyses of Section 6 6.2.2. This deterministic approach ad-challenges. Accident sequences from the PRA shall be dresses an exhaustive list of containment challenges, regard-selected for analysis of containment performance. PRA se-less of probability. The probabilistic requirement of Section-quences shall be grouped into functional sequence types for 6.6.2.3 complements the deterministic approach as required the purpose of determining the mean total frequency of all ac.

In the NRC Severe Accident Policy. The difficulty of assign-cident sequences with approximately the same type of chal.

Ing accurate numerical estimates notwithstanding, use of lenge. The sequence types shall be those resulting from the PRA in this manner provides valuable design insights and failure of any one of the following functions:

added confidence that containment margin exists for severe accidents and that important risk contributors have been ad-Reactivity insertion; dressed.

RCS depressurization; This set of functions is considered necessary to assure con-tainment integrity based on the report Passive ALWR Severe Core or core debris coolant inventory control; Accident Containment Performance Requirements, January 1992. This report concludes that the only potentially sig-Containment pressure / temperature control; nificant severe accident challenges to a standard ALWTl plant design which implements the provisions in the Requirements Combustible gas control; Document are those associated with core damage events that occur into an intact containment with the RCS at low Containment isolation and containment bypass control; pressure with containment systems functioning as designed.

Other functions, the faRure of which could lead to containment challenge.

Page 5.643

VOLUME Ill, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No Requirement Rationale RN 6.6.2 3 Severe Accident Sequence Selection for Reporting Severe Accident Sequence Selection for Reporting 0

Containment Response (Continued)

Containment Response (Continued)

Functional sequence types with mean frequency greater than The approximately 10'#/yr threshold for furetional sequence 5

approximately 10'#/yr shall be analyzed for containment types to be analyzed for containment response is consistent response.

with the NUREG-142010/yr limit for insignificant risks and is consistent with Standard Review Plan guidance to evaluate potential accidents from hazards in the plant vicinity which ex-ceed approximately 10'# yr. Also. NUREG '150 uses a cutoff

/

of 10 /yr for accident progression analysis NUREG-1338 stated that any se ence appearing to have a frequency down to about 10'p/yr wR1 be examined from the standpoint of residual risk Finally, consideration of functional sequence types greater than approximately 10-# yr provides assurance

/

that the cumulative effects of such sequence types will not ex-4 ceed the 10 /yr probabRity goal for off-s!!o consequences.

Functional sequence types with frequency less than 10-# per The purpose of this requirement is to assure tiet there is un-5 year shall be reported for discussion:

derstanding of those features designed to preclude contain-ment failure resulting from a severe accident. It is also ex-Identifying the design features and operating characteristics pected to show that those phenomena which could lead to credited to reach this low frequency; exceeding the capacity of containment early in a postulated severe accident event are a small fraction of the ALWR PRA Singling out the frequency of those sequence types which goals for core damage frequency and consequences.

may result in early containment failure.

If the loads resulting from the analyzed severe accident se-The loads resulting from any analyzed functional sequence quence types are enveloped by the conditions determined for types shall be no more limiting than the peak LOCA plus LOCA plus hydrogen in accordance with Section 6.6.2.2, the i

hydrogen loads of Section 6.6 2.2 for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> comparison of these severe accident loads may be rnade after the start of fission product release from the fuel.

directly with the LOCA plus hydrogen loads. In the event the loads exceed those determined in accordance with Section 6 2.2 2, it is expected the Plant Designer will be able to Page 5.6-44

VOLUME 111, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rationale Rev.

6.6.2.3 Severe Accident Sequence Selection for Reporting Severe Accident Sequence Selection for Reporting 0

Containment Response (Continued)

Containment Response (Continued) demonstrate that the containment still meets the functional 5

criteria for Service Level C or Unity Factored Load as per-mitted by 10CFR50.34(f)(3)(v) and provide confidence that the structural integrity and leak tightness of the passive plant containment will be maintained following a severe accident.

Should any functional sequence type selected for analysis 5

result in loads which exceed the functional criteria for Service Level C or Unity Load permitted by 10CFR50.34(f)(3)(v) or result in containment bypass, the Plant Designer should iden-tify the reasons for the high loads or the bypass and explain why the accident sequence frequencies cannot im further reduced, and provide recommendations for an attemate basis on which confirmation of acceptable containment per-formance can be justified.

t Page 5.645

VOLUME lil, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rationale Rev.

6.6.2.4 Containment Ultimate Capacity Analysis Containment Ultimate Capacity Analysis 0

The Plant Designer shall perform an analysis to determine the An analysis of containment ultimate capacity is required by 0

ultimate structural capability of the containment. For steel con-Standard Review Plans 3.8.1 and 3.8 2, including the deter-tainments, the ultimate capacity shalt be defined as the pres-mination of pressure retaining capacity of localized areas.

sure and temperature loadings which correspond to the col-The failure analysis criteria included here are identical to or lapse load defined by the method detailed in paragraph 11 more conserva1No than thoso developed during NRC/IDCOR 1430 of the ASME Code, Section lit. Appendix II. For con-Issue resolution (see ARSAP Technical Task 2.3 seport) or are crete containments, the ultimate structural capacity shall be more realistica!!y based on recent experimental tests for con-defined as the pressure and temperature Icading which crete containments by Sandia National Laboratories. These produces liner plate strains equal to the liner strain limits of tests have indicated that concrete containment capabil;1y the ASME Code Section Ill, Subarticle CC-3720 for the Fac.

may be limited by leakage resulting from liner plate tears tored Load Category. The analysis shall consider the penetra.

EPRI report NP-6261 describes computer modeling techni-tions and their interaction with the containment, the shield ques used to predict the failure mode of the scale model con-building, and other structures intemal or extemal to the con.

crete containment tested by Sandia. Interaction of the con-tainment, which might cause localized failure prior to the limit tainment penetrations w!!h the shleid buiding or other struc-load for the overail pressure boundary. Results from testing of tures may produce leakage paths.

prototype detaRs or models of prototype detais may be used to augment such analyses. The failure mode associated with the ultimate structural capability shall be identified.

Page 5.6-46

VOLUME lit, CHAPTER 5: ENGINEERED SAFETY SYSTEMS Paragraph No.

Requirement Rarianale Rev.

6.6.2.5 Long-term Containment Overpressure Protection Long-term Containment Overpressure Protection 5

Protection of the containment for overpressurization beyond Containment overpressure protection provides additional 5

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sha!I be provided. Overpressure protection beyond defense-In41epth to protect the conta!nment from long-term 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may be provided simply by the size and strength of catastrophic faBure. The analysis shall credit design features the containment by demonstrating that the ASME limits for containment heat removal and debris cooling on the basis specified in Chapter 5. Section 6 6 2 2, are not exceeded for of Passive ALWR requirements directed at decay heat approximately two to three days after the beginning of the ac-removal ard providing water to the debris. The analysis cident.

should utaize best estimate analysis methodologies locluding realistic assumptions.

On the order of two to three days is judged to be adequate time for actions by the plant staff to bring the accident under control.

6.8.3 Cavity / Pedestal-Drywell Configuration Cavity / Pedestal-Drywell Configuration 0

6.6.3.1 Retention of Core Debris Retention of Core Debris 0

The reactor cavity / pedestal drywell shall be evaluated to con-The specified evaluation wHI confirm that direct containment 0

firm that quantitles of core debris sufficient to jeopardize con-heating is not an issue for passhre ADVR designs, based tainment integrity wil not be transported from the primarHy on the assured provisions for RCS depressurization, cavity /drywell after RPV failure and then either mix with the but also considering the specific proposed cavity / pedestal containment atmosphere while in a finely particulated form or drywe41 geometry. The PRA will define the extent to which in-establish direct contact with the containment boundary. For complete depressurization wBI be considered for a specific passive ALWRs, the evaluation shall address low-pressure design _

(nearly complete depressurization) conditions prior to vessel failure unless a higher pressure sequence is identified as risk-significant in the PRA for a specific passive ALWR design.

Page 5.6-47

i APPENDIX B Summary of ALWR Requirements to Address Severe Accident Containment Challenges (Reproduced from Reference 12) b i

1

)

i lable 4

,~

St#9tARY Of REQUIREHLHIS 10 AI) DRESS CONIAINNINI CllAlLLNGES TilAT ARE INUFPFNDENI 0F OR C0lHCIDENI Will! 00ltE UAHACE

$1 Y l'A551VI AlW REQUIRf Ht NI5 AtitCit0 SAFilf PtANI At W CllAl t [NG(

FliMCijM ITPE BASISt t lHit POIE Nil Al inR tilAtlIN6t "

ALC0HH0llAll (HAl l t Nht 7 1.

Cent a lnment Isolat ton Isolat ton PWR/PWR 2

A Reduced fluid line penettations P Passive Res tilis.il liest Rece va l minialres soir Isolation prowlsions and leakage s ate test ing damage r isk given isolat 'on f ailure (with per standards.

RHR on line even wittmut DC pn=er).

valves capable of closus e with gam.ihh I low and f ull contalrment pressure.

Conteol roan position traficat ton Int au t tana t Ic and remote manual valves.

A Manual valve configurat ion permit s ha 6 log t

only 6n closed posit ion.

A Closed systems penetrat ing tont alsment evaluated for ex-vessel severe accidents f all closed or DC powered isolat ion valves A Capability f or periodic gr oss ctek of containment integrity.

2 Intet f acing System (OC A Eypass PVR/BVR 2

A Redia.ed interf aces between the Reactot t oo lant Pressuie Relief System (RC5) and low pressure systems A Design piessure snth that t ull Mrs piews..e A lingh to low pressure Interf aces provided with is below ruptuie pr essus e ami eu, leat s =ll t isolation valves leak testinq capability, occur which cit eed 405 ndeup upacit y isolation valve position laidnator in control room, and high pressure alaim.

Interlocks prevent isolation valve open6ng when RCS pressure exceeds RSDC system design pressure (PVR).

A RSDC desIgne i tor f ull ecattor pi er.no e (t<WR )

Double isolat lon.

f lie acceptability of ALV5t requirements to adsh ess contalevnent olialletiges was based on tie f ollowing o iles ia-1.

Current LVR resistance to challenge acceptable for ALVR.

2.

Suf f icient AL VR design f eatures added to increase resist ante to t.hallenge by reduclog the sevee lt y aseillor ensuring contalissent.

• Passive plant design f ea tut es wh k h e=< eeil r eigulr ement s f ue curient lWs ate bient it led with A (simmm to all AlWs) or P (passive A!Ws smly)

Iable 4 (continued)

SUHHARY Of HEQUIREMlHIS 10 AI) DRESS CONTAlHMINI CllAllENGES IllAT AllE INDEl'ENDENT OF 011 COINCIDENT Willi LD11E DAMAGE DtY PA551Vt Al W Pt 0 alWt Mt NI5 Ali f Cif D SAFElf PLANI At W

[t14t l [ NrA fitNCil0ft liPE BA51F LlHI! P01[NilAl f M fitAl i t IEl

• Art MH0il4 t i 184l l I IH>t 1*

1.

Blmnlown forces Con t a inn.en t PW/ew I

Design avul ISI en accaritariti-wit h ASMt BPV Design contaniw nt fus.L eub le code.1 Pressure Code guillut ine ta cek i>t largest pipe.

Control teak Before Break.

4.

Pipe Whip drid bypass PWH/BWR I

Desagte atut 151 in accordance with ASHL UPV Protect ion f r uin jet / pipe whip =lwi e leak Jet Impingement Code.

before break ts not demonstrated.

leak Before Break.

lise of only pr oven mater is l> 4eul los.: itat sosi processes.

lise of EPRI water chemist r y gui le t i. cs

• ,. Steam Generatur Bypass PW t

improved water chemistr y P Oper atur att susas tan te initiat e leak s sta pi si.e Tube Rupture Proven matertals.

to ADS actuat nun f or de,sgn leas ts le. k A Mechan 6 cal design of tubes, teihe suppusts. and P Autisnatic Depressur trat ion Systein I ADS) tube sheets reduce likelihood of SGIR.

operat ion ten minates tid >ea-leak a*je A leproved design features f acilitate 56 aut una t ica lly.

c leaning and replatenent.

P Passive lilR p his aiktit sona l f e.st nics pitwnt secessulary side reitet f ollowing 5618t to AlW5 Reactivity BW" 2

A Diver se Neactor Protect nun Systen INP$l.

St anitby i siguid finite ol (Li t 1 Control A Diverse means of rod insertion A Checkert.oard pattern of scras gioup ei l.

manistres gioup worth.

FW 2

A Diverse RPS (or capabilit y tu e ide uast AIW5).

Bosated Safety i n jet t ion (til A llegat Ive modera t or t empei a t ut e t uet I u sent over ent tse f uel ocle in. proves Alus response.

f Ibe acceptability of AlWR requirements to address containment challeinje:S was based on the f ollowing or iles ia.

l.

Current LWR resistance to challenge acceptable for AtWR.

2.

Suf t tclent AtWR design f eatures added to increase s estst ante to challenge by reducing the sever it y auit/ue ensuring contaisiment

• Pa ss ive p losit des igii t eatieres =1 si.fi e...eed + eriss esii-nt s f oi tier e erit t W s are ident if ted 3,itle A (cuiseiun tu all At Ws) or P (passive At w> un ty) 15

lable 4..(continued)

SIMiARY Of REQillRLMENIS 10 AllilllLSS CONIAlHHINI DIAILLNGES lilAT ARE IHil[l'EHilfMI Of OR ColHCillfN1 Willi utile ilAMAGE fl V PASSIVt At W R10411REMf NIS AII E C l[D SAFilf FtANI AlVR (ItAt I ( NGE f f!NC110N lYPE BASIS #

IIMIT Pol [NIl Al I OR filAl t l Nbl

• Alf0Mt W All (HAltlNhtS*

7 Simpr ession Pool thpass fon t a irwnent BLR 2

Vacuese Brean ce s : potenttil In nis

.n.i inmt eil ADS use of 5tJVs h h h diu ttit ye to Psessure for. position ludicat ion, minima l led age suppress ton pool and thus ensui e v ie.o Control P Ho high energy lines in welwell ais spm e suppr esslote despite le ak age P Passive Rt1R (inc luding FU S) 8 Catasteophic RPV falluee Int er na l I'WR/BWR

?

A RI 1 10 F; initial RI NOT NDI 1 Con t a isnaent core belt line; low f luence at veur t w-i ll l oad ing A No welds in belt line region A Relief valves prevent oves pe essin e, l..n ke.t up by depressur trat ion system and low he.ul inlection.

Design in at.cordante with ASHI u n to Design features to avoid seltel valw opening f or e=pected plant transients.

9 Int ennal Vacuinn Cont a tionent PWR/BWR I

Vacuinn Br eaken s Pressure Design for entein.it piesvue in.ed.

Control 10, internal (Plant)

(sternal FWR/BWR 2

turbine overspeed protect ion lut b ane or ientat nun avoids missile <anit.u t Missiles Cont a inisent A lenproved turbine integrity /one piec e sonors.

with containment.

t oading Missile protection for any saf ety telete 4 components in missile path (5RP J.51 3)

Il tornado and lovnada External PWR/0WR 2

Conf ormance with ANSI 2 I? and AttSi '31 5 P Passive cos e c ooling syst nns ins.stant wit hin Missiles Cont a irwnent containment l oadinq

1. lie acceptabilit y of A!WR s eguir ement s to adde ess constaisimerit s ha lleeiges was based ori the lo t hiwisig i s at es i.s-.

I Current IWR resistance to challenge acceptable for AlW.

2, Sullicient At WR design teatures added to increase vesistant:e to (fullenge by t educing the seve ity and/u: ensur ing cont a lruisevit.

• Passive plant design f ealue es which e=ceed requirewnts f or t ue rent t WRs are ident ified with A (uwennn to all AlWRs) or P (passive AlWRs only).

er

Table 4 (continued)

SUMMARY

Of HEQtilREHlHIS 10 A110RESS CONEAINHINI CilAllLNGES TliAT A!!E IH11El'ENDENT OF DR COINCIDENT Willi CORE DAMAGF t t Y l' ASS {Vi Al W Mt00lRIMINth All t[It D SAFElf PiANI At W (IIAll f h6{

[Mjg

]tPE Mht ilMll P0l[NIl At F OR (Hall ( t4:,l.

• Af ( OHMGliait (llAt i t Ni t S

• i 12 Han-Made 5 tt e Enternal PW /aw 2

Conf orenance with ANSI 2 12 P Passive tuv e t ooling systems Iwat eil wit hin Prualmity Hazards Cont a irmnent contatteent loading 13 5etsmic laternal PWR /0W 2

5 tt ing requirements esc lude the si=o t A 55t. at 0 39 Cont a iniment vulnerable sites.

A Evaluat icn at > sst with PH A os ma i g in s loading assess.nent as part of alesign piocess A Address vulnerobilit les t run past esper li nces, e 9, pr ov ide e tmt.on 1.n.w it i Ibe acceptability of AlWR requirennts to atkiress cont ainerient challenges was based on the f ollowing Li tt er ia.

l.

Current LWR resistance to challenge etceptable for AIWR.

2.

Suf f icient Alb11 design feate res added to increase s es tstaru e tu s hallenge by reducing tiie sevee st y av 1/or ensuring tuntatrimesit

• Passive g= latit ales egri f eat ures wlittle estec41 s e, tis t r e-seiient s t us Less s eint 4 Ws es e islent if led with A (temiainen l es a ll At Ws) or P (gia>> twe Al Ws ana ly) ti

lable 4a SUMARY Of RLQUlHLHENIS 10 ADDRESS CONIAINH[NI CHALLINGES itESUL11NG IROM CORE DAMAGE ti T l'A551VE Al W Rf 0tilRf Mi rals Af f f Cl!D S AF E I'T FLANT AL W CilAt l (Nr,E TWffil0N ItPE BA$157 t IHit P0l[ Nil Al FOR CitAll t N6t

• A( COMM00411 [14Al t l N6[ $*

14. High Pressure Melt Reactor Bbt 2

P Diverse depressurtration systerns.

Suppress tun gnol cools tu.it ed gases E Ject ion (Iff'Ml l Pressure P Passive PilR can aid deps essue trat son Inerted conta bafnt (no cmbustion f. eat Control addition).

Pb'R 2

P Unverse depressurtration systrims.

A f avit y conf igui at tun to l im i t tianspoit of P Passive RHR can aid depressue trat ton fragmented cone debeis.

15 flydeogen Generat ton to Combusttble BW I

Ineited.

A ivaluitton sequised it liu,i l ilet ona t inn n Datontable t imits Gas Control possible.

generat ton with d.esign f eatus es such A i v a luat ton reyis es e.t it ha.el sletunatsim e.

l'W 2

A t la s t i[l as AD and cawtty flooiling passtble.

A flydrogen control system (e.g., sem saf et y related igniters) designeil to keep hydrogen concentrat ion below 10% f or 100% aitive clad equivalent r action.

A Containment site prevents global detonable ti l concentration (< 13%) for generatton up to 75% act ive clad equivalent rear.tlen A Design ensures convect swe mim ing and minimites D01-pr one geomet r y I IIE Ncepia'bliliy of ALW requirements to address contalrment thallenges was based on the f ullowing u itei sa:

1.

Current tWR resistance to challenge acceptable for ALVR.

2.

Suf flctent ALVR design features added to increase resistance to challenge by reducing the severit y and/or ensuring containment.

• Passive plant design feetures which exceed requirements for current (WRs are identified with A (tonmn to all AlWs) or P (passive AlWs only) u

.m m

~

liable 4a (continued)

SUMMARY

Of RII)UIRIMINIS 10 ADDRESS CONTAINMLHI CllAllLNGES 11f5111 TING ilt0M CORE DAHAGE kl y pASMVt AlW RtQtilRt Hf N{5 Af flCit 0 5Af f f f PtANI Al Wl3 f

J I!NCI10N liPf

[AM1f I[H1i POIiNil Al ) OH t li Al i I to.t

• At t OMMoD A IL t liti i t m>l 5*

yi [ Nr,{

ltpleogen Dellegr at son Cie. bent ihle BVH I

loes i eil A Demm>t e aleil.s. i smiuml. t son et egnes.it n.n Gas Control equ iv a lent t o 100% at t ive c ini s eact line A 5tria;t.us a l eva lu st tone lia t M A plus hpis u.p-n loads (JM on t est cl#1seaction).

FW 2

A Dellagrat son likely at low contentsations A Dononst rat eil actienesheil48 soit ut sj+ sete at isne

(= 10%) given hydrugen control system (lWW5I equivalent to 100% act ive cl.id s cat t hwi anal PCC5 lim it steam enert trig pot erit ta l) wit h molt iple bur ns.

A Sti oct ut a I esJ lisat eine lin I LH A p his h,ile e. gen ten,imling 9 oh.s i but n of hpis ugen l

loads, equiva lent to /54 Jct Ive c lait r eact ion 16 in-Vessel Debr is Water Internal BW / PW t

t arge-scale phenomena limited in gn obabi lit y.

Rugged s cattua vessel umtains tut te a,

Interaction Containment In-wessel geometry limits interetting backup, ruggeel lowei 4h ywell/ r eactea t it y loading quantitles and size of any int er ac t i..n contains lower lic ail f ailuie.

Il in vessel Debt is Water Intesnal BWR/PWR 2

l a r ge - sta le gehenur+na l atii t t eil in pr i,le.sh i l i t y A Ricjgent lowes air ywe ll/s ear t... tav it y Interactlun Cont a irheent l a-vessel geometry llinit s interatt ing conf is med by evalait son loading quantities and site of any inter act h n l ont a lranent des ig,i acews ol et es s t e.ua generatton i Ihe acceptabilItTof A!W s equis ements to atlitress contaisws.i:nt thallenges was based on the f ollowling ci sti e ia.

l.

Current twl reslutance to challenge acceptable for AtWR.

2.

Suf f icient AlVR design f eatures added to increase s es tstente tu sha llenge by r educing the sevei st y on.1/or ensuring contatem+ent

• Passive pleiit sles ign f eat us es n.hith eateed s tipe se tiv.icnt s ti+r tisr rent IWs die ideeit if seil with A (timswen t u a ll At W:,) ur p (gia>> it,i: AtWH3 ini l y )

19 i

.q-.

lable 4a (continued)

~

St#EARY OF RLilUIRIMINIS 10 ADilRESS CONIAINMENI CllAlllHGES RiSIR11HG TRUM CORE DAMAGI FI T PA%1VI At W Rf QtflR(MINIS Af f((l10 S AF E lf PtANI At W Cil At t ( NG(

I DMCl[0M IYPE PA5158 I f Mil P0lf Nil At FOR fitAt t t N6t.'

Aff 0MMODAll fil A.li t Nut S*

18. Noncorniens 6ble Gas f uel/Debr is BW / f'W 2

A leatures limittnq concrete erosion (se e C on t a t smn t s i t e 4,91 39 enni e setentne Ganeration Cooling Iteen 191 Ilmit noncondensible gas gener at ion capab i lit y as well.

A Sacr if ic ia l concr ete sect.ll ied as 1:~ q is generation type.

A Over lying pool cools gases le ta= cot e un r ete interaction.

cav it y/ lower da yme ll spi e.ntina) as e.s ut A Sacr if itia l uniu ele *lwir ak in is un I li.m

19. Basewat trosion and f ue l-Delir ts BW/PW 2

A Reattug/telt promotes core ikbr 6s cooling.

cont act s lioundar y st ructui es (which di e ti.e Veggel Support Cooling 0.02m Degradstion A t ower dr ywell/cawlty f looding pe=s tve BW vesse l suptan t )

A Io=er deywe1I ffooding iheemaily attmetr.1 dir ec t I rarn BW gr av it y de a in t ank or suppression pool.

A Over f low f rom contaltwnt v el lu= v ia I W INW51 preiloods reactor cavity.

A liackup capability f or water athlit seus l e ew:s sources enternal to contaiennent.

20 Core Debeis in Sump F ue l/Debr is itW / PW 2

A 5pecial cavity sump design ps events Im.a lized Cooling unterminated core-concrete interact lon-A Sump dralnline conf igurat ion prec h&s gr av it y transport of debris es-containment.

A Reactor cavity / lower dr ywell f lorulinit IIli acceptability of ALW req elrements to address containment challenges was based on the folk =ing cr ites sa:

1.

Current LWR resistance to challenge acceptable for ALVR.

2.

Suf f icient At W design f eatures added to increase vesistante to ctullenge by rediscitig tiie sevei 6t y and/or ensiir lesg conta trienent

• Passive plant design f eatures which ent.eed s e+pilsements f or cur rent tWs are ident if ied with A (r.onamni to all AlWs) or P (passive AlWs only).

~

lable 4a (continued)

SilHMARY Di~ RlQlllHtHINIS 10 ADDRESS CONIAINHINI CilAlltNGES ITESUI. LING FROM CORE DAMAGE t!Yl'AS51Vt AlW Nf0 8 M MINI $

Af flCILD

$Af[IY PLANI Al W Ctest i f M6[

FUNf g

})PE

!! AS!5#

L.lMl! P01[NJI AI FOR CHAll { N6l

• Af f fW 004lt ( Hall {fpf,jg*

21. Core Deltres Contact f ue l/ Debris BWR/PVR 2

A i iner protected by concrete With t iner Cooling A t ower dry. ell /cavit y f looding.

A Design features to limit debris dispeisel including ADS.

22. Decay lleat Generation Containment SWR 2

Hann Condenser.

P Pass ive Cont a inense vit (t,u ling Pressure A Reactor Water Cleanup System Control P Passive Rleil (NCS beat renoval smuH PVR 2

Stessa (.enciatoss/Maese iced ates (Ht W)/ Bat k eep P Pass ive Ctnit a nnisit tit t uo lienj feedwater.

P Passive lie t Memova l througti wnte <m..ns Heactor Shutdown Coolerpj shel l = s t hout PCC S =a t er li its con t a in.nent pi essui e.

.23. Iube Rupture frorn Bypass PVH 2

Steam Gene:r alors/Hf W/Dat4up f eedwates Not 6ases A Depressurtratson System f IEe acceptabil6ty of AlW requirement s to mikfress conte niment challenges was based on the f ollowing i e iter i.;

1.

Current LWR resistance to challenge acceptable for AIVR 2.

Suf f iclert AIVR design f eatures added to increase e esistence to t.hallenge tiy reducing tim: sever s t y and/or ensuring containment.

• Pass ive plant - design f eatur es which e=ceed s eignis ements tus curient tWNs ese ident al ted with A (unmin to all At Ws) or P (passive AluHs onir) i 21 1

,w

9 APPENDIX C ALWR Design Characteristics And Features Which Address Dominant WASH 1400 And Subsequent PRA Accident Sequences And Failure Modes

Appendix C ALWR Design Characteristics and Features which Address Dominant WASIl '1400 and Subsequent PRA Accident Sequences and Failure Modes LOCA No recirculation piping in BWR; minimal number of welds in RCS piping in PWR RCS depressurization system allows low pressure systems to be effective regardless of the break size.

It is unnecessary to switch to recirculation since passive containment heat removal condenses steam released into containment and retums it to the vessel by gravity.

Safety system dependencies essentially climinated (include only de power for the purpose of depressurization).

Vessel Runture Reduced RCS peak pressure for plant transients.

Improved materials:

- Less than.0129, phosphorus, weld and base metal

- Less than.039. copper, PWR hase metal

- Less than.059 copper, BWR base metal

- Less than.089. copper, weld metal

- Less than.059. vanadium, weld metal Initial ductility transition reference temperature less than 10*F (less than -20 F for PWR core belt region), reference temperature shift less than 30 F over plant life.

Low fluence at vessel wall.

No welds in beltline region.

Interfacing System LOCA Low pressure systems normally isolated from the RCS are provided with

=

interlocks to prevent their exposure to RCS pressure and are enunciated should i

high pressure conditions occur.

]

The ultimate rupture strength of potential interfacing systems is capable of 1

withstanding full RCS pressure.

i C-1

Transient (loss ofiniection)

Core passive residual heat removal system automatically actuates on loss of ac power. Passive system is fail safe and can operate independent of any support system.

Automatic depressurization and gravity injection are capable of providing adequate core cooling independent of normal makeup systems and passive residual heat removal system.

Transient (station blackout)

Core passive residual heat removal system automatically actuates on loss of ac power. Passive system is fail safe and can operate independently from any support system.

Automatic, backup ac power systems.

Battery capacity in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Canned rotor reactor coolant pumps are provided in the PWR, climinating the potential for seal LOCA (the BWR is natural circulation and has no recirculation pumps).

ATWS PWR capability to ride out an ATWS.

PWR negative moderator temperature coe ficient over entire operating cycle.

r PWR borated safety injection.

BWR capability to mitigate short term ATWS effects and shutdown automatically by diverse means:

Safety relief valve capacity > 100% power Motor drives diverse from hydraulic drive mechanisms Auxiliary Rod Insertion system diverse from reactor protective system Automatic Standby Liquid Control independent of all support systems except de power Shutdown Risk Pennanent, operable, redundant water level instrumentation designed for use during shutdown conditions.

Antisiphon provisions in refueling pool cooling and cleanup system piping to prevent pool drain down.

Features to prevent or mitigate the effects of losing suction to decay heat removal pumps during shutdown condition (e.g., piping design to minimize vortexing and air entrainment).

C-2

.. =

t a

Features to assure required net positive suction head is always available to decay heat removal pumps, Passive decay heat removal systems are capable of removing decay heat and -

preventing RCS overpressure.

Detailed requirements for analyses of mid-loop operation (PWRs) and low-level operation (BWRs) to provide assurance that known loss of shutdown cooling problems have been addressed and that information to operate the plant safely.

- 4 during shutdown has been developed.

Provision of a separate power supply circuit to the plant peimanent nonsafety-

~

+

leads for use in the event of extended unavailability of the nomial power supply.

such as may occur during shutdown.

Capability of closing valves for draining the reactor vessel or RCS without reliance on ac power.

Limitations on boron dilution flow in PWRs such that the operator has at least.

30 minutes after indication of dilution to terminate the incident prior to any recriticality.

Overoressure (steam)

Passive containment cooling systems transfer heat directly from containment without dependence on support systems, the BWR through a heat exchanger in a water pool, the PWR directly through the containment steel shell.-

Ovemressure (noncondensables) and Basemat Penetration l

Reactor cavity / lower drywell configured to promote spreading of core debris to increase coolability Ample water is available to cool debris in the reactor cavity / lower drywell passively, by means independent of potential causes for core damage.

d in-Vessel Steam Exolosion Containment failure due to in-vessel steam explosion was unlikely in WASH 1400, and has been reexamined several times since and is now considered to be j

extremely unlikely[25). This is due to improved understanding of steam explosion phenomena, particularly the extent to which water depletion in the debris-water interaction zone (due. to high heat transfer rates from debris fragments to the water and to the dispersive effect of the subsequent high steaming rates on the surrounding water pool) limits molten debris premixing and mechanica' energy yield.

Hydrogen Combustion The BWR containment is inerted.

1 i

i C-3 1

e 1

-t w

e m,s

=;

The PWR containment is required to have a hydrogen control system. Even without crediting this system, the PWR containment is capable of withstanding -

a burn associated with hydrogen generated from oxidation of as much as 75% -

of the active fuel cladding without exceeding ASME Service Level C limits.

Containment Isolation t

The passive plants have fewer penetrations as a result of safety systems being located inside containment and other changes to reduce the number of penetrations.

Most penetrations are isolated during power operation.

Penetrations which may be open during power operation are fail safe or de powered making them effectively independent of support systems.

A periodic, on line leakage monitor is specified to avoid pre-existing opening.

Liner Melt-throuch Reactor cavity and lower drywell are configured to protect the containment boundary from direct contact by core debris.

1 Ex-Vessel Steam Explosion Similar to in-vessel steam explosions, water depletion in the debris-water

+

interaction zone limits ex-vessel molten debris premixing and mechanical energy yield: also, voiding (i.e.. steam content in the debris-water-steam system) limits pressure pidse propagation to structures.

1 A rugged BWR reactor vessel foundation design is provided together with a URD requirement to demonstrate that ex-vessel debris water interactions will i

not cause loss of reactor vessel structural support.

A shield is provided in the BWR lower drywell to protect the containment r

+

boundary from the effects of debris-water interactions.

Direct Containment Heatine Both PWR and BWR have an automatic RCS depressurization system.

containing redundant trains and diversity in valve designs to prevent common cause failures. The depressurization systems require only de power for operation.

Passive decay heat removal systems are capable of mducing and maintaining the RCS at low pressures.

Cavity / lower drywell configuration is such that much of the debris will be trapped as opposed to being entrained in the steam flow. Also, recent. work i

suggests that any debris which is entrained is exposed to only a small fraction i

of the steam flow from the RCS, thus greatly limiting the potential for thermal / chemical interactions [26].

C-4

_y e

.. ~.

f Overtemocrature

[

Automatic RCS depressurization system and RCS passive decay heat removal system minimize high pressure melt ejection and resulting core debris transport into upper drywell I

Ample water available in lower drywell to cool debris and avoid high temperatures t

BWR drywell spray to reduce temperatures Steam Generator Tube Ruoture (SGTR)

Reduced primary coolant temperatures to reduce corrosion

+

Improved water chemistry and tube materials (i.e., NiCrFe alloy 690 TT).

Improved mechanical design of tubes and tube bundles.

Passive RHR prevents need for secondary side relief and steam generator overfill.

s

. Automatic RCS depressurization terminates tube leakage with no operator action.

Depressurized RCS minimizes convection of hot gases which could cause tube rupture.

L P

P C-5 i

4 s

3 1

APPENDIX D Assessment of AP600 Design Conformance with ALWR Containment Requirements P

+

1 u

~

i i

r I

I Table D-1 ASSl;SSMENT OF AP600 DESIGN CONFORMANCE WITil AI.WR REQlilREMENTS %IllCII ADDRESS CONTAINMlWF CilAl.l.ENGES CIIALLENGE AFFECTED KEY ALWR REQUIREMENTS AND ASSOCIATED SSAR OR PRA SECTIONS III III ACCOMMODATE CilALLENGE SAFETY LIMIT POTENTIAL. FOR CIIALLENGE FUNCTION

1. Containment isolation Isolation P

Reduced fluid line penetrations [6.2.3.2.1 &

P Passive Residual lleat Removal (RllR) 6.2.3.1.3-A ).

minimites core damage risk given isolation

- Isolation provisions and leakage rate testing per failure (with RiiR on-line esen wilhout DC standards 16.2.5.2.21 power) [6.31 Valves capabic of closure with maximum flow and f ull containment pressure [6.2.3.1.3.Fl.

+ Control room position indication for automatic and remote manual valves 16.2.3.1.3-11.11 P

Manual valve configuration permits laking only in closed position 16.2.3.1.3-J 1 P

Closed systems penetrating containment evaluated f or ex-vessel severe accidents

[ 6.2.3.1.1-111 Fail closed or DC powered isolation valves l 6.2. 3.1.3.K l.

P Capability for periodic pross check of containment integrity 121

2. Interlacing System Bypass P

• Reduced interf aces between the Reactor Coolant a fiessure Rehet 15.1.2 & 5.2.21 LOCA System (RCS) and low pressme systems (PR A P

• Design pressure such that f ull RCS pressure App. A.3.21 is below rupttire pressure and no leaks will P

+ liigh to low pressure interfaces provided with occur which exceed RCS makeup capacity isolation valve leak testing capability [6.2.5.2.2 15.4.7.2 & PRA App. A.3.21

& for RilR, Fig. 5.4-7], isolation valve position indientor in control room [6.2.3.1.3-11. I & lor RilR see note 21, and high pressure alarm (RilR. 7.6.l.I.1].

. Interlocks prevent isolation valve opening when l

RCS pressure exceeds RSDC system design l

pressure 15.4.7.2.21

• Doubic isolation 15.4.7.2.21 f

l l

lII Passive plant design features which exceed requirements for current LWRs are identilkd with a P.

[2} No reference in SSAR: however, Westinghouse has committed to this capability.

lw

Table D I LCemt'd)

ASSESSMENT OF AP600 DESIGN CONFORM ANCE WITil AI.WR REQUIREMENTS WillCII ADDRESS CONTAINMENT CIIAll.ENGES CilALLENGE AFFECTED KEY ALWH REQUIREMENTS AND ASSOCIATED SSAR OR PRA SECTIONS III ACCOMMODATE CilALLENCEIII SAFETY 1.lMIT POTENTIAL FOR CIIALI ENGE FITNCTION

3. Blowdown Forces Containment

• Design and ISIin accordance wuh ASME BPV

• Design containment f or double-ended Pressure Control Code 15.2.1.11 guillotine break of largest pipe [6.2.1.1.11

• Leak. Before Break 15.1.3.4 & 3.6.1.1-Pl.

4. hpe Whip and Jet Bypns

• Design and 151in accordance with ASME BPV Protection trom Jet / pipe whip where leak Impingement Code 15.2.1.1].

before break is not demonstra'ed 13.6.1.1-Leak Before Break 15.1.3.4 & 3.6.1.1 -Pl.

C; 3.6.2.3.4.2 & 3.6.2.4.11

• Use of only proven materials and lahrication processes 15.2.3.11

• Use of EPRI water chemistry guidelines 15.4.2.4.11

5. Steam Generator lube Bypass improved water chemistry 15.4.2.4.31 P

operator actions can termmate leakage Proven materials 15.4.2.4.11 prior to ADS actuation for design basis Rupture a

P Mechanical design of tubes. tube supports and leak i15.6.31 tube sheets reduce likelihood of SGTR P

Automatic Depressurization System ( ADS) l5.4.2.3.3, 5.4.2.3.4 & 5.4.2.4.21 operation terminates tube leakage P

Improved design features f acilitate SG cleaning automatically 115.6.11 and replacement 15.4.2 & 5.4.2.5).

P

• Passive RilR prevent secondary side rehef following SGTR (15.6.3].

h.ATWS Reacuvny Control

+ Diverse RPS tor capabilny to ride out A IWS

+ borated Safety injection (SI) 15.4.131 P

14.3.1.71) IPR A App Cl2].

P

• Negative moderator ternperature coctlicient t

over entire f uel cycle improves ATWS response 14.2.2.31

7. Suppression Pool Containment NOF APPLICABLE Bypass Pressure Control

8. Catastrophic RPV Internal P

= Rl'NDTs to F; initial RTNDT s -20 F ter Failure Containment PWR core beltime; low fluence at vessel wall Loading 15.3.3.11 P

No welds in beltline region 15.3.4.11 P

• Relief valves prevent overpressure, backed up by depressurization system and low-head injection 15.4.91 Design in accordance with ASME code 15.3.1.11

• Design teatures to avoid relief valve opening for espected plant transients 16.3.1.1.1 &

15.2.R.31 ill Passive plant design features which exceed requirements for current LWRs are identified with a P.

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i' lahb D-1 (Cont'd)

ASSESSMENT OF AP600 DESIGN CONFORMANCE WITil Al.WR REQUIREMENTS WillCII ADDRESS CONTAINMENT CilALI.ENGES CHALLENGE AFFECTED KEY ALWR REOUIREMENTS AND ASSOCIATED SSAR OR PRA SECTIONS III IN ACCOMMODATE CilALLENGE SAFETY LIMIT POTENTIAL FOR CilALLENGE FUNCTION Reactor cavity / lower drywell spreading arca of P

• Sacrificial concrete where debris on Door

.[

19. Basemat Erosion and Fuel / Debris P

0 02m /MWt promotes core debris cooling contacts boundary structures 13.8.2.1.21 2

Vessel Support Cooling Degradation IPRA 10.2.41.

P Reactor cavity flooding (PRA 10.2.21 P

Overflow from containment rethis via PWR 1RWST prefloods reactor cavity (PRA 10.2.21 P

Backup capability for water addition from sources external to containment IPRA App.

t C.4.4. I 1

[

Special cavny sump design prevents locahied

20. Core Debris m Sump F' dl/ Debris P

Cooling unterminated core <oncrete interaction i10.2.41 Sump drainline contiguration precludes P

gravity transport of debris ex. containment (PR A 10.2.4}.

Reactor cavity flooding (PRA 10.2.21 P

=

Liner protected by concrete i.L8.2.ill,

21. Core Debris Contact Fuci/ Debris P

Reactor cavity flooding [PRA 10.2.2}.

with Liner Cooling P

a Design teatures to limit debris dispersal P

includmg ADS (5.4.6 & 3.H.3.1.51 Steam Generators / Main Feedwater P

• Passive Containment Coolmg th.2.21

22. Decay IIcat Generation (MFW)/Startup Feedwater (10.4.91 P

+ Passive lleat Removal through containment t

Normal Residual IIcat Removal System shcIl without PCCS water limits containment (5.4.71 pressure iPRA App. L.3.1 & L.3.21 Steam (3cnerators/MFW/Startup Feedwater

[

23. Tube Rupture Irom

, llot Gases

[10.4.91 Depressurization System 15.1.21 l

P UI Passive plant design features which exceed requirements for current t VRs are identified with a P.

g

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.w.

m -

.c e e w-e.

Table D-2 Exceptions for AP600 Design Conformance With ALWR Requirements 1.

No SSAR provision exists for periodic gross check of containment integrity. Ilowever. Westinghouse has committed to the ALWR Program to provide this capability in AP600. The ALWR Program will track this item.

2.

No SSAR requirement exists for a high pressure alarm on the high-to-low pressure interface for the Primary Sampling System and the l

Chemical Volume and Control System (CVCS). The ALWR

. l Program will track this item.

3.

An inconsistency exists between SSAR Section 7.6.1.1.1. which identifies a high pressure alama on the low pressure side of the RHR System, and Figure 5.4-7 which does not show it. Westinghouse has confirmed in response to an NRC Request for Additional Information that the high pressure alarm is part of the system and that Figure SA-7 will be corrected.

4.

The existence of isolation valve position indication for the RHR System is not mentioned in the SSAR. but Westinghouse confirmed that this capability is provided in the design.

5.

No SSAR commitment to ANSI 2.12[27] exists for man-made site proximity hazards. However Westinghouse has stated that AP600 will conform to ANSI 2.12. The ALWR Program will track this item.

6.

No explicit statement is made in the SSAR regarding containment size being large enough to limit dry hydrogen concentration to less than 13% given 75% active clad oxidation. However, the ALWR Program has evaluated hydrogen concentration based on AP600

- zircaloy mass and containment volume.'and has concluded that the 13rle requirement is met.

7.

The SSAR does not currently specify low gas generation concrete in the reactor vessel cavity. - However, based on sensitivity studies for ex vessel debris coolability. the intent of the requirement is met, i.e.. avoid rapid containment overpressure due to noncondensable gas generation. even under very conservative. molten core concrete interaction assumptions.

y 3

I i

i i

i APPENDIX E Assessment of SBWR Design Conformance with ALWR Containment Requirements

'I

,b I

Talite E-1 ASSESSMENT OF SilWR 1)ESIGN CONFORM ANCE WITil AI,WR REQUIREMENTS TO Alli)RESS CONTAINMENT CilAI,1.ENGES CIIALLENCE AFFECTED KEY ALWR REQUIREMENTS AND ASSOCIATED SSAR SECTIONS

!l!

III ACCOMMODATE CIIALLENGE SAFETY LIMIT POTENTIAL FOR CllAI.LENGE FUNCTION

1. Containment isolatim isolation P

Rufuced fluid line penetrathms {2l P

Passive Residualliest Removal (Isolatiem Isolation provisions and leakage rate testing per Condenvr System [ICS)) minimises core d.unage standards [6.2.4. ll.

rist given isolation failure (mith ICS on linc even e

Valves capable of closure with madmum flow and full widmut DC power)15 4 6 & 5.4 6 21 I

amtainment pressure [6.2 4.2.51 Control room position indication fe automatic and remote manual valves (6.2.4.21 P

Manual valve omfiguratk n remuts twking onty in dosed position l3 j.

P Closed systems penetrating containment evaluated for ex-vessel severe accidents [19.lL5 2.11 1:ait chwed or DC powered isolation valves 16 2 All.

P Capability for penodic gross chetk of umtainment interrity [16.1.4]

2. Interf acing System IJK'A 11ypass P

Reduced mterf aces between the Reactor Coolant Pressure Rehef 163 31 System (RCS) and low pressure systems 17 6.1.1.

P Design prenure such that full RCS Pressure is 1911.2.25 & 1911.2 441 below rupture pressure 19112 2 for I.PCI)

P

• Ifigh to low preuure interfaces pnnided with isolation valve leak testing capability [9.112.2 for I.PCil.

isolation valve posithm indicator in control rmm (9.112.2 for i PCl]. aral high preuure alarm [il P

RWCU/SDC designed for full reactor prewure 15.4 H.I.2l

• Ihmble isolation 17 612]

Ill Passive plant design features which exceed requirements for current LWRs are identified with a P.

[2]

No reference in SSAR; however, General Electric has committed to this capability

[3]

No reference in SSAR '

9 2- -

n a

Talite E-1 t

ASSESSMENT OF SilWR DESIGN CONFORMANCE WITil

~

A1,WR REQUIREMENTS TO ADDRESS CONTAINMENT CIIALI.ENGES (Cont'd)

CIIALLENGE AFFECTED KEY AI.WR REQUIREMENTS AND ASSOCI ATED SSAR SECTIONS IIl N

SAFETY LIMIT POTENTIAL FOR CIIALLENGE ACCOMMODATE - CIIAl,LENGE 1

FUNCTION

3. filowdown ihren Cemtimnent Prewure

• Design avalISIin acumfance with ASME HPV Code

• Design umtainmeat for double cudal guilh4me Control l 5.2.1 j.

break of largest pipe [6.2.1.1.1).

• leak Hefore Break {16 3]

4. Pipe Whip and let Itypass

• Design and ISIin acconlance with ASME HPV Code

• Protet: tion from jet / pipe whip where leak before impingement 15 2.11 break is not dernonstrated ]16.I.3 & 3.611

• Imak Ilefore fireak l3 6.3)

• Use of only proven materials & fabrication processes

[ Tables 5.2.1 & 5 2-4l

• Use of FPRI water chemistry ruidelines [12.12 21

5. Steam Generator Tube 14ypass NOT APPLICABLE Rupture I

6.ATWS Reactivity Control P

• Diverse Reactor Prutection System and Alternate Rod

• Standby laymd Control (SLC)l9 3.51 Injection [4 6.1.2.5. 7.2.1. 7.4.5].

P

• Chetkerboard pattem of scram group nuts P

• Diverse means of ro11inse1 ion [4.6.1 21]

maximires gnmp worth (21

7. Suppression Pool Hypass Centainment Presmre

• Vacuum Hreakers: potentialloads accounted for.

• ADS use of SRVs which diwharge to suppreuion Control position indication, minimal leakage [6.2.1.1.2 &

pool and thus ensure vapor suppression despite 1.9A4.1.11l.

leakage 15.2.2 & 6.3.3.21 P

• No high energy lines in weiwell airspace [ Figures 5.2 P

• Passive ICS tincluding PCCS)[5A 6 & 6 2 2l 1 and 21.1.2-2].

8. Catastrophic RPV Faihste Intemal Containment P

• RTNDTsl0*F; low fluence at vessel walil5.3 2.Il r

Inadmg P

• No welds in beltline region 15.3.3.31 P

• Relief valves prevent overpressure, baked up by depressurization system and low. head injection {5.2 2.

6.3.2. & 6.3.3 l.

• Design in acconlance with ASME code [5.3.I.l]

• Design features to avoid relief valve opening for expected plant transients [19AE.8 3.21

9. IntemalVacuum Containment Pressure

• Containment internal design loads specifications. Table

• Vacuum Hreakers [6.2.1.1.21 Control 6 2.1 through 6.2.6.

• Design for external pressure loads (6.2.1.1.2 &

6.2.1.1.3 ).

Ill Passive plant design features which exceed requirements for current LWRs are identified with a P.

[2]

No reference in SSAR; however, General Electric has committed to this capability.

1 4

w

,-4

Table E-I ASSESSMENT OF SitWR DESIGN CONFORMANCE WITil ALWR REQUIREMENTS TO ADDRESS CONTAINMFNr CilALLENGES (Cont'd)

CIIALLENGE

-AFFECTED KEY -AIAVR REQUIREMENTS AND ASSOCI ATED SSAR SECTIONS SAFETY LIMIT POTENTIAL FOR CIIALLENGEIU ACCOMMODATE CIIALLENGE FUNCTION '

10. Intemal (Plant) Missiles Ex temal Containment

+ Turbine overspeed protectbm [10.2.2.4[.

• Turbine orientation avoids minite wntact with Inabng P

Impmved turbine integrity Asne piece rotors i10.23 &

wntainment (3.5.1.1,1l.

10.2.3 A].

• Missile protectiem for any safety related cornponents in missile path (SRP 3.513)135I

&.1.5.3)

I

11. Tornaki and Tomado Extemal Contamment Conformance with ANSI 2.12 & ANSI $1.5 (m P

• Passive core coolmg systems heated within Missiles Inaling acterdance with ANSI A38 I ant! ASCE Paper containment [5 4.6 & 6.2.2]

Number 1269)[ }

12. Man Made Site Pnixumty Ex temal Containment

• Cimiormance with ANSI 2.1212]

P Passis e core coohng systems krated w ithm Ilarants Inaline containment {5 4 6 & 6 2 21

13. Seismic Extemal Containment

• Sitmg reymrements exclude the most vulnerable sites P

SSE of 03g [2.5 2 & 3.71.11 Inading inn effect on design].

P

• Evaluation at > SSE with margins assenment as part of design prmess l19Dl-P

• Addreu vulnerabilities fmm past experienue, e g.

provide common basemat [ 4 R 4.ll.

14. liigh Pressure Melt Reactor Pressure P

• Ihverse depressurizathm systems [6 3 3 &

• Suppressi<m pool ctets heated gases [6.21.1.2 &

Ejectism (HPME)

Control 19 4.4.l_5).

19.4.4.1.41 P

- Pauive ICS can aid deprenurisatsm l5 4 63]

• Inerted containment (no comte. tem heat aAhti.ml

[9 4.R & 19 4.4.1.101 15a,11ydmgen Generathm Combustible Gas

- Inerted [9A 8.19.4.3 A & 19.4 41.10l P

Eva!uation required if local detonation is possible to Detonable Limits Contro!

{l943A).

il1 Passive plant design features which exceed requirements for current LWRs are identilied with a P.

[2]

No reference in SS AR: however, GE has crimmitted to this capability.

f m

Talite E-1 ASSESSMENT OF SRWR DESIGN CONFORMANCE WITII ALWR REQUIREMENTS TO ADDRESS CONTAINMENT CilALI.ENGES

~

(Cont'd)

C11ALLENGE AFFECTED KEY ALWR REQUIREMENTS AND ASSOCIATED SSAR SECTIONS U!

UI SAFETY LIMIT POTENTIAL FOR CHALLENGE ACCOMMODATE CllALLENGE FUNL710N 15b. llydmgen ikflagration Combustible Gas Inerted. [9 4.8 & 19.4.4.1.101 P

• Denmnstrated accommestathm of generatam Control equivalent to 10lY1 active clal reactium l19H.31.5

& 19.G.2.451 P

+ Stmetural evaluathm for i OCA plus hydmgen loads (100% active clad reaction)l MH3 2.5 &

19 G.2 451

16. in-Vessel ikbris. Water Internal Contamment 1.arge-scale phenomena hmited in probability

- Rugged reactor vessel omtams forces; as backup.

Interacthm Inading

[19HH]

rugged lower drywell contains lower heat failure In-vessel geometry limits interacting quantities and i19.4 2.21 sire of any interactum [19HH].

17. Ex Vessel Debns. Water External Cont sinment targe-scale phenomena tunited m pnibabihty P

- Rugged lower drywell confirmed by es aluatu m Interaction inadrng

[19.4.11 & 19HH.51

[19.4 2.2,19II..l.2.5 & 19HH.3 4]

EUvessel geometry limits interacting quantities and

. Containment design accommalates steam size of any interaction [19HH 211 ceneration [19Bil 5.41 1H. Noncondensible Gas Fuel /Ikbris P

Features hmiting concrete erosion isee item 19)

Containment site and pressure retenton capabihty Generation Cooling limit noncondensible gas generation as well.

[1911.6.2.1 & 19H.6.2.21 P

Overlying pool cools gases from core.conctric interaction i19 AF 7.1]

19. Itasemat Emson and Fuel /Debns P

Inwer dtywell spreading area uf 0 02m /h1W P

• Sacrificial amcrete where debris on floor contacts 2

Vessel Support Cooling pn, motes core debris auling l19 4 4.1 gj.

luundary stmctures (which are the passive llWR Degrafation P

1.ower drywell flooding [6.2 I.1.2 & 19 4,4.1.31 vewel surgwrtI [19.4.4.1.71 P

I nwer drywell thuuling thermally actuated directly from gravity drain tank or suppression pool 16.12.2

& 19.4 4.1.3j.

P llackup capability for water addition fnun sources external to containment 19.1.3.1]

ll} Passive plant design features which exceed requirements for current LWRs are identified with a P.

l l

l

3 Table E-1 ASSESSMENT OF SilWR DESIGN CONFORMANCE WITil AI,WR REQUIREMENTS TO ADDRESS CONTAINMENT CIIAl.l.ENGES t C<mt'd)

CHAI,LENGE AFFECTED KEY ALWR REQUIREMENTS AND ASSOCIATED SSAR SECTIONS IU III SAFETY LIMIT POTENTIAL FOR CIIALLENGE ACCOMMODATE CIIALLENGE FUNCTION Special cavity sump design pn vents localized

20. Core ikbris in Sump Fuel /lkbris Couting P

untenmnated cure-concrete interaction [2].

Sump drainline configuration precludes gravity P

tranvr1 ef debris cuontainment (2).

lower drywell llonding (6.2.1.1.2 & 19 4 41.31 P

21. Core ikbris Contact Fuel / Debris Coolmg P

tmer protected by concrete i19 4 41.7l with Liner P

Irwer drywell 11n. Ming l6.2.1.L2 & 19 4.4.1.11 P

Ikugn features to limit debris disperut incluihng ADS 16 33 (ADSL 19H.10 2.4 & 19HH.13 teorium shichlil Main Condenser l 5.4 71 P

- Pmive Containment Coolmg 16.2.2 &

22. Ikcay lleat Generatum Containment Pressure Reactor Water Cleanup System tRWCS)15.4 7 &

19.4.412).

Control P

5481 P

Pmive RilR (RCS heat removal m.ide)is a 6 &

5 4.7l.

23. Tube Rupture from Hypass NOT APPL.lCAHl.E Hot Gases 4-(I] Passive plant design IEtures which exceed requirements for current 1,WRs are identitied with a P.

[2] No reference in SSAR: GE is currently considering this requirement.

F Q

Table E-2 Exceptions for SBWR Conformance Assessment With ALWR Requirements 1.

The fact that the number of containment fluid line penetrations has been reduced is not explicitly mentioned in the SSAR, but General Electric has stated that this is the case in the SBWR design.

2.

There is no SSAR provision that manual containment isolation valves permit locking only in the closed position. The ALWR Program will track this item.

3.

No SSAR reference to isolation valve leak testing capability and position indication in control room was identified for sampling line_s. General' Electric indicated that this capability exists. The ALWR Program will confirm this item as part of conformance assessment.

4.

No SSAR reference to high pressure alarms for high to low pressure interfaces was identified. The ALWR Program will track this item.

5.

There is no SSAR requirement for a checkerboard pattern of control rods within a scram group to maximize group worth, but General Electric has stated that this provision is in the SBWR design (it became standard practice in recent operating BWRs).

6.

Conformance with ANSI 2.12[27] for man-made site proximity hazards was not identified in the SSAR, although the design approach appears consistent with ANSI 2.12. The ALWR Program will track this item.

7.

No provision currently exists in the SSAR for cavity sump and sump l

drainline design to prevent localized core concrete interaction and ex-containment gravity transport of core debris in the event of ex-vessel core damage. General Electric is considering design features in this regard. The ALWR Program will track this item.

I

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