ML20237G357

GE Advanced BWR Licensing Review Bases

ML20237G357
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Issue date:	08/31/1987
From:	Office of Nuclear Reactor Regulation
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PROJECT-671A NUDOCS 8708140043
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Enclosure i

GE Advanced Boiling Water Reactor Licensing Review Bases August, 1987 rikoj3070007 PDR

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1 INTRODUCTION GE intends to submit an application for final design approval (FDA) and design Initial certification portions of the (DC) forAnalysis Safety the Advanced Boiling)

Report (SAR willWater be submitted Reactor beginning (ABWR). in the fall of 1987. Both the NRC staff and GE believe that staff's safety review of the SAP will proceed more smoothly if certain licensing review bases are i estaci!shed before the review starts. These bases are intended to address aspects of the review process and certain technical issues that have caused difficulties in past reviews of standard plant designs.  ;

The Standard Review Plan (SRP) is the basic staff document which will be used in the ABWR review. The Licensing Review Bases addressed in this enclosure provide supplementary guidance on regulatory issues and areas which are either not addressed at all, or not covered in detail, by the SRP. In many cases, (

these are areas where the staff's positions and requirements are evolving.

These Licensing Review Bases contain no new regulatory requirements. In certain key areas where Commission policies and staff positions are still i

under development, both the staff and GE have committed to implement  !

acceptance criteria which, if satisfied by the ABWR standard design, would l result in a licensable design. Should, however, substantial new information  !

become available that results in new requirements being promulgated by the i NRC, they will be addressed during the course of the ABWR review. j i

The staff supports the efforts of the Department of Energy (DOE) and GE to obtain design certification (DC) of the ABWR. Once the design has been certi-fied, it could be referenced by a number of applicants for use on a number of different sites without further review, except for matters which cannot be reviewed or accepted until a specific facility is constructed. These matters would be specifically identified in the staff Safety Evaluation Report (SER).

When an applicant references the pre-approved design, the staff would conduct a compliance review to confirm that the plant was built in accordance with the DC. The design would be certified for the period specified in the Commission's >

Policy Statement on Standardization, with an option for renewal.

GE has agreed that all Generic and Unresolved Safety Issues relevant to the ABWR will be resolved for the ABWR design before a Final Design Approval (FDA) is issued. After an FDA is issued, new issues will be considered for back-fitting under the provisions of 10 CFR 50.109, or under other applicable l Commission regulations. j GE is to provide a Safety Analysis Report (SAR) for the entire nuclear island design. The SAR will meet all applicable Commission regulations and contain enough information for the staff to complete its safety review. -

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1.1 Scope and Content of the ABWR SAR The scope of the ABWR standard design, as illustrated in Figure 1-1, is a nuclear island. The SAR is to include all of the information necessary for the staff to complete its SER. This includes interfaces (design, construction, testing and operational) between the ABWR standard design and the remainder of plant. GE may later expand its submittal to include some or all of the remainder of plant.

The ABWR standard design employs an advanced boiling water reactor enclosed in a steel-lined reinforced concrete containment vessel integrated with the reactor building which, in turn, forms the secondary containment boundary.

The reactor building houses the equipment associated with auxiliary systems (such as emergency core cooling, residual heat removal and reactor water cleanupsystems). It also houses the fuel handling and storage and the diesel generators, which have traditionally been housed in separate buildings. A separate control building is located adjacent to the reactor building. The control building includes the control room, change rooms and plant supervisor's office and provides plant access control.

Because GE wishes to obtain an FDA and DC for the ABWR design before any applicant, site, architect / engineer or equipment suppliers are identified, it is necessary that GE provide the necessary level of detailed information to enable the staff to complete its review without preempting competitive bidding on any future project that references the certified design. The technical

.information for the ABWR standard design portion of the submittal must meet the '

requirements of 10 CFR 50.34(g) and the guidance in Regulatory Guide 1.70, Revision 3, appropriate to the degree of design available for standard designs. The corresponding contents of the SAR are listed in Section 8.4.1.

Tests, inspections, analyses, and acceptance criteria necessary for an appli-cant to assure that the designs are properly implemented in the plant will also be defined in the SAR. The applicant will later demonstrate compliance with this design and implementation information.

Section 8.4.1 lists the design documentation that GE intends to submit to

, support the ABWR standard design. GE does not plan to submit these types of supporting documents for the remainder of plant for the DC effort.

1.2 Scope and Content of Future Applications Referencing the ABWR When the approved ABWR standard design is referenced in an application, the staff's review of matters related to the approved design need consider only whether the interface requirements have been satisfied in the referencing application (the applicant's Final Safety Analysis Report (FSAR)). Specif-ically, for those areas in the remainder of the plant and the site envelope where the ABWR SAR has specified interface requirements, the applicant will ,

have to demonstrate compliance with them. No further review of the referenced '

design will be required when the site envelope parameters fall within the ,

design envelope and the interface requirements are met.

2 SCHEDULE I The schedule for the FDA review of the ABWR design is shown in Table 2-1. The f schedule for the subsequent design certification rulemaking phase depends on,

! among other factors, the type of rulemaking proceeding selected by the l Commission. The range of dates shown is the staff's current best estimate of the design certification duration.

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,. .- l Table 2 ABWR FDA Review Schedule i Cumulative i Review Elapsed Time, I Review Element Schedule

• Months **

Chapters 4, 5, 6, 15 9/87-3/89 19 l (Reactor,ReactorCoolantSystem, Engineered Safety Features, ,

Accident Analyses) j Chapters 1, 2, 3, 17 3/88-9/89 25 ]

(General Description, Site Char- i acterization, Design of Structcres.

Components, Equipment, and Systems., QA)

Chapters 7-9, 11-14, 16 6/88-12/89 28 (I&C, Electric Power, Auxiliary J Systems, Radioactive Waste, Radiation Protection, Conduct of Operations, j-Initial Tests aM Operation, Technical Specification )

Chapters 10, 18 i

1/89-12/89 28 (Steam and Power Conversion, '

Emergency Planning)

PRA and Failure Modes & Effects Analysis (FMEA) 1/89-12/89 28  !

Integrated Review / Final SER 3/89-2/90 30 l ACRS Review 9/87-4/90 32 Proposed Decision Date for FDA 9/90 37 Design Certification 49-61

• Time from submittal of chapters to issuance of draft SERs

• • From beginning of NRC review 3 CONTENT OF APPLICATION 3.1 Safety Analysis Report Format The ABWR SAR and all subsequent SAR amendments are to be organized in accordance with Regulatory Guide 1.70, Revision 3, and the SRP in effect on March 30, 1987. The application will include the infonnation specified by 10 CFR 50.33, 50.34, and Appendix 0 to Part 50. GE will comply with the provisions of 10 CFR 50.34(g)(1)(ii).

3.2 Use of Metric Units k

Because the ABWR has been designed for international applications, the i SAR may use metric units in describing equipment dimensions and - -

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performance. However, the values in the SAR are to have their corresponding l English units included, in parentheses, next to the metric values (e.g., maximum 4 fuel cladding temperature during an accident is 1204 'C (2200 F)). GE intends to include a table of conversion factors between metric and English units at the front of each volume of the SAR.

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3.3 Data Interchange Format Because of the increased use of computers in the licensing _ process, GE plans to provide a copy of the SAR on a diskette suitable for use on an IBM (or compatible) personal computer (except for drawings and graphs that are not amenable to such portrayal). GE also is to provide the requisite number of hard copies of the SAR specified in 10 CFR 50.30(a), (c)(1) and (3).

4 INCORPORATION OF FUTURE ISSUES As stated in its Severe Accident Policy Statement (see Section 7 below), the Commission expects all new power plant designs to address all Unresolved Safety Issues (USIs) and all medium-and high-priority Generic Safety Issues (GSIs).

NUREG-1197, " Advanced Light Water Reactor Program", December 1986, presents

l. these issues and their status as of July 1,1986. GE is to identify which issues are applicable to the ABWR design and address them. These issues will l

include both applicable issues identified in NUREG-1197 and any new generic issues raised up to the time of FDA issuance. It is the intention of the staff that there will be no open items regarding the resolution of USIs or GSIs or other plant features for the ABWR at the time of the FDA decision.

Issues introduced after an FDA is issued would be analyzed and resolved in accordance with the backfit requirements of 10 CFR 50.109, except to the extent that the DC rulemaking provides otherwise.

5 STAFF REVIEW PROCEDURES The staff will follow its review procedures in the SRP, supplemented and modified as follows:

(1) The ABWR SAR is to be submitted chapter by chapter, over a period of about 16 months. Correspondingly, the staff SER will also be issued in draft form, in sections in accordance with the schedule shown in Section 2. The draft SER sections will be made publicly available.

(2) At the completion _of the review of the individual SAR chapters, the staff will perform an integrated review of the application. This review will complement the Probabilistic Risk Assessment (PRA) review, in that it will be an overall assessment of the design. The staff will issue a composite final SER in accordance with the schedule described in Section 2.

(3) It will be important to carefully document open or unresolved issues that may be identified early in the review process, but which cannot be resolved until the completion of later chapters. Each draft SER section will contain a description of such issues. In addition, with the submittal of each chapter of the SAR, GE is to provide an updated check-list which identifies outstanding issues and the future chapter (s) in which resolution is anticipated. -

(4) Each draft SER will contain a target schedule for closing outstanding SER l issues that is compatible with the target FDA decision date.

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6 ACRS PARTICIPATION  ;

One step in the design review of a standard plant is the independent review by the Advisory Committee or Reactor Safeguards (ACRS). The ACRS review of the  ;

ABWR design certification process started before submittal of the first chapters of the ABWR SAR. An initial briefing of the ACRS by the staff and GE  ;

took place early in 1987. Periodic reviews will address the safety aspects l of the design on matters selected by the ACRS. The ACRS review is scheduled to continue through April 1990, when the ACRS will be requested to issue a letter i report on its review.

l The staff will keep the ACRS informed of the progress of the review by for-warding to it copies of the SAR chapters as they are submitted, along with copies of the draft SERs as they are issued. In addition, the staff will meet with the ACRS, as needed, to discuss the draft SERs.

7 SEVERE ACCIDENT POLICY STATEMENT (SAPS) 7.1 Introduction On August 8, 1985, the Commission issued a Policy Statement on Severe Accidents I (50FR32138, " Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants," and NUREG-1070, "NRC Policy on Future Reactor Designs"). The policy statement provides criteria and procedures for the licensing of new plants, and sets goals and a schedule for the systematic examination of existing plants. The Commission encouraged the development of new designs that might realize safety improvements and stated that it intends to take all reasonable steps to reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an accident, should one occur. The Commission's licensing criteria for new plant designs are spec-ified in the policy statement.

The Commission also recognized the need to provide defense-in-depth by striking a balance between accident prevention and consequence mitigation, through a better understanding of containment performance, with the understanding that new performance criteria for containment systems might need to be established.

It also recognized the importance of such potential contributors to severe accident risk as human performance and sabotage, and determined that these issues should be carefully analyzed and considered in the design and operating procedures for a nuclear facility. Specific discussions of each of the policy statement licensing criteria follow.

7.2 Construction Permit / Manufacturing License Rule GE will comply with all applicable Commission regulations, including those listed in 10 CFR 50.34(f) applicable to the ABWR, except 10 CFR.50.34(f)(2)(i) and10CFR50.34(f)(3)(iv). Any future applicant that references the ABWR design must satisfy 10 CFR 50.34(f)(2)(i) by providing simulators. With regard to 10 CFR 50.34(f)(3)(iv), GE has stated that the ABWR design has specific features that function to mitigate the consequences of severe i accidents within the offsite dose objectives discussed in Section 7.5 (Severe i Accident Performance Goals). GE intends to provide justification to demon-strate that a dedicated containment penetration is not required in the ABWR design. The staff will consider this justification as part of its review.

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7.3 Resolution of USIs and GSIs i See Section 4, above.

7.4 Probabilistic Risk Assessment (PRA)

GE has proposed certain criteria and methodologies relevant to PRAs in the following sections. These criteria and methodologies will be used by the staff as the bases for their review of the ABWR unless new criteria and methodologies are promulgated by the NRC.

7.4.1 Scope GE has ~ committed to provide a level-3 PRA for the ABWR design, as defined by the "PRA Procedures Guide," NUREG/CR-2300. The level-3 scope PRA includes the following elements:

(1) An analysis of the plant design and operation focused on the accident sequences that could lead to a core melt, their basic causes, and their frequencies (2) An analysis of the physical processes of the accident sequences and the response of the containment (3) An analysis of the transport of radionuclides to the environment and an assessment of potential public health consequences Bounding analyses of external events that can be quantified (e.g., seismic, internal fires, internal floods, tornados) are to be included in this evaluation.

7.4.2 Methodology The PRA is to be based on a methodology that originated with the approach taken in the Reactor Safety Study (WASH-1400), and that has been developed and systematized through applications in numerous plant-specific studies.

The general procedures have been documented in NRC NUREG reports, such as the "PRA Procedures Guide" (NUREG/CR-2300) and the "Probabilistic Safety Analysis Procedures Guide" (NUREG/CR-2815).

GE intends to utilize the IDCOR-developed Modular Accident Analysis Program (MAAP) which has been modified by GE for utilization in BWR analyses. If technical disagreements surface between the NRC methods (such as the Source Term Code Package) and MAAP, specific sensitivity studies will be performed. GE is to use the CRAC-II code or other suitable model acceptable to the staff to compute the potential consequences of fission product releases. _

7.4.3 Reference by Utility Applicants The PRA is intended to be applicable to all sites within the ABWR envelope. It is contemplated that applicants will not have to prepare or submit plant-specific PRAs before an operating license is issued, but that the GE PRA would be updated by the licensee within 2 years after a plant is licensed.

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7.5 Severe Accident Performance Goals This section describes the goals for severe accident performance criteria, to which GE has committed for the ABWR design, consistent with existing staff regulations and policy statements pertaining to severe accidents and defense-in-depth through a balance between accident prevention and consequence mitiga-tion. The staff will utilize these goals as the basis for their review unless new criteria are promulgated by the NRC.

7.5.1 Prevention of Core Damage GE intends to demonstrate by analyses that the likelihood of core damage will have a megn value of less than one in one hundred thousand reactor years (i.e.,

1.0 x 10' ), including both internal and external events. (The staff will determine the adequacy of this goal and the analyses.)

7.5.2 Mitigation of Core Damage GE has stated that the ABWR design will provide protection against containment failure if a severe accident occurs and results in core damage. GE intends the containment capabilities to include:

a. Measures to reduce the probability of early containment failure for dominant accident sequences

b. Measures to accomodate hydrogen generated from the reaction of the equivalent of 100% of the zirconium in the active fuel clad, consistent with 10 CFR 50.34(f), as provided for by the Severe Accident Policy Statement

c. Highly reliable heat removal systems to reduce the probability of  ;

containment failure by loss of heat removal (

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d. Reliable means to prevent hydrogen deflagration and detonation, consistent with 10 CFR 50.34(f), as provided for by the Severe Accident Policy Statement.

7.5.3 Offsite Consequences for Severe Accidents GE has committed to meet the following goals:

(1) The expected mean frequency of occurrence of offsite doses in excess of 25 Rem beyond a half mile radius from the r million reactor years (i.e.,1.0 x 10~6)eactor is to be both

, considering less internal than once andper external events (2) The containment design is to assure that the containment conditional failure probability is less than one in ten when weighted over~ credible core damage sequences 8

8 ADDITIONAL TECHNICAL ISSUES 8.1 Introduction The ABWR design will incorporate several features that are novel or which have been used in relatively few other nuclear power plants in the United States.

In addition, because standardized plant reviews are conducted before actual facility applications are made, these reviews cannot address every aspect of a .

facility. This section is intended to address some of the issues which arise i from these circumstances and that have caused difficulty in previous standard plant reviews.

8.2 Physical Security i

8.2.1 Basis for Requirements 1 The basis for the requirements will be as defined in 10 CFR 73.55,

" Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage," and other applicable portions of 10 CFR Part 73.

8.2.2 Acceptance Criteria The ABWR SAR is to include enough information to demonstrate the existence of adequate physical barriers to protect vital equipment in accordance with 10 CFR 73.55(c), " Physical Barriers," and to identify access control points to all vital areas in accordance with 10 CFR 73.55(d), " Access Requirements."

The ABWR SAR will not provide details but is to identify design requirements to be satisfied by an applicant for the following sections of 10 CFR 73.55 (the applicant must then address all remaining requirements):

(b) Physical Security Organization (e) Detection Aids (f) Communication Requirements (g) Testing and Maintenance (h) Response Requirements The design requirements are to include reference to existing NRC documents such as Regulatory Guide 5.44, " Perimeter Intrusion Alarm Systems" and NUREG-0908,

" Acceptance Criteria Evaluation of Nuclear Power Reactor Security Plans", as well as to industry standards such as IEEE-692-1986, "IEEE Standard Criteria for Security Systems for Nuclear Power Generating Stations." _

8.3 Site Envelope Parameters and Soil-Structure Interaction Analysis Selected site envelope parameters and methods for soil-structure interaction analysis proposed by GE are provided in Appendix A to this enclosure.

8.4 ABWR Design 1

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. 8.4.1 ' Completeness of. Design The ABWR SAR is to provide essentially complete design information. The term

" essentially complete" is defined as follows:

- (1) The SAR will-define the major design components and include the results of sufficient engineering to identify, as appropriate:

a. design basis criteria

b. analysis and design methods

c. functional design and physical arrangement of auxiliary, B0P, and NSSS systems

d. plant physical arrangements sufficient to accommodate system and components

e. functional'and/or performance specifications for components and materials sufficiently detailed to become a part of associated procurement specifications

f. acceptance / test requirements

g. risk assessment methodology (2) Design documentation for systems, structures, and components should include as appropriate:

a. design basis criteria

b. plant general arrangements of structures and components, including piping system layouts

c. process and instrumentation diagrams, electrical system layouts, and major conduit and cable tray layouts

d. control logic diagrams

e. system functional descriptions and supporting studies and analyses

f. component and procurement specifications, including acceptance criteria and test requirements ,

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g. construction and installation specifications, including acceptance )

criteria and test requirements

h. program for the assurance of quality -

1. design-related aspects for the emergency plans

j. supporting design documentation such as site envelope data and calcula-

.tions sufficient to support the level of design detail noted above .

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k. design-related aspects of the physical security program '

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1. an ALARA radiation protection plan m.- ' accident analyses n.. Technical Specifications

o. risk analysis In the limited cases where design information is not available, GE is to provide information on methods, procedures, and performance criteria. GE also is to define those related tests, inspections, analyses, and acceptance criteria that are necessary to assure that the designs are properly implemented in the plant. These tests, inspections, analyses, and acceptance criteria are intended to be implemented and verified in a series of reviews by the applicant during construction and pre-operation. The staff will monitor the performance of these reviews and implementation of the design through its inspection program.

The. degree of design detail necessary for providing an essentially complete design is to be that detail that is suitable for obtaining specific equipment or construction bids and to demonstrate conformance to the design safety limits and criteria.

8.4.2 Program for the Assurance of Quality in Design I The design process and resultant design documents must meet the quality assurance (QA) requirements delineated in Appendix B of 10 CFR Part 50, as addressed in Section 17.1 of the SRP. GE must submit justification, acceptable to the staff, for any deviations from Appendix B.

8.5 Interface Assumptions Affecting Safety Determinations for the Nuclear Island The nuclear island scope of the ABWR reduces the number of interfaces between the nuclear island and the remainder of the plant. GE is to provide a list of the assumptions relied upon to make safety determinations for the nuclear island design. This listing is to identify the nuclear system, the instrumen-tation and control requirements, reliability assumptions and specific performance criteria. GE is to use the results of the PRA to indicate which interfaces are particularly sensitive to deviations.

8.6 Instrumentation and Controls GE has committed to use standards and criteria and provide information pertaining to instrumentation and controls for the ABWR as discussed in Appendix B to this enclosure.

8.7 Water Chemistry Guidelines -

The maintenance of proper water chemistry in BWR cooling systems is essential to the prevention of stress corrosion cracking of austenitic stainless steel piping and to the minimization of plant radiation levels due to activated corrosion products. GE has committed to using at least the following documents j in this area:

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1 (1) EPRI NP-3589-SR-LD, "BWR Water Chemistry Guidelines," April 1985 (2) EPRI NP-4500-SR-LD, " Guidelines for Permanent Hydrogen Water Chemistry Installations," March 1986; revised, November 1986 (3) EPRI NP-4474, "BWR Radiation-Field Control Using Zinc Injection Passivation," March 1986 8.8 Maintenance and Surveillance GE is to provide in the SAR the reliability and maintenance criteria that a future applicant must satisfy to ensure that the safety of the as-built facility will continue to be accurately described by the certified design. The SAR is to include the key assumptions of the PRA and other PRA licensing commitments.

8.9 MSIV Allowable Leakage and Related Dose Calculations GE has committed to an ABWR standard design that will provide a non-safety related main steam isolation valve (MSIV) leakage processing pathway consistent witi, those evaluated in NUREG-1169, " Resolution of Generic Issue C-8," August 1986. The allowable MSIV leakage is to be determined based on the calculated total dose (using methodologies consistent with NUREG-1169) from all leakage sources and the exposure guidelines of 10 CFR 100.11. In addition, leakage for the final installed MSIV test is to be less than 50% of the value allowed to account for equipment degradation during the design lifetime.

8.10 Safety Goal Policy Statement On August 4, 1986, the Commission published a policy statement on " Safety Goals for the Operation of Nuclear Power Plants" (51 FR 28044). This policy statement focuses on the risks to the public from nuclear power plant operations. Its objective is to establish goals that broadly define an acceptable level of radiological risk.

Although the implementation requirements for the Safety Goal Policy Statement are still being developed by the staff, GE has committed to severe accident performance standards and criteria that are intended to assure compliance with those eventual requirements.

9 FINAL DESIGN APPROVAL The staff may issue an FDA after it and the ACRS complete their reviews of the final design. The FDA means that an entire nuclear power plant design or major portion thereof is acceptable for incorporation by reference in individual applications for construction permits, operating licenses, and manufacturing licenses. The staff and the ACRS intend to use and rely on the approved final design in their reviews of those.

applications. However, an approved final design is subject to litigation in individual licensing proceedings on those applications. An FDA is a j prerequisite for a design certification.

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. 10 DESIGN CERTIFICATION l 10.1 Introduction The Commission currently is considering staff-proposed revisions to its 1978 l policy statement on standardization of nuclear power plant designs. The l Commission also is developing proposed regulations that will address licensing i reform and standardization and provide a regulatory framework for implementation l of the standardization policy, including Commission certification of standard l designs by rulemaking. Since design certification is the ultimate goal of the  !

ABWR program, and since the focus of the proposed policy statement and i regulations is reference system design certification, the essence of these l proposals, and GE's commitment to them, is summarized here. It should be noted, however, that the Commission has not yet acted on these proposals and that they are subject to change.

The staff-proposed revisions to the policy statement encourage the use of standard plant designs in all future license applications. The staff believes that the use of standard plant designs can benefit public health and safety by:

(1) Concentrating the resources of designers, engineers, and vendors on particular approaches (2) Stimulating standardized programs of construction practice and quality assurance (3) Improving the training of personnel (4) Fostering more effective maintenance and improved operation The staff believes that the use of such standardized designs can also permit more effective and efficient licensing and inspection by the NRC.

10.2 Design Certification Concept The design certification concept, as described in the staff's proposed standardization policy statement, provides for certifying a reference system design (such as the ABWR) through rulemaking. In this process, the Commission would certify a design after the staff issues an FDA and a rulemaking pro-ceeding is completed. The design certification means that the portions of the nuclear power plant design that have been reviewed are accepteble for incor-poration by reference in an individual license application. The conclusions of the certification rulemaking would be used and relied on by the staff, the ACRS, the hearing boards, and the Commission in their reviews of applications that reference the design. The certified design would not be subject to litigation in individual licensing proceedings, except as provided in 10 CFR 2.758.

Under the staff proposal, the Commission could certify the standardized ABWR design for referencing by applicants for a period of 10 years. Renewal of the design certification could be granted for an additional period of up to ten years unless the Comission found that the design would not comply with the Commission's then-current regulations. Applicants could reference the certified ABWR design in applications for cps and OLs docketed during the period beginning with the docketing date of the FDA application and ending t

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at the expiration date of the design certification. However, no CP or OL could be issued for an application referencing the ABWR design until the FDA is issued.

10.3 Completeness of Scope and Design Detail The ABWR application for design certification is to include a plant design that is essentially complete in both scope and level of detail. The scope of design developed to support the design certification process is addressed in Section 8.4.

The ABWR application for design certification also is to demonstrate compliance with the licensing criteria for new plant designs set forth in the Commission's Severe Accident Policy Statement (Section 7). The ABWR SAR is to address the tests, analyses, and inspections that are necessary to provide reasonable assurance that the plant will be built and operated within the specifications of the certified design. For those individual aspects of the design where safety-related structures and components differ from those of existing designs, empirical information is to be included as part of the application for design certification.

The ABWR SAR is to include information that will permit construction verification and compliance. This will permit reviews during the construction and startup phases of the plant and will eliminate the need for further design reviews on those portions of the plant that have been certified, except to verify that interface requirements have been met.

10.4 Changes to Approved and Certified Designs The staff believes that standardization will be best achieved if changes to approved or certified designs are kept to a minimum. Nevertheless, there are situations in which changes may be needed or desirable. It is the staff's intent that after issuance of the design certification, the Commission would require backfitting only when it determines, using the standards in 10 CFR 50.109 and the results of the DC rulemaking, that a substantial increase in the overall protection of the public health and safety would result.

GE may request modifications to an approved or certified design by applying for an amendment to the design approval or certification in accordance with the proposed regulations.

10.5 Rulemaking Appendix 0 to 10 CFR 50 provides the opportunity for the Commission to 1 approve the ABWR reference system design in a rulemaking proceeding. The  ;

regulations that are currently under development will specify the procedures to be used for the rulemaking. In general, however, upon receipt of.a request from GE, a notice would be published in the Federal Register announcing the '

request for a design certification for the ABWR. The notice would set out the i matters at issue, as specifically as possible, and the possible hearing procedures that could be used if the Commission decided to hold a hearing, and would request that all persons wishing to participate in a hearing notify the Commission within a stated period of time. As a condition to participating in  ;

a hearing, however, interveners could be required to state the issues they wish to have considered at the hearing and to commit to providing expert testimony 14

- on those issues. Written comments could be invited from those not intending to participate in the hearings. As a result of responses to the notice, or on its own initiative, the Commission could then hold hearings on the proposed rulemaking.

If a hearing were held, the Commission could use a number of formats, from the simple hearing and recording of testimony to interchanges among the parties and a limited right of cross-examination. Because such rulemaking procedures go beyond the notice-and-comment requirements for rulemaking, the Commission has broad discr;etion to establish hearing procedures best suited to the matters at issue.

After any hearing, the Commission could review the complete record of the rulemaking, including both the hearing record and any other written comments. The notice of final rulemaking would have to include responses to written comments and the resolution of issues considered at a hearing.

The views of the ACRS would be sought and considered. The ACRS would review the design before the rulemaking, and the results of the ACRS review would be made available when the proposed rule is announced.

10.5 Renewal of Certifications l Under.the staff's proposal, the Commission could certify the standardized ABWR design for referencing by applicants for a period of 10 years.

Additionally, before the expiration of the design certification, GE could apply for certification renewal. The design certification could be renewed for an additional 10-year period, provided the design complies with the Commission's regulations in effect at the time of the renewal application.

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APPENDIX A SITE ENVELOPE PARAMETERS AND SOIL-STRUCTURE INTERACTION ANALYSIS 1 SITE ENVELOPE PARAMETERS This Appendix addresses some of the important site envelope parameters proposed by GE for the ABWR. Additional parameters are to be provided in the SAR. All of the proposed parameters and analyses will be considered by the staff during the review.

Utility applicants must verify that their proposed facilities lie within the site envelope parameters assumed by GE in its safety analyses and approved by the staff. No further analysis or actions will be required by an applicant when the site-specific parameters fall within the design envelope. If site or interface parameters fall outside the design envelope, utility applicants must provide justification for the deviations.

2 SOIL-STRUCTUREINTERACTION(SSI) ANALYSIS 2.1 Scope GE will perform soil-structure interaction (SSI) analyses of the reactor building design for a rance of site conditions within the site envelope design parameters defined in Table A-1. GE will satisfy the staff acceptance criteria for the design ground motion and SSI analysis methods as specified in Revision 2 of the SRP, Sections 2.5.2, 3.7.1, and 3.7.2.

2.2 Methodology GE intends to employ the state-of-the-art computer program, SASSI (System for Analysis of Soil-Structure Interaction) code. SASSI is a linear analysis program using the finite element approach. Solutions for a complete soil-structure system are sought in the frequency domain employing the complex response technique. Problem formulations are based on the flexible volume substructuring method in that the complete soil-structure system is partitioned into the foundation and the structure.

In compliance with the duality (finite element and half-space) requirement, limited cases are to also be analyzed using the CLASSI/ASD computer code, which is an improved version of the CLASSI family of computer codes.

CLASSI/ASD is a linear analysis program using the structure approach based upon continuum mechanics for half-space.

A-1

_ _ _ _ _ _ _ _ . )

Table A-1 Envelope Of Selected Plant Site Design Parameters Applicable To ABWR MAXIMUM GROUND WATER LEVEL: 2 feet below grade PRECIPITATION (for roof design): .

1 Maximum rainfall rate: 10 in/hr Maximum snow load: 50 lb/sq ft j DESIGN TEMPERATURES:

Ambient: 1 1% Exceedance Values Maximum: 100'F dry bulb /77'F coincident wet bulb )

Minimum: -10'F l 0% Exceedance Values (historical limit)

Maximum: 115 F dry bulb /82*F coincident wet bulb Minimum: -40'F Peak Emergency Cooling Water Inlet: 95'F Condenser Cooling Water Inlet: 100*F SEISM 0 LOGY:

i SSE PGA: 0.30g*

SSE Response Spectra: per Regulatory Guide 1.60 SSE Time History: Envelope SSE Response Spectra EXTREME WIND: Basic Wird Speed: 110 mph / 130 mph S0IL PROPERTIES:

Minimumbearingcapacity(demand): 15 ksf Minimum shear wave velocity: 1000 fps Liquificntion potential: None at plant site resulting from OBE and SSE Free-field, at plant grade elevation. " ~

l **

50-year recurrence interval; value to be used for design of non-safety-relate structures only.

• • • t 100-year recurrence interval; value to be used for design of safety-related structures only.

sses Values of bearing capacity and shear wave velocity are included in this table to ensure wide application of a standard mat-type foundation design. The design must be evaluated parametrically against ranges of possible soil properties to verify wide application.

A-2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

2.3 Key Desian Parameters (1) Design Ground Motion For the purpose of standard plant design, GE plans to use a value of 0.3g for the peak ground acceleration (PGA) for the Safe Shutdown Earthquake (SSE) in accordance with Table A-1.

(2) Control Motion Location The. control motion (or' design ground motion) is to be defined at the finished grade in the free-field. This is consistent with' Revision 2-of the SRP, Section 3.7.1, and with the recommendations in h0 REG /CP-0054, " Proceedings of.the Workshop on Soil-Structure Interaction". June 1986, for the application of ground motion defined by Regulatory Guide 1.60 for standard design ground spectra.

(3) Design Ground Spectra The design ground spectra are to be the Regulatory Guide 1.60 standard design ground spectra, normalized to the design peak ground acceleration.

(4) Design Time History A single set of three artificial time histories (two horizontal components and one vertical component) are to be considered. They are to satisfy che spectra enveloping requirement of the Regulatory Guide 1.60 spectra. The power spectral density (PSD) function of the two horizontal components are to be calculated and compared with the target PSD specified in Revision 2 of SRP, Section 3.7.1.

(5) Soil Conditions To support the ABWR all-soil envelope design concept, GE plans to I analyze a reasonably large number of cases by considering the variations of soil-site conditions within the site envelopes of Table A-1.

The site conditions are to be selected based on those considered in the GESSAR II design and adjusted to accommodate the ABWR design. These site conditions cover a wide range of soil deposit depths, shear wave velocities, and water table locations.

A-3

(

l i

(6) Strain-Dependent Soil Properties GE plans to use strain-dependent soil properties for shear modulus and  !

material damping, as defined in the GESSAR II design. In accordance with NUREG/CR-1161, " Recommended Revisions to NRC Seismic Design Criteria",

May 1980, the strain-compatible shear modulus will be limited to 40% of its low-strain value and material damping will be limited to 15% of critical. The effects of pore pressure are to be taken into account by varying the water table location.

(7) Floor Response Spectra Peak Broadening The floor response spectra to be generated for subsystem seismic design are to be the enveloped spectra for a wide rarige of potential site conditions. The ur. certainties associated with soil properties, therefore, will be automatically accounted for in the enveloped spectra. The only other sources of uncertainty that may exist in the calculated floor response spectra are those associated with modelling approximations and structural material properties. To account for those uncertainties that may result in variations in structural frequencies, GE considers it sufficient to broaden the peaks at the structural frequencies of the enveloped computed floor response spectra by +10%. This is consistent with the GESSAR II design.

j i

A-4

.m

APPENDIX B INSTRUMENTATION AND CONTROLS I INTRODUCTION The instrumentation and control (I&C) systems of the ABWR are to use state-of-the-art fiber optics, multiplexing, and computer controls. Staff guidance in this area has not been developed, however GE has committed to the standards and criteria currently specified in the SRP, and use of the documents and criteria identified below. GE is to provide the information listed below so the staff can determine the acceptability of the ABWR I&C systems.

In lieu of actual test or qualification reports for equipment that will be selected later by GE or a utility applicant, GE is to provide a detailed description of the testing or qualification standards to be used to assure that the equipment that is ultimately selected will perfcrm as intended.

2 MULTIPLEXING SYSTEMS In the SAR, GE is to:

(1) Provide a complete list of components (pumps, valves, etc.)

whose actuation, interlock, or status indication is dependent on the proper operation of each Class IE multiplexer.

(2) For the components cited above, describe the means of remote or local control (other than by cutting wires or jumpering) that may be employed should the multiplexer fail.

(3) Describe the multiplexer pre-operational test program.

(4) Describe the test and/or hardware features employed to demonstrate fault tolerance to electro-magnetic interference.

(5) Describe the interconnection, if any, of any Class IE multiplexer to non-Class 1E devices such as the plant computer.

(6) Describe the online test and/or diagnostic features that may be employed, including any operator alarms / indicators and their locations.  ;

(7) Describe the multiplexer power sources.

(8) Describe the dynamic response of the multiplexer to momentary interruptions of AC power.

(9) Describe the applicability of the plant Technical Specifications to multiplexer operability.

(10) Describe the hardware architecture of all multiplexer units.

B-1

.~

l (11) Describe the "firmware" architecture.

~

(12) Provide an explicit discussion of how the systems conform to the provisions of IEEE-279, Section 4.17.

(13) Provide an explicit discussion of how the systems conform to IEEE-279, paragraph 4.7.2, as supplemented by Regulatory Guide 1.75 and IEEE-384.

(14) Provide confirmation that system level failures of. any multiplexer  !

system detected by automated diagnostic techniques are indicated  !

to the operators consistent with Regulatory Guide 1.47.

i (15)Provideanexplicitdiscussionofthesusceptibilityofthe l multiplexer systems to electromagnetic interference, i 3 ELECTRICAL ISOLATOP.S 4

GE has committed to provide the following on isolation devices:

(1) For each type of device used to accomplish electrical isolation, a '

description of the testing to be performed to demonstrate that the device is acceptable for its application (s). The test configuration and how the maximum credible faults applied to the devices will be included in the description.

3 (2) Identification of the data that will be used to verify that the maximum credible faults applied during the test are the maximum voltage / current to which the device could be exposed, and to define how the maximum voltage /

current is determined.

(3) Identification of the data that will be used to verify that the maximum credible fault is applied to the output of the device in the transverse mode (between signal and return) and other faults are considered (i.e.,

openandshortcircuits).

(4) A definition of the pass / fail acceptance criteria for each type of device.

(5) A comitment that the isolation devices will comply with all environmental  ;

qualification and seismic qualification requirements.  ;

l (6) A description of the measures taken to protect the safety systems from electricel interference (i.e., electrostatic coupling, EMI, common mode, and crosstalk) that may be generated.

(7) Information to verify that the Class IE isolation devices ~

are powered from a Class IE power source (s). ,

B-2 l t

1

. (8) A' comparison of the design with the guidance in NUREG/CR-3453/ EGG-2444,

" Electronic Isolators Used in Safety Systems of U.S. Nuclear Power Plants," March 1986.

(9) A comparison of the design with the guidance in draft Regulatory Guide EE502-4, " Criteria for Electrical Isolation Devices Used in Safety Systems for Nuclear Power Plants".

4 FIBER OPTIC CABLE The staff is working with EG&G to develop comprehensive guidance on this subject. The guidance will be based on the existing IEEE cable standards, such as IEEE-323 and IEEE-384, on the ANSI standards for fiber optic cables which are listed at the end of this Appendix), and the results of the EG&G work.

5 PROGRAMMABLE DIGITAL COMPUTER SOFTWARE As a starting point, the following documentation is to be used by GE in the design and by the staff in its review:

(1) ANSI /IEEE-ANS-7.4.3.2, " Application Criteria for Programmable d Digital Computer Systems in Safety Systems of Nuclear Power Generating Stations," 1982  ;

(2) Regulatory Guide 1.152, " Criteria for Programmable Digital Computer System Software in Safety-Related Systems of Nuclear Power Plants," November 1985 (3) NUREG-0308, " Safety Evaluation Report - Arkansas Nuclear 1 Unit 2," November 1977 (4) NUREG-0493. "A Defense-in-Depth and Diversity Assessment of the RESAR-414 Integrated i'rotection System," May 1985 j (5) NUREG-0491, " Safety Evaluation Report of RESAR-414," February 1979 6 PROGRAMMABLE DIGITAL COMPUTER HARDWARE As a starting point, the following documentation is to be used by GE in the )

design and by the staff in its review:

(1) IEEE 603, "IEEE Standard Criteria for Safety Systems for Nuclear l Power Generating Stations," 1980 ,

i i

(2) NUREG-0308, " Safety Evaluation Report - Arkansas Nuclear 1, Unit 2,"

November 1977 (3) Regulatory Guide 1.153, " Criteria for Power, Instrumentation; and Control Portions of Safety Systems," December 1985 l (4) NUREG-0493, "A Defense-in-Depth and Diversity Assessment of the RESAR-414 Integrated Protection System," May 1985 (5) NUREG-0491, " Safety Evaluation Report of RESAR-414," February 1979 B-3

, ANSI Standards on Fiber Optics FIBER OPTICS pu, Opa cwt o; continuities, Ax58/ Spectret Artemat,on W wrement w Acd'ersted Oryten A#ns. AN5f/UA RS DA R545k321983.15D0 lan44*arth, cred.d% opt,c.: /;

45 b10,1 19 84. 5 7.00 fiber Optic Corontor Dust (Gne Sand)

Acceleration Tecing of Componeats and bers ANSI /UA R545546-1983 Test, ANSI /DA A5 455351981, 85.00 Auembl;es, AN5VtlA 45k18 A 19 64, $ 5.00 Temperatum Dependence of Artenu.oon

$5h0 6ber Optk Shock Test (Specified Pulse), for Optical'Wavegulde fibers. Wthod

• Air tmitage Testing of 6ber Op5c Cormo- AN5t/UA R545514-1983.19.00 lor Messuring. AN5t/UA RF45552 ensnt Soln AN5f/[l A 4 5523 A 198 5, 71oid immersion Test for fber Optic Ca. 1983,58.00 55.00 tJes. AN5t/DA R5455401983, ye,minal Devices. Gene,ic 5pecjf.cz tion Ahitude immersion. ANSI /[!A R545 kt b $ 8,00 for, ANSUDA R55091984. 510.00 1983.55.00 Cas flame Test for fiber Optic Cable.

Ambient OsM SusceptibEty, AN51/ eta RS- ANSI /EIA R545599-1983,55.00 Test Procedure 3 ANSI /CfA RS-4551

^ 3,g 3, y 3 o.co 455221983.57.00 Cas Leakage Test for Eber OpR Cable, Test Precedures for Fiber Optk Rbers, Ca.

+6end Test for 6bw Optic Cable T 'st, AN5t/DA R5455100-1984. 57DO AN51/[l/ 455911985,4 impact Testing Measurements for De4ces, gg, y,%,,,, g g ANSI / CIA R54552A 1954. 55ho mbating Devices AN51/ CIA RS455-Ikiftered fLer Bend Test AN51/ttA RS . 1980,313 00 4551031984,55.00 *intenonnnGen Doke Inser%n Less Tot

• Catde to interconnecting De.ke Asiat ANSI /UA 4553t.1985,510 Test Procedures 12. 2 7, 28. and 2 9 sive Leading. ANSI / IIA RS E ngst ano us or WMR 554N 5N E8 198 A $ 5.00 Test Procsdures 17. 2 5. and 31, ANSI /

• • * *** 1983 5500 DA RS4553-198 0(R198 5), 513.00 tic, AN51/UA R5472,1985. $ 18.00 .la t selfifhesion (Bloc **.ng.) Test for 6- eTest Procedures 26 (Crush Resistance),

bw Optic Cable. ANSI /UA RS45584- AN5UCA R54 5 526 A 198 5, 5 5.00

• Communication Cables for hdoor Use' 1984 5500 " Test Procedures 30,33,36, 5 4. 55 and A 1 85 58 JacAet n age Test b % OpCc Ca. 82, MW RWL1982, $ 1h kab Caws & We Avia! Id'* AN5t/EIA RS 455861983,55.00 Torsb Test for Optical Waveguide 6bers, US** Set,onal Spx.4caUon & N -

Neot Test for Aber Optic Cable. AN5UCA AN51/UA R54 5 b65-19 84, 5 7.00 i

Optic, ANSI /U A 472 A 1985. 5 8 00 RS455 871983, 58.00 Wual and Mechanicalinspect.on of nbers-g g g,g 3 gg g

1. Commsmication Cables for Outside Tele.

phone Plant Use, Sectionat Sonific.s- Creded.inder Opticat Aber spectral AL. Opt;c Devices, ANSL/UA Rb45513 tions for 6ber Optic, AM.51/t!A 472 D- tenuation Measurements. ANSI /UA RL 1984.15.00 i 455501933,58.00 1985.58 M Maung Durability for Aber OpGe Intercon-

• CommunicaGon Cables W Uruserground cad Buried use. 5ectional 5pecil. cat;ons riec6ag Devices. ANSUDA 45521 tor 6ber Optic. AN5UEIA 472 B 1985 1984.55.00 Wasurement of Change in Optical Trans. i' 38M mittance, ANSUELA RS455201383

• Compound now (orip) Test for Glied 6ber 58.00 Optic Cable, ANSI /W. 455811985, Measuring RstaWe Abrasion Redstance of 55.00 - Opt; cal Weveguide Coatings oms Buf.

I eComprusive Loading Redstance of Fee' lers, AN5l/DA RS455461984, )

- Op6c Cables, ANSI /DA R545541' .

57.00 1985.A On-Une D;amator Wasurement of Optical Connector Terminology, nber Optic, ANSI / wanguides. AN58,11A RS-45b48-DA R54401978, 5 5.00 1983,55.00

• Connector / Component Temperature Ufe, + Optical Crosstalk in 6ber Optic Compo.

AN51/DA 455 4 A.1985. 55.00 rients, AN58/DA 45k421955. 36.00 Connectors, Generic Specif. cation lor. Optical Weveguide Aber Matedal Classes AN5UElA ftL4751981. 513.00 1 and Preferred Sites, ANSUDA RS 454 A 19 84. 5 7.00 J Cde Osmet., y*4urement of Creded b **~

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• • 8 O 6 cal F6ers. AN5uu4 RS 455

• Opt. cal Weveguide F6ers, Conc 6c Speen;.

58 I m. 5740

• C^'sh Redstance of Abw Optic intercon. cation for. ANSI /OA 492 1985*

514.00 .

es. AN51/UA 45526A. Output far-Field Rad;ation Pottem Was.

g%s of Tcm1 M8dng to fiber Op- prement ANSI /UA Rb455471983*

$ 10.00 C, 812 1984, 57,0o, y p Output Naar netd Radiatiorr Pa rtem Was.

Ws.eguide reer . w- mt f Op6 cal Waveguide reers, 561983 33,00' A $[ A 4Sk AN5UUA 45b431984. 55.00 Lasemal ft$1 Opuc Cab's t* " '* * * *** N" * ' **d;* M**

Msn r Opuc Components, he 55.00Test, Ansyng g3435 80'III3'

. AN51/DA R5445-4S 1983. 58 00 Putse Distort.on Measurement of Mag.

Fiber Geometry 6ers, Microsoep;c Athadofso, Optical wa Wev, vide [F-ss Optical 6ber informason

• • d' Cla l

%. AN5yDA RS-45545-19$4, Tran sm.iss.on Capechy. AN5UUA R5

.58.00 45551 1983,31040 .

,r6er Optic Cable Seq 74 g g Refract;.e andes pr,r,se. Refracted Ray )

Temperature. AN5yc4 R5455 37 Method. ANSI /DA R5455-441984, 1983.55.00 59.00  !

l A Opuc cable Twist Test, AN5yuA 5ett Speer (Conook) Test ter compo.

5 b8 5-1954, sypo  !

1. =ents. AN5VDA RS45516-1984* I A Opuc Cable Wieseng T.eg, Ansaft:A 3740

1. 5455-39-1983,55.00 - {

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