ML20149M218

Draft SER on Chapter 1 of EPRI Advanced LWR Util Requirements Documents

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s ORAFT SAFETY EVALUATION'REPOR_T on CHAPTER 1 0F EPRI's ADVANCED LIGHT WATER REACTOR UTILITY REQUIREMENTS DOCUMENT prepared by the Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ,

September 1987 (as revised on February 5, 1988)

. (with changes on pages 1-6, 2-2 ,

) 4-10 and 4-10a indicated by vertical lines in margin) 8802250417 080218~

PDR PROJ 669A PDR- 1

e i .

4 l TABLE OF CONTENTS ,

Pagg 1 INTRODUCTION ..................................................... 1-1 1.1 The ALWR Program ............................................ 1-1 1.2 The ALWR Utility Requirements Document ...................... 1-2 1.3 The Staff Review ............................................ 1-4 2 GENERAL DESIGN REQUIREMENTS ...................................... 2-1 3 DESIGN-BASIS EVENTS .............................................. 3-1 4 STRUCTURAL DESIGN BASIS .. ....................................... 4-1 5 MATERIALS ........................................................ 5 6 RELIABILITY AND AVAILABILITY ..................................... 6-1 7 CONSTRUCTABILITY ................................................. 7-1 8 OPERABILITY AND MAINTAINABILITY .................................. 8-1 9 QUALITY ASSURANCE ................................................ 9-1 10 LICENSING ........................................................ 10-l' CONCLUSIONS .......................................................... C-1 REFERENCES ............................................ ............... R-1 CH 1, TC 09/13/87

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l 1 INTRODUCTION ,

The purpose of this Draft Safety Evaluation Report (DSER) is to document the results of the NRC staff's initial review of Chapter 1 of the Advanced Light Water Reactor (ALWR) Utility Requirements Document. That document is a compre- l hensive statement of utility industry requirements (performance specifications) for the utilities and their plant designers to use in developing future light water reactor (LWR) power plant projects. The basic objectives addressed by  ;

the Requirements Document are: (1) increased public safety, (2) enhanced plant l performance, (3) increased protection of plant investment, (4) reduced cost, and (5) assurance of licensability. The ALWR Program proposes to meet the j first three objectives through the development of more conservative designs using proven technology. The cost objective is to be met by simplifying plant i design, with an emphasis on operability and maintainability, and adhering to a 6 year schedule for site-specific engineering, construction, and startup  ;

efforts. The licensability objective is to be met through NRC approval of the l Requirements Document, which is intended to satisfy regulatory requirements and appropriately treat the applicable generic safety and licensing issues.

1.1 The ALWR Program  !

i The ALWR Utility Requirements Document is a primary work product of the ALWR ,

Program, which was initiated by the Electric Power Research Institute (EPRI) to provide leadership for utility industry efforts to ensure that there is a  ;

viable nuclear power option in the United States. The Program is directed by EPRI's ALWR Utility Steering Committee.* Those participating in the program include utilities with nuclear plant experience, nuclear steam supply system l (NSSS) vendors, architect-engineering firms, and consultants in related fields.

• For convenience, the term "EPRI" is generally used in this document to designate EPRI and/or its ALWR Utility Steering Committee.

CH 1, SEC 1 1-1 09/13/87

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As part of the ALWR Program, in 1983 EPRI embarked on a major effort to  !

stabilize the regulatory climate by working with the NRC staff to identify and resolve key safety and licensing issues. That joint effort resulted in a process whereby the unresolved and generic safety issues applicable to the ALWR as of July 1, 1986 were identified. This process was consistent with the procedures given in NUREG-0933, "A Prioritization of Generic Safety Issues."

EPRI's proposed resolutions of these issues are included in the Requirements Document and in separate topic papers that provide the bases for those I resolutions.

With respect to issues identified after July 1, 1986, EPRI plans to establish appropriate requirements after such issues are resolved if one or more of the  ;

following criteria are satisfied in the affirmative:

(1) Would the core melt frequency goal established in the ALWR Requirements Document be exceeded as a result of this issue? l (2) Would the offsite accident radiological consequences dose criteria l established in the Requiremants Document be exceeded as a result of '

this issue? .

(3) Would the Commission's safety goals be exceeded as a result of this i issue?

In addition, EPRI has determined that a number of technically supportable alter-natives to some current regulatory requirements are needed in order to achieve the program goals. These alternatives, called "plant optimization subjects,"

are also included in the Requirements Document, and further details are provided i in separate subject papers. (Further information about the regulatory phase of this program is available in NUREG-1197," Advanced Light Water Reactor Program - Program Management and Staff Review Methodology.") l In 1985, two new phases were added to the ALWR Program: the development of the Utility Requirements Document, which is addressed below, and the assessment of  ;

small plant options, which is not part of this review.

CH 1, SEC 1 1-2 09/13/87 I l

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1.2 The ALWR Utility Requirements Document The primary emphasis of the Requirements Document is on resolution of signifi-cant problems experienced at existing nuclear power plants. The document is to be used with companion documents, such as utility procurement specifications, which would cover the remaining technical requirements applicable to new plant projects. The Requirements Document consists of two parts, as follows:

Part I, the Executive Summary, presents a concise discussion of the design objectives and philosophy of ALWRs, the overall configuration and features of ALWR plants, and the steps necessary to proceed from a conceptual design to a functioning plant. This part was published by EPRI in June 1986.

Part II is the set of utility industry-developed requirements for the design, construction, and performance of an entire ALWR power plant--either a boiling water reactor (BWR) or a pressurized water reactor (PWR)--that ranges in size up to 1350 MWe. This part will consist of 13 chapters, beginning with overall plant design requirements (Chapter 1) and proceeding through detailed requirements for each major portion of the plant. The principal subject of each chapter is indicated by its title in Table 1.

Table 1 ALWR Utility Requirements Document Chapters CHAPTER 1 -

OVERALL REQUIREMENTS CHAPTER 2 -

POWER GENERATION SYSTEMS CHAPTER 3 -

REACTOR COOLANT SYSTEM AND NON-SAFETY AUXILIARY SYSTEMS CHAPTER 4 -

REACTOR SYSTEMS CHAPTER 5 -

ENGINEERED SAFEGUARDS CHAPTER 6 BUILDING AND ARRANGEMENTS CHAPTER 7 -

FUELING AND REFUELING '

CHAPTER 8 -

PLANT COOLING WATER SYSTEMS CHAPTER 9 -

SITE SUPPORT SYSTEMS CHAPTER 10 - PLANT CONTROLS AND INSTRUMEN-TATION CHAPTER 11 - ELECTRIC POWER SYSTEMS CHAPTER 12 - RA910 ACTIVE WASTE PROCESSING SYSTEMS  !

CHAPTER 13 - TURBINE GENERATOR SYSTEMS l l

CH 1, SEC 1 1-3 09/13/87

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e Chapter 1 of the Requirements Document was submitted for NRC review by the ALWR 1

Steering Committee Chairman on July 8, 1986. Chapter 2 was received in October i 1986, and Chapters 3 and 4 were received in July 1987. The subsequent chapters i are scheduled for submittal by EPRI at various times throughout 1988. EPRI l also plans to make changes in the 13 chapters as a result of comments by the NRC and others, and will submit them about mid-1990 for final NRC review.

l'. 3 The Staff Review The NRC staff plans to issue a draft SER on each chapter of the ALWR Utility Requirements Document between now and September 1989. This schedule is based on the assumption that 6 to 8 months will be sufficient for the review of each chapter beginning with C'h apter 3. Allowing a similar period for review of EPRI's "rollup document" containing modifications of the 13 chapters, the staff presently expects to issue the final SER on the entire Requirements Document by December 1990. Before completing Chapter 1, EPRI had submitted six topic papers containing the elements of resolution for seven unresolved I generic safety issues and a report describing technically supportable alternatives ("optimization subjects") to current regulatory requirements in five areas that would be incorporated into Chapter 1. The staff responded to those submittals by letters dated January 24 and June 6, 1986, providing comments that EPRI considered in the preparation of Chapter 1. The subjects covered were the following:

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- Generic Safety Issues j A-17 Systec.s Interaction I.F.1 Quality Assurance - Expanded QA List l l

II.F.5 Classification of Instrumentation, Control and Electrical Equipment II.B.8 Safety Review Consideration - Rulemaking Proceeding on Degraded Core Accidents II.E.6.1 In Situ Testing of Valves - Test Adequacy Study j 21 Vibra. ion Qualification of Equipment HF.02.1 Maintenance and Surveillance Program i

CH 1, SEC 1 1-4 09/13/87

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- Regulatory Optimization Subjects (alternatives to current NRC requirements)  !

Leak-Before-Break Methodology Operating Basis Earthqi:akes and Dynamic Analysis Methods ,

Source Term Equipment Seismic Qualification by Experience Tornado Design The NRC staff requested additional information relative to the review of Chapter 1 by letters to the ALWR Utility Steering Committee Chairman dated January 5 and March 11 and 18, 1987. Responses to all three requests were provided by a single letter dated March 27, 1987.

As EPRI and its Steering Committee requested, the staff has endeavored to review the Chapter 1 requirements at the level of detail presented. The level varies widely. For example, the material requirements in Section 5 specify some materials in great detail, while the effects of water chemistry are hardly mentioned. The requirements also vary considerably in the effects they have on plant design and operation. Some--such as the use of specific materials--are of considerable importance, whereas others are so trivial that  ;

they could be replaced by a simple statement such as "good engineering practice should be followed." This unevenness in treatment probably results largely from the ALWR Program's intentional emphasis on resolution of problems experiencsd in existing plants instead of on the development of a complete compendium of design requirements.

The staff used the NRC Standard Review "lan (SRP, NUREG-0800) as guidance for its review; however, the SRP was written for review of safety analysis reports on specific plant designs. Reviewing utility requirements (performance specifica-tions) that will guide the plant designer is a different ~ kind of exercise, I because the degree of compliance with some regulatory requirements cannot ce j completely determined without a detailed plant design, i l

l CH 1, SEC 1 1- 5 09/13/87 j I

Based on the above considerations and in accordance with the ALWR Program, the staff's evaluation has become one of "review by exception." In this mode, the staff has assumed that all current regulatory requirements would be met by an ALWR plant except where deviations are identified in the Requirements Document or where the staff identifies (1) a potential incompatibility between proposed design requirements and current regulatory requirements or (2) an apparent misinterpretation of the regulatory requirements. In addition to addressing such matters relative to safety, the staff also has provided constructive comments on the document that, while not specifically regulatory in nature, would offer improvements in its requirements. Review of an actual ALWR plant by the staff will be performed in accordance with the most current version of the SRP, and the staff will follow the SRP criteria except for those instances where the staff has specifically accepted other positions indicated in the ALWR Requirements Document and those positions have been endorsed in the final SER for the ALWR Program.

The staff's evaluation of Chapter 1 is documented in the following sections of this Draft Safety Evaluation Report (DSER). The sections are numbered and titled like those in the Requirements Document, except for Section 2, which i addresses all of the general design requirements.

Copies of this report are available for inspection at the NRC Public Document Room, 1717 H Street N.W., Washington, D.C. i The NRC project manager for the staff's review of the ALWR Utility Requirements Document is Paul H. Leech. He may be contacted by calling (301) 492-8209 or by writing to the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. l l

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CH 1, SEC 1 1-6 02/05/88 l

2 GENERAL DESIGN REQUIREMENTS

• Sections 1.4 and 2.0 of Chapter 1 present the utility general design require-ments to be met by an ALWR plant. Those that address public safety and plant protection, plant characteristics and performance, plant cost, and regulatory stabilization are in Section 1.4 (Table 2 of this DSER presents a summary of selected general design requirements related to safety). The utility require-ments in Section 2.0 cover the design process, the use of proven technology, design standardization and simplification, plant investment protection, and design bases.

The NRC staff has reviewed these sections and, with certain exceptions that are indicated below, has determined that they are acceptable or do iiot warrant regulatory review, Also stated below are several changes proposed by EPRI and EPRI responses to staff comments. t i

Section 1.4.A.1, "Public Safety" The utility requirement in this section states, in part: "In the event of a severe accident, the dose beyond a half mile radius from the reactor shall not exceed 25 rem." In response to a staff comment that the requirement should be

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clarified, EPRI stated that it will be revised to read: "In the event of a severe accident, the dose beyond a half mile radius from any individual reactor shall not exceed 25 rem whole body."

The staff also requested the basis for a statement in the Engineering Rationale discussion to the effect that a plant trat satisfies the 25-rem requirement for accidents of > 10.s/yr frequency of occurrence is permitted to have a simpler evacuation plan. EPRI responded that "it was intended in the development of this conservative design requirement that any emergency plans considered appro-priate would be based on the recognition of this additional levil of protection provided by the ALWR design. Recognition of the ALWR dose requirement could be Cli 1, SEC 2 2-1 09/13/87

. .s e, Table 2 Summary of selected ALWR program general design requirements

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General Requirements Numerical Value l E

na 1. Plant Safety and Protection

- Public Safety - Severe Accident Radiation Dose < 25 rem (I)

- Plant Investment Protection - Core Damage Frequency Target < 1 x 10 5/yr

2. Plant Performance

- Availability > 87%-

- Planned Outage Time < 25 days /yr (average)

- Refueling Interval 24 month capability 7 - Refueling Time < 17 days -

ro

- Unplanned Automatic Scrams < 1/yr

- Plant Design Life 60 years

- Radioactive Waste Shipped (Low Level) < _ 2500 ft3 /yr/ unit

- Plant Construction Schedule 72 months (2)

? - Plant Personnel Exposure < 100 man-res/yr (1) For accidents more frequent than 1 x 10 8/ year at distance greater than half a mile from the reactor.

(2) From owner commitment to construct to commercial operation.

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• Source: Table 1-1, Chapter 1, ALWR Utility Requirements Document

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4 6 used as the basis for developing a more balanced emergency plan which reduces the emphasis on evacuation and considers other alternatives consistent with the plant design philosophy." The staff believes that such a proposal will require careful review since n may involve a change from current policy in that the proposal appears to state that additional engineered safety features in the ALWR design can justify modifications to existing emergency planning requirements and .

practices.

Section 1.4.B.1, "Availability" The staff suggested that this requirement should be clarified to indicate l whether 60 year plant life means 60 years at full power or 60 years at an 87%

availaoility rate. In response, EPRI said this requirement will be modified to state: "The plant shall be designed to operate for 60 years at an 87 percent ,

availability, which corresponds to an expected capacity factor of 83 percent."

The Atomic Energy Act currently limits the operating period to a maximum of 40 years; however; a study of the ramifications of authorizing 60 years of r operation was recently initiated by the staff. l Section 2.2, "Desian Process" The Engineering Rationale for this section does not include Section 11 of the

{

Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME Code) related to Inservice Inspection. In response to this observation, l EPRI has agreed to include an appropriate reference to that code section in a future revision of Chapter 1.

Sections 2.2.F.3 and 2.2.F.4, "Desian Integration" These sections discuss the development of "living" computer models for design studies and plant performance simulation. The models are to be maintamed as a permanent part of the plant documentation. Although the subject of probabilis-tic risk assessment (PRA) has not yet been discussed in any detail, it is understood that, in compliance with the Commission's Severe Accident Policy, CH 1, SEC 2 2-3 09/13/87

PRAs will be an important part of the ALWR Program to establish the risk level associated with the design (s). It is also expected that PRA methods will be utilized in certain decisionmaking. Although the development of overall plant PRAs may not be feasible until a subsequent phase of the ALWR Programs the staff suggested that the use of "living" PRA models in development of the overall plant design be included in the discussion in these sections.

EPRI agreed with this comment and stated that "it is the intent of the ALWR Program that 3 top-level PRA be used during the design phase in order to assist the designer in decisionmaking." A requirement to that effect will be added to the Requirements Document. The specific requirements for addressing severe accidents will be provided in Chapter 5, "Engineered Safeguards."

Section 2.2.F.7, "Design Integration" The second item in this section requires spatial separation of systems and i equipment to reduce common-cause failures due to seismic, fire, pipe rupture, and ficoding events. In response to a staff suggestion, EPRI has stated that failures caused by sabotage will be added to this list of initiating events.  !

Specific requirements defining acceptible separation of systems and components will be included in Chapters 5 and 6 of the Requirements Document.

As suggested by the staff, EPRI has stated that a requirement will be added to  !

Section 2.2.F.7 on protection against and mitigation of internal flooding.

Section 2.3.C, "Plant Siting," and Table 2-1, "Envelope of ALWR Plant Site Design Parameters" The envelope design parameters in Tat,le 2-1 are intended to allow siting at  ;

most sites available in the United States, but they do not encompass all worst-case conditions, such as the high seismicity along the Pacific Coast, i For corvenience of reference, Table 2-1 of the Requirements Document has been reproduced on the next two pages. The staff's preliminary evaluation of these pararceters follows the table.

1 CH 1, SEC 2 2-4 09/13/87

Table 2-1

• ENVELOPE OF ALWR PLANT SITE DESIGN PARAMETERS II}

MAXIMUM GROUND WATER LEVEL: 2 feet below grade EXTREME WIND: Basic Wind Speed: 110 mph ( /130 mph I4I 1 foot below TORNADO:(5)

MAXIMUM FLOOD (OR TSUNAMI) LEVEL:

to 26 feet above grade level - Maximum tornado wind speed: 260 mph

- Translational velocity: 57 mph PRECIPITATION (FOR ROOF DESIGN):

- Radius: 453 ft.

10 in/hr - Maximum atm AP: 1.46 psid

- Maximum rainfall rate:

- Maximum snow load: 50 lb/sq. ft. - Missile Spectra: per ANSI /ANS-2.3-1983 .

DESIGN TEMPERATURES: 501L PROPERTIES:(6)

- Ambient - Minimum Bearing Capacity (demand): s15 ksf -

1% Exceedance Values Maximum : 100*F dry bulb /77'F coincident wet bulb - Minimum Shear Wave Velocity: 51000 fps Minie.mn : -10*F - Liquification Potential: None 0% Exceedance Values (historical limit) (at Site-Specific SSE Level)

Maximum : .115*F dry bulb /82*F coincident wet bulb Minimum : -40*F

- Emergency Cooling Water inlet: 95*F

- Condenser Cooling Water inlet: <100*F .

3-lhis table is a reproduction from Chapter l of the ALWR Utility Reguirements Document

Toble 2-1 (Continued)

SEISMOLOGY:

- OBE Peak Ground Acceleration (PGA)( ): 0.10g(0}

- SSE PGAII}: 0.30g(8 M

- SSE Response Spectra: per Reg. Guide 1.60

- SSE Time History: Envelope SSE Response Spectra (1) Further definition on application of site design parameters may be found in subsequent chapters covering various areas of design. For example, Chapter 6 - Building design and plant arrangements ~

Chapter 8 - Cooling water systems 7 Chapter 9 - HVAC

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I2I 50-year recurrence interval; value to be utilized for design of non-safety-related structures only.

(3) 100-year recurrence interval; value to be utilized for design for safety-related structures only.

(4) Probable maximum flood level (PMF), as defined in ANSI /ANS-2.8-1983 "Determining Design Basis Flooding at Power Reactor Sites." Minimum value to be basis of standard plant design with provisions as defined in-Chapter 6 for accommodation of flood levels up to maximum value.

Ib} 1,000,000-year tornado recurrence interval, with associated parameters based on ANSI /ANS-2.3 - 1983.

Pressure effects associated with potential offsite explosions are assumed to be non-controlling for the design.

(6) Values of bearing capacity and shear wave velocity are included in this table to assure wide application of a standard mat-type foundation design. Design must be evaluated parametrically.agair.st ranges of possible soil properties to verify wide application. .

I#I PGA = Peak Ground Acceleration.

IO Free-field, at plant grade elevation.

I9I Envelopes all present U.S. nuclear sites, except those on California coastline.

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Maximum Ground Water Level: 2 feet below grade This criterien should encompass most potential nuclear power plant sites except borderline swampland sites or sites with artificially controlled groundwater levels.

l Flood (or Tsunami) Level: 1 foot below to 26 feet above grade level 1

l Although this utility requirement does not violate a regulatory requirement, the staff recommends that it should state only an upper limit related to plant grade. Assuming that this design level includes static water level plus wind waves, then the upper limit should be plant grade or less.

Precipitation (for roof design):

If tne maximum rainfall rate of 10 inches / hour is intended as a design value for sizing scuppers in parapets, it is much too low. For the Great Lakes area, the 1-square-mile, 5-minute probable maximum precipitation (pmp) is 6.3 inches, or 75.5 inches /hr. Along the Gulf Coast, the 1-square-mile, 5-minute pmp is 6.2 inches or 74.5 inches /hr.

The maximum snow load of 50 lb/ft2 may limit sites to below about 38 north latitude in some regions. For example, the Beaver Valley plant design is for 69 lb/ft2 snowload and 72 lt/ft2 snow and ice.

Design Temperatures:

The staff is not certata how the ambient temperature values would be used. For purposes of reviewing the safety-related water supply, the staff uses worst 1-hr, 24-hr, and 30-day values of record.

Extreme Winds:

The criteria meet NRC requireaents.

CH 1, SEC 2 2-7 09/13/87

9 Tcrnado The criteria in the table are different than the NRC requirements spelled out in WASH-1300 and used in SRP reviews. Accordingly, the maximum velocities should be, 290 mph rotational and 70 mph translational, and the maximum pressure drop 3 psi with a 2 psi /sec maximum rate of pressure drop. The staff has not endorsed American National Standards Institute /American Nuclear Society (ANSI /ANS)

Standard 2.3-1983, which is referenced in Table 2-1, as an adequate basis for changing its criteria. However, the staff is considering a modification of them based on tornado statistics in NUREG/CR-4461, "Tornado Climatology of the Contiguous United States." (This effort is related to ALWR optimization subject 3.C.2.) '

Soil Properties:

The i signs in front of the criteria numbers appear to be reversed. If this correction is made, these criteria meet regulatory requirements.

Seismology:

1 The parameters in the table indicate a 1 to 3 ratio between the operating basis. l earthquake (OBE) peak ground acceleration (PGA) and the PGA for the safe shut-down earthquake (SSE), whereas 10 CFR 100, Appendix A requires the relationship to be at least 1 to 2. The staff is considering whether the OBE should be redefined, but no decision has been made. (This is related to ALWR optimiza-tion subject 3.A.2.)

CH 1, SEC 2 2-8 09/13/87 i

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3 DESIGN-BASIS EVENTS ,

Section 3 of Chapter 1 defines the design-basis events, analysis processes, and acceptance criteria. The introduction to this section states that the analyses of safety-related events shall meet NRC requirements except where explicitly noted. On that basis, the staff has reviewed this section and found its utility requirements acceptable, except as indicated otherwise below.

Section 3.3.A.3, "Safety Analysis Events" The Engineering Rationale for this section states that not every postulated initiating event must be analyzed. The staff agrees with this philosophy but noted that it will be necessary to justify and document the choice of certain events to be analyzed. EPRI has stated that a requirement to this effect will be added to this section.

In response to a staff query, EPRI has defined "coincident occurrences" as  !

postulated single failures in addition to the initiating event. These single f failures are to be chosen so as to bound the consequences of all such possible ~ l single failures during the events analyzed as part of the safety analysis. l These coincident occurrences are listed in Tables 3-2 and 3-3 of Chapter 1 anJ j are based on requirements that are to be included in other chapters of the docu-ment. Hence, the treatment of such occurrences must be judged during the staff's review of those chapters.

Table 3-1, "ALWR Event Initiators and Frequency Categorization" In response to staff comments questioning the frequency categories assigned to )

various events, epi!I stated that these assignments are based on an assessment of current operational data and of the system quality and features that will be required in other chapters of the Requirements Document. Therefore, the staff review of these event frequency classifications will not be completed until that information is available.

CH 1, SEC 3 3-1 09/13/87

8 e EPRI has explained that the use of 10 CFR 20 in the frequency category column for radioactive releases from failed components or subsystems is not intended to establish their frequency but to indicate the event acceptance limits if the components are not designed as safety related. If the components.are designed as safety related, the limits applicable to a limiting fault will be used. EPRI also said that a clarifying note to this effect will be added to this table. On that basis, the staff concluded that this item does not c'onflict with regulatory requirements.

Table 3-2, "ALWR Coincident Occurrences for PWRs" Table 3-3, "ALWR Coincident Occurrences for BWRs" Station blackout is currently listed only in Table 3-8, "Utility Investment Protection Analyses for an ALWR." However, because of the importance of this issue, the staff believes it should also be listed in Tables 3-2 and 3-3 to ensure that the plant designer considers it in determining the most limiting event for analysis. EPRI has stated that, when this issue has been resolved for the ALWR, as intended during review of Chapter 5, the station blackout requirements will be reevaluated.

In its request for additional information, the staff called attention to various types of events that are not listed in Table 3-2 and/or Table 3-3; EPRI responded that these events are covered in Table 3-1. This fact should be recognized in Section 3.3.A.3, which requires the plant designer to determine the most limiting event for analysis, i l

l Table 3-6, "ALWR Performance Event Capabilities" Sections 6.b, 6.d, and 8 identify certain plar.t performance capabilities involving step and ramp power changes and inadvertent control rod insertion without reactor trip. Although such activities may not impose significant challenges to fuel integrity, the staff will require that confirmatory analyses be provided to ensure this is the case in specific plant designs. ,

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CH 1, SEC 3 3-2 09/13/87 l l

4 STRUCTURAL DESIGN BASIS ,

Section 4 and Appendix A of Chapter 1 include the general structural design criteria and requirements applicable to all buildings, structures, systems, ,

and equipment. Also addressed are the passive structural requirements and requirements to ensure active eauipment functions. Requirements for a unified system of classifying the structures, systems, and equipment with respect to function and structural integrity are established, as are codes and standards and acceptance criteria. Design loads and load combinations, as well as the required measures to mitigate the effects of in plant hazards, are defined.

These include the design for site-related criteria, including natural phenomena and environmental conditions. Requirements are also established to ensure that equipment will function under specified dynamic and environmental conditions.

The staff has reviewed Section 4 and Appendix A and has found them acceptable, except as otherwise indicated below.

Section 4.3.A, "Safety Classification" General Design Criterion (GDC) 1, "Quality Standards and Re;;ords," requires that nuclear power plant structures, systems, and components important to safety be designed, fabriccted, erected, and tested to quality standards commensurate with the importance of the safety function to be performed. Regulatory Guide i 1.26, "Quality Group Classification and Standards for Water, Steam and Radio- ]

active Waste Containing Components of Nuclear Power Plants," is the principal l document used by the staff in the review of this subject. However, the ALWR I Requirements Document proposes to use ANSI /ANS-51.1, "Nuclear Safety Criteria  ;

for the Design of Stationary PWR's," and ANSI /ANS-52.1, "Nuclear Safety  !

Criteria for the Design of Stationary BWR's," as an alternate way of complying with RG 1.26. Because these two industry standards have not been completely endorsed by the staff, it will be incumbent on the plant designer to resolve any deviations from RG 1.26, as discussed in Section 4.4 below.

CH 1, SEC 4 4-1 09/13/87

l Section 4.3.B, "Seismic Classification" GDC 2, "Design Bases for Protection Against Natural Phenomena," in part, requires that nuclear plant structures, systems, and components important to safety be designed to withstand the effects of earthquakes without inss of capability to perform their safety function. RG 1.29, "Seismic Design Classification," is the principal document used by the staff in the review of this subject.

In response to staff comments on Section 4.3.8 of the Requirements Document,

. EPRI agreed that a Quality Assurance Program in accordance with the applicable parts of 10 CFR 50, Appendix B should be applied to seismic Category II items (i.e., structures, systems, and equipment) that perform no safety function and whose continued function is not required, but whose structural failure or  ;

interaction would degrade the functioning of a seismic Category I structure,  !

system, or component to an unacceptable level. However, the EPRI response referenced Section 9.2.C of Chapter 1 for the type of program to be applied to such items. Section 9.2.C states only that the utility shall have the right to impose additional requirements to supplement the 10 CFR 50, Appendix B requirements and it does not explicitly address seismic Category II items. 1 The staff recommends that commitment to Positions C.2 and C.4 of RG 1.29 be I required for seismic Category II items. -

Section 4.3.B.2, "Seismic Category II"  ;

l This section states that to determine the amount of separation required between i seismic Category I and II structures, systems, and equipment, appropriate seis- j mic ductility factors shall be selected for the design to take credit for real-istic amounts of energy dissipation in seismic Category II items. The ductility factors are expected to be obtained from research, which will be documented by the time a specific ALWR is designed. It should be noted that for the staff to be able to justify the use of such factors, the limits of ductility must be .

1 clearly defined and correlated with the research results.

In response to this staff comment, EPRI stated that it will revise the last sentence of this requirement to read: "Appropriate approved seismic ductility CH 1, SEC 4 4-2 09/13/87

a factors shall be selected for design to take credit for realistic amounts of energy dissipation in C-11 items." The corresponding Engineering Rationale will be revised to read: "The seismic ductility factors and ductility limits selected are anticipated to consider the results of research which,is documented by the time of the ALWR, and to provide significant improvements over current practice." These changes are acceptable to the staff.

Section 4.4. "Codes and Standards" The Codes and Standards Rule, 10 CFR 50.55a, requires that the Edition and Addenda of Section III of the ASME Code applied in the construction of compo-nents be determined by the provisions of Paragraph NCA-1140, Subsection NCA, of Section III incorporated by reference in the rule. Subparagraph NCA-1140(a)(2) states that in no case shall the Code Edition and Addenda dates established in I the design specifications be earlier than 3 years prior to the date that the I nuclear power plant construction permit application is docketed. However, the Requirements Document Rationale for this section noted that the applicable Edition of these codes and standards will be that in effect approximately 42 months prior to the start of construction.

In response to a staff comment, EPRI made a commitment to revise the first -

sentence in the Engineering Rationale to state that the applicable edition of the codes and standards will be that in effect approximately 42 months prior to the placement of the first structural concrete. This commitment is not in conformance with 10 CFR 50.55a and, therefore, is not currently acceptable to the staff.

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EPRI recognized this apparent inconsistency with NCA-1140 and stated that it could probably be resolved by a Code Case or another technical solution at the time an ALWR is designed. It is the responsibility of the plant designer to j ensure that there is such a basis in effect to ensure compliance with 10 CFR 50.55a.

Cri 1, SEC 4 4-3 09/13/87

Section 4.4.C, "Regulatory Positions" In this section, EPRI has opted to take advantage of several recent or expected modifications to staff positions. These are as follows: ,

(1) Elimination of Arbitrary Intermediate Pipe Breaks: This has recently been implemented in a revision to SRP 3.6.2.  ;

(2) Leak-Before-Break Methodology (i.e. , limited and Broad Scope Rule revisions to GDC 4): A new Standard Review Plan, Section 3.6.3, has been issued for coment (Federal Register, p. 32626, August 28, 1987).

(3) Use of ASME Code Case N-411, "Alternative Damping Values for Seismic j Analysis of Classes 1, 2 & 3 Piping Sections": This was conditionally approved by the staff in RG 1.84, Revision 24, dated June 1986. In addition, it should be noted that the staff does not currently accept the l damping values of Code Case N-411 for use in analyses that employ the j independent support motion response spectrum methodology. '

(4) Use of Approved Independent Action Response Spectrum Analyses: This was recomended in NUREG 1061, "Report of the NRC Piping Review Comittee," _

dated April 1985, but it has not yet been incorporated into the SRP or an RG.

(5) Use of ASME Code Case N-397, "Alternative Rules to the Spectral Broaden-ing Procedures of N-1226.3 for Classes 1, 2 and 3 Piping": This was conditionally approved by the staff in RG 1.84, Revision 24, dated June 1986.

i (6) Analysir of Vibratory Loads with Significant High Frequency Input: In these analyses, the plant designer may algebraically combine high frequency results, which is a recomendation in NUREG-1061. However, that document also recamends that investigations be undertaken to establish the transition frequency between high and low frequencies when the designer is implementing the algebraic sumation rule for high CH 1, SEC 4 4-4 09/13/87

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frequency modes. Pending the results of such investigations, the staff l

position is that this deviation from RG 1.92 will be evaluated on a j case-by-ca n basis.

. l (7) Application of Seismic Equipment Qualification Experience to ew Designs:  ;

Although such application has been accepted by the staff for operating  !

plar.ts (in the resolution of USI A-46) and is being addressed in a i revision to ANSI /IEEE-344, the NRC has not yet accepted such applications to new designs. Therefore, the staff's position on this subject is still uncartain.

(8) Use of ANSI /ANS-2.3 (Instead of RG 1.76) To Define Tornado Effects: The staff agrees that the current tornado criteria may be too restrictive, but ANSI /ANS-2.3 has not been accepted as a basis for changing the criteria. Consideration is being given to using a staff-sponsored study of tornado statistics as such a basis. EPRI will be informed when any revisions to the criteria have been determined.

In responding to staff comments on Section 4.4.C, EPRI agreed to modify several of the requirements to provide a basis for the staff and the plant designer to reach agreement on the details of implementing the above positions. Subject -

6 to the foregoing exceptions, the staff has concluded that the information pre-sented in Section 4.4 of Chapter 1 is generally consistent with regulatory requirements on these issues.

Section 4.5, "Design Loads and Conditions" This section states that, when certain specified loading combinations dispro-portionately control the design of plant structures without rational bases, the plant designer say, with the concurrence of the utility, develop quantitative mechanistic design loads and combinations using probabilistic methodology.

The staff is not currently accepting a probabilistic approach as a basis for changing existing loads and/or loading combinations. In response to a staff comment to that effect, EPRI stated that this requirement will be changed to require approval by the NRC, as well as by the utility.

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Another issue discussed in this section is decoupling the magnitude of the OBE l from the SSE, which would conflict with 10 CFR 100, Appendix A. The staff agrees that the OBE should not control the design of safety systems; however, a staff position on this issue has not yet been fully developed. It,is currently ,

being evaluated by the staff on a medium priority basis as a part of NRC Generic Issue 119, "Piping Review Committee Recommendations." In addition, the ASME is addressing the possibility of revising ASME Code,Section III, NB-3650 rules for pip Mg design to keep the OBE from controlling the design. The results of these evaluations will provide the basis for the staff's final position on this issue. In the interim, the staff will consider requests to decouple the OBE from the SSE on a plant-specific basis.

1 Section 4.6, "Design Met'hodology" l l

This section states that the design of the ALWR shall be based on the industry codes and standards specified in Section 4.4 and that al's analysis and design techniques shall be in accordance with these codes and standards. However, l this cection also requires the plant designer to implement methods that  !

l minimize unnecessary conservatisms and to utilize accepted advanced dynamic analysis techniques where appropriate. The staff's evaluation of tnese matters

, is in Section 4.4.C of this report.

Section 4.7.A, "Seismic and Dynamic Qualification By Experience" This section includes the relatively new technique of equipment qualification by experience. This could be accomplished for a piece of equipment by justi-fying its dynamic similarity to previously qualified equipment or to equipment that has been exposed to other, more severe environments. The staff accepted this approach in resolving USI A-46 with respect to operating reactors. How-ever, this approach may not be acceptable for new plants because it may be difficult to show that experience envelopes the demand requirements for new equipment. Revision of RG 1.100, SRP 3.9.2, and SRP 3.10 to include potential use of experience data is uncertain at this time. In response to a staff comment on this subject, EPRI committed to revising a requirement in this section of the Requirements Document to state that the plant designer shall CH 1, SEC 4 4-6 09/13/87

make use of qualification by experience as permitted by governing codes and standards. The staff finds this comitment acceptable.

Another staff coment on this section related to a statement that allowed the plar.t designer to adopt the recomended anchorage from the equipment manufacturer.

The staff concern was that the manufacturer's recomended anchorage does not always prove acceptable for seismic qualification. In response, EPRI made a comitment to revise this requirement to read "... adopt the recomended seismically qualified anchorage from the equipment manufacturer." Tne staff's interpretation of this commitment is that the plant designer will determine that the manufacturer's seismic qualification procedures are consistent with all of the ALWR qualification requirements approved by the staff.

The staff has concluded that the information presented in this section, aug-mented by the above commitments, provides reasoriable assurance that the testing and qualification requirements for the ALWR will result in an acceptable program for seismic dynamic and environmental qualification of electrical and mechanical equipment.

Tab'e 4-1, "Structural Codes and Standards for Structures, Systems and Equipment" _

As suggested by the staff, EPRI has committed to adding a note to this table requiring that "all ASME Section Ill items must also satisfy the requirements of ASME III, Divis. as 1 and 2, Subsection NCA."

Table 4-2, "Industry Technical Standards" A number of ANSI /ANS standards identified in Table 4-2 (such as ANSI /ANS-51.1-1983 and ANSI /ANS-52.1-1983) are not endorsed by the NRC, and their use as the basis )

for plant construction may result in a conflict with NRC guidance in F.Gs, the SRP, j and the regulations. In response to this coment, EPRI stated tha; t will l l

w vise the introduction to Section 4.4 to require the plant destgrier to resolve i l

.nflicting requirements with regulatory positions and current codes and standards.

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Table 4-3, "Major Structural Design and Construction Codes" This table lists ACI-531-1979, "Building Code Requirements for Concrete Masonry Structures," for masonry wall design. The staff commented that although Appen-dix A to SRP Section 3.8.4 endorses the use of this code, it also gives warning that when the provisions of the code are less conservative than the corresponding provisions of the criteria given in Appendix A to SRP 3.8.4, the use of the code should be justified. A cautionary note of this kind may be appropriate where the ACI-531-3979 Code 12 mentioned as a major structural design code. The designer should Liso consider the requirements of IE Bulletin 80-11, "Masonry Wall Design."

In response to the above staff comment, EPRI stated:

ACI-531 should remain in Table 4-1. The requirements in this document meet the intent of the SRP requirements, as well as those of IE Bulletin 80-11. More detailed requirements for the design of masonry walls are given in Appendix A, Section 5.0 A.7, which requires that masonry walls be engineered as a substructure of the building, reinforced, designed and constructed in accordance with the Uniform Building Code (UBC). The ACI-531 code is also shown as a design code because it treats more masonry technical matters in more detail, and with better explanation of intent, than does the UBC. Becsuse the masonry walls of the ALWR will be rein-forced, many of the SRP qualifications on the use of ACI-531 do not apply.

The requirement to engineer the masonry walls as a substructure incorporates such considerations as in plane and out-of plane interstory drift effects, l in plane shear stresses, and collar joint shear stresses in multi-wythe walls, which were addressed in IE Bulletin 80-11. )

l The staff considers EPRI's response acceptable because it meets the general l intent of Appendix A to SRP Section 3.8.4 l l

Appendix A to Chapter 1. "Structural Design Requirements" l

Appendix A specifies more detailed structural design requirements for the equip- l ment, structures, and buildings of the plant than are provided in Section 4.5.

CH 1, SEC 4 4-8 09/13/87  !

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The appendix also states that the plant designer shall implement these require-ments and such other design requirements considered necessary to ensure the safety and licensability of the plant and to obtain reliable operation for the specified plant design life. ,

Section 2.0 of Appendix A addresses the required loads and load combinations for the ALWR, as well as the design methodology to be used, with respect to the following topics:

- All site-related criteria, including site proximity hazards, natural phenomena, anti environmental conditions.

Definition of all structural effects of operating loads and in plant hazards.

Definition of all design loads and load combinations.

General structural design criteria applicable to all buildings, structures, systems, and equipment.

Establishment of acceptance criteria. -

Section 3.0 of Appendix A contains detailed requirements for design loads and conditions fer all cf the topics listed above. The staf f's commem.<.i on the site design parameterz in Table 2-1 and Section 4.4.C.8 are also pertinent to

, the structural requirements in this section relative to wind loadings including tornados, hydrology, and geology and foundation conditions.

The discussion in Section 4.5 of this DSER relative to decoupling the SSE and OBE is also applicable to the information on earthquakes in Section 3.0. A.4 of Appendix A. In add' tion, this section states that in establishing the single free-field seismic ' notion input spectrum for the SSE design, spectral input motion shall be deconvoluted to specific points for application to structural analytical models, taking into consideration soil structure interaction phenomena. The deconvolution technique has been accepted by the staff in a few CH 1, SEC 4 4-9 09/14/87

e plant-specific cases, and it would be considered for a specific ALWR design if the design complies with the enveloping requirement in SRP Section 3.7.2.11.4.

This section requires a comparison of the floor responses from the finite soil boundary technique with those from the half-space modeling approach, and the selection of the most conservative response. Future modifications to the SRP may result in significant changes in this area, and the most current version of the SRP would be applied.

With respect to plant operating loads, the staff is currently evaluating the information relative to the methodology for defining the BWR safety relief valve loads to be used in the design of the suppression pool and will report the results when they are known. The remainder of the information on plant operating loads is consistent with the staff positions on this subject.

The requirements regarding pipe rupture loads in Section 3.0.D.1 of Appendix A contain a statement that dynamic effects resulting from a double-ended break of any pipe inside containment will not be explicitly considered in the design of the containment and emergency core cooling system (ECCS) for the ALWR. The dynamic effects which will not be considered include pipe whip, jet impingement, rapid subcompartmerit pressurization, hydraulic system internal loads and the rtotion of attached equipment. It is the staff's understanding that this re-quirement is only applicable if the ALWR plant has justified the use of the leak-befare-break concept. To be consistent with the implementation of the l leak-before-break technology under the recent broad : cope amendment to GDC-4 l

of Appendix A to 10 CFR Part 50, containment and ECCS functional and performance requirements are maintained. The staff's current interpretation of this rule change is that the above local dynamic effects (pipe whip, jet impingement, I etc.) can be excluded from the containment design b.:ser To retain high safety margins, the containment must continue to be sosigned to withstand all global loading and environmental effects up to and including the douvie-ended rupture of the largest pipe in tne ret.ctor coolant system. The heat removal and mass replacement capability of the ECCS (flow ratss, pressures, storage volumes) should continue to be desigr.ed to accoremcJate pipe ruptures up to and including the double ended rupture of the largest pipe in the reactor coolant system, even when leak-before break is dem 7strated. However, containment components l CH 1, SEC 4 4-10 02/10/88

and ECCS hardware (piping, pumps, valves) need not be protected against the l dynamic effects of pipe ruptures in systems qualifying for leak-before-break.

Section 3.0.0.1 of Appendix A should be revised to clearly state the above requirements. ,

In response to a staff comment relative to referencing ANSI /ANS-58.2 as providing acceptable criteria for postulating pipe breaks, EPRI agreed to revise several sentences in the Requirements Document. Two such references on pages A-17 and A-20 will be changed to read: "ANSI /ANS-58.2 as supplemented by applicable regulatory documents." The staff finds this commitment acceptable for defining mechanistic criteria for postulating pipe breaks. However, EPRI also stated l

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that a sentence on page A-19 would be changed to read: "Criteria for applica-tion of leak before break (LBB) to pipe rupture analyses are presented in ANSI /

ANS 58.2 (DRAFT) and NUREG-1061." These documents are not acceptable to the staff for defining leak-before-break criteria. Acceptable criteria for leak-before-break are currently in the proposed SRP Section 3.6.3, "Leak-Before-Break Evaluation Procedures," March 1987 Oraft, which is scheduled for release in 1988. The sentence on page A-19 should be revised to be consistent with the f'irst two revised sentences discussed above.

The staff is reviewing the in plant hazards requirements addressed in Section 3.0.0.2 relative to BWR suppression pool loads that remain after leak-before-break has been demonstrated. The results of the staff evaluation will be reported later.

Section 4.0 of Appendix A contains detailed requirements relative to loading combinations used in the design of buildings, structures, systems, and equipment.

In response to staff comments on this section, EPRI comitted to modifying the definition of the load designated as DF in Tables A-4 and A-6 of Appendix A.

DF consists of loads resulting from dynamic events associated with the faulted condition (ASME Service Level 0). EPRI has comitted to state that DF shall .

include the dynamic effects of postulated pipe ruptures that are not eliminated through the application of leak-before-break. The staff finds this acceptable;  !

however, the current staff position is that OF should be added to Service Level 0 loading combination (e) in Tables A-5 and A-7 of Appendix A. As a part of Generic Issue 119, "Piping Review Comittee Recomendations," the staff, on a low priority basis, is considering the possibility of decoupling the SSE from the loss-of-coolant accident (LOCM . Until this issse is resolved, the staff position will remain as stated above.

The staff hss concluded that, pending resolution of this open item, the l information in Appendix A relative to loading combinations, augmented by EPRI's comitments, is generally consistent with the SRP and is acceptable for a design requirerr.ents document.

i CH 1, SEC 4 4-11 09/13/87 4

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Section 5.0 of Appendix A contains detailed requirements relative to the design methodology used in the design of buildings, structures, systems, and equipment.

The staff submitted several comments on this section to the EPRI that.are aimilar to its comments on Section 4 of Chapter 1. Those comments and the responses are discussed in Sections 4.4 and 4.5 above.  ;

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5 MATERIALS

• The purpose of Section 5 is to provide general guidance regarding the selection of materials froin which nuclear plant components can be fabricated, based on past experience in LWR design, construction, and operation. The ALWR require-ments also provide guidance regarding precautions to be taken to avoid failure from known causes of degradation. Although these objectives have been generally achieved in Section 5, the etaff review led to the suggestions given below for l modifications and additions sc the Requirements Document will better meet GDC 14 relative to the reactor coolant pressure boundcry. '

Section 5.3. A.1, "Materials in the Reactor Coolant System and Related Systems" This section must be revised to bring it into agreement with NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for Coolant l

Pressure Boundary Piping," Revision 2, which was issued for public comment on July 11, 1986. For example, limiting the carbon content of 304 and 316 stainless steels to 0.065%, as specified in requirement 5.3.A.1.d, does not meet the 0.035% 1 i

maximum carbon limit in Revision 2, which is considered necessary to make the material resistant to sensitization and intergranular stress corrosion cracking l

of BWR piping. EPRI has stated that it will modify this section, as appropriate, l after the final version of Revision 2 and a generic letter have been issued by the NRC.

j Section 5.3.A.2, "Martensitic Stainless Steels" The ALWR requirements regarding "Martensitic Stainless Steels" are acceptable,  !

but they do not go far enough. In keteping with the spirit of utilizing past experience, the hardness of precipitation-hardened stainless steels must be I kept below that being specified in the nuclear industry today because failures are occurring. It may be that the Rockwell "c" hardness of 40 is too high for '

am application; a "c" of 25 may even be too high for some. (This comment i

also applies to Section 5.3.A.6, "Precipitation Hardened Stainless Steel.")

CH 1, SEC 5 5-1 09/13/87

EPRI responded, in part, that "new developments and experience will have to be evaluated by the designer at the time the heat treatment is specified." To ensure that this occurs, EPRI will add the following requirement to Set-tion 5.3.A.2.a and Section 5.3.A.6: "Hardness limits shall be specified on the basis of existing industry experience."

Section 5.3.A.3.b(1), "Ni-Cr-Fe Alloys (for BWR Applications)"

The staff suggested modifying this section to require avoiding designs utilizing Alloy 600 in which a crevice might be created. EPRI will revise this item to read as follows:

(1) Crevices in applications should be avoided; however, if they cannot be avoided, the design and residual stresses at operating temperature shall i:e limited to a level which has been shown to be acceptable.

Section 5.3.A.5, "Carbon and Low Alloy Steel Materials" The staff commented that the requirement that allowance be made for corrosion is potentially inadequate because TVA (among others) has shown that the accepted standard corrosion allowance is inadequate.

EPRI responded that the current requirement in no way accepts any "standard" corrosion allowance. Rather, it requires that the designer specifically consider all types of corrosion (general, pitting, and crevice) and allow for o them in the design. The following sentence will be added to Section 5.3.A.5:

"The allowance shall be in conformance with current industry experience ,

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6 RELIABILITY AND AVAILABILITY ,

This section of Chapter 1 establishes analysis and design requirements that will ensure that the plant reliability and availability requirements are con-sidered in all design activities. The objective is to ansure that the ALWR will have a significantly higher availability than the average of existing plants. For example, Section 6.2.A requires, in part, that "the Plant Designer for the ALWR plant shall prepare an availability analysis which shows that the plant systems end supporting maintenance systems and recommended spare parts are adequate to meet the availability goal."

The staff has not identified anything in this chapter that conflicts with existing NRC requirements. However, the addition of one item was recom.nerded to EPRI, as indicated below.

Section 6.2.B.4, "Failure Mechanisms" In response to a staff suggestion, EPRI has stated that it will add the follow-ing item to a list of failure mechanisms for which the ALWR shall be designed ~

so the plant is not prohibited from achieving its design life or availability:

Loss of strength and/or fricture resistance because of change (s) in the metallurgical state of alloys resulting from exposure to high temperature, thermal cycling or high radiation.  ;

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S 7 CONSTRUCTABILITY ,

The performance requirements in Section 7 support the overall objectives of the ALWR, particularly by reducing the utility's risk and capital costs through a significantly improved construction schedule, productivity, and construction techniques. The ALWR Program view is that the interface between the plant designer and the contractor must be recognized and included in the planning.

In addition, the following elements are treated in this section:

Integrated Design / Construction Schedule Detailed Construction Plan Utilization of Advanced Construction Techniques and Practices Utilization of Advanced Computer Technology in Design, Construction, Planning, and Control 1 Integrated System Completion, Testing, an< r/ Operator Acceptance The staff has reviewed Section 7 and has not identified any items that are in-compatible with NRC requirements except as discussed below.

Section 7.2.C.2, "Construction program Control" The integrated seftware system to be used by the constructor is to maintain and transfer data including quality assurance / quality control (QA/QC) requirements.

As suggested by the staff, EPRI will modify the data list by including mile-stones for scheduling construction verification inspection points for inspection and enforcement personnel and startup testing events. The software used will be subjected to the QA program requirements given in Section 9.  ;

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Section 7.2.C.3, "Construction Program Control" The list of activities for which the constructor must provide a site organiza-tion should include quality assurance / quality control. ,

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8 OPERABILITY AND MAINTAINABILITY ,

Section 8 of Chapter 1 specifies the utilities' requirements for enhancing the operability and maintainability of the ALWR by incorporating lessons learned at operating plants. The subjects covered include standardization of compensnts, -

with extra design margin to allow for 60 year operating life; increased ventila-tion and illumination; reduction of the types and numbers of valves used; better access and adequate space for testing, maintenance, and replacement activities; considerations for keeping radioactive emissions as low as reasonably achiev-able (ALARA); adequate supporting services and availability of spare parts; improved communications and personnel processing facilities; and other provi-sions that should minimize plant downtime.

The staff review of this section identified no conflicts with existing NRC requirements, with the exception that operation for more than 40 years is not  ;

currently permitted. As a result of staff comments, EPRI has stated that changes will be made in Section 8 as discussed below.

Section 8.2.B, "Cesian Features to Enhance Operability and Maintainability" 10 CFR 50.55a(g) requires that pumps and valves that are classified as ASM?

s Code,Section III, Class 1, 2, or 3 and whose function is important to safety De designed and provided with access to enable the performance of inservice testing of the purps and valves for assessing operational readiness set forth in Section XI of the code. A qualifying statement should be added in this section specifying that the plant designer shall also selr.ct pumps and valves that can meet the requirements of Sections IWP and IWV of Section XI of the  ;

code.

l In response to this comment, EPRI stated that it will modify this section to '

provide better coverage of inspectability and provisions for inservice inspection.

As suggested by the staff, EPRI will also add appropriate acoustical monitoring requirements to this section.

CH 1 SEC 8 8-1 09/13/87

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Section 8.2.B.4, "Human Factors Provisions To Facilitate Operation and Maintenance The staff recommends that the Institute of Electrical and Electroni,cs Engineers (IEEE) "Guide for the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Generating Stations" (P1023/05) be used as a reference. EPRI-2360, "Human Factors Methods for Assessing and Enhancing Power Plants Maintainability," also is an excellent reference.

Section 8.2.C.2, "Preventive Maintenance and Inspections" In response to a staff comment, EPRI has stated that it will expand this section  ;

to include the integration of equipment inservice inspection and testing requirements imposed by codes and standards that apply at the time of design. '

Section 8.2.C.3, "Document Control" As suggested by the staff, EPRI has stated that it will modify the Engineering Rationale in Chapter 1 to include utilization of experience gained from two pilot plant studies (for Nortnern States Power and Virginia Electric Power plants) in the EPRI/00E Plant Life Extension Program. ,

Section 8.2.C.4, "Personnel and Staffino" In response to tha staff's suggestion that this section should include personnel l qualification requirements, EPRI said it will add the following statement:

i Personnel selected and trained for the plant operating organization shall ,

meet the requirements of ANSI /ANS 3.1, "Selectior., Qualification and Training of Personnel for Nuclear Power Plants" (latest revision at the l time of submittal of the aoplication for an operating license).

Section 8.3.A, Minimizino Dose Levels to Personnel (ALARA) 1 Discussions of the control room in Section 8 do not cover protection against l accidental releases of toxic and radioactive gases and habitability during CH 1, SEC 8 8-2 09/13/87 J

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i accidents. EPRI stated that these subjects will be treated in Chapters 6 i and 10.

. Table 8-2, "LWR Operation Problem Areas to be Addressed i_n,ALWR Desian" As suggested by the staff, EPRI stated that it will add the following item to this table: "insufficient structural integrity and mechanical reliability of pump components."

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9 QUALITY ASSURANCE ,

Section 9 of Chapter 1 identifies the major elements of the ALWR overall Quality Assurance (QA) Program and specifies the requirements for the supporting QA programs for the primary organizational entities involved in and supporting the design, procurement, construction, and pre-operational testing of an ALWR plant.

EPRI has also affirmed that this section is intended to be in accordance with til current regulatory requirements for quality assurance at the level of detail provided.

The stcff has reviewed Section 9 and found it acceptable with the changes discussed below.

Section 9.2.B, "Program Requirements" In response to a staff comment that the designer must set up a program to moni-tor and control the performance of each and every vendor, EPRI stated that actual monitoring of vendor performance is considered beyond the scope of the Requirements Document. However, the following requirement will be added to Section 9.2.B: "The prime contractors shall be responsible for monitoring the performance of subcontractors to assure compliance with the quality assurance program."

EPRI also stated that it will revise the last paragraph of this section to read as follows: "The plant designer (s) shall ensure the Quality Assurance Program developed in Item 9.2.B is consistent with the requirements of ANSI /ASME HQA-1 and its supplements as endorsed by Regulatory Guide 1.28 Rev. 3."

Table 9-1, "Quality Problems During Design and Construction" In response to a staff suggestion, EPRI stated that it will add the following two line items to the table:

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10 LICENSING ,

Section 10 provides the licensing requirements for the ALWR. These include a ,

statement that "the ALWR shall comply with the NRC regulatory requirements that are in effect when the ALWR is to be licensed." This compliance is to be demonstrated by compliance with a defined set of regulatory requirements that will be finalized in Appendix B of the Requirements Document.

At present, Appendix B is described as a list of documents that contain the NRC criteria applicable to LWR design. Exceptions to this list will be identified as the later chapters are developed, and Appendix B will be updated accordingly.

All of the items in Appendix B are supposedly cross-referenced to the sections in the document where they are to be addressed. However, the staff has found that the appendix requires considerable reworking to ensure that the cross-references provided are correct. In many cases, where sections in Chapter 1 ,

are referenced, those sections do not identify or address the NRC regulation or guideline indicated in Appendix B. EPRI has stated that Appendix B will be modified to make it consistent with the foregoing sections of Chapter 1.

4 The staff suggested that it would be useful if Chapter 1 included a listing of the NRC unresolved generic safety issues that were identified as applicable to the ALWR as of July 1,1986. EPRI responded that an appendix will be developed for Chapter 1 that will identify all the issues that have been or will be the i subject of EPRI topic papers proposing elements of resolution for applicable unresolved issues. This appendix will identify the sections of the Requirements )

Document that contain the specific requirements that have been incorporated to resolve those issues. j l

il The plant designer should alto be required to determine that such requirements j are compatible with the issue resolutions adopted by the NRC after the staff l

{ completes its review of the Requirements Document.

l CH 1, SEC 10 10-1 09/13/87 )

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Section 10.2.C.2 requires the plant designer to review the resolution of issues identified after July 1, 1986, consistent with the process identified in NUREG-1197 to establish any changes to the regulatory requirements. While NUREG-1197 outlines such a process, at this point the guidelines for its imple-mentation have not been developed. Specifically, the steps involved in imple-mentation and the responsibility for each step must be defined. Additionally, guidelines must be developed on how the screening criteria in Section 3.1.3.2 of NUREG-1197 are to be used; i.e., it must be determined (1) how the ALWR core melt frequency goal and compliance with the Commission's safety goals are to be assessed without a design and (2) whether the ALWR requirements relative to offsite radiologicC dose consequences of an accident would be exceeded as a result of the issue involved.

The staff recommended that a set of principal des 1gn criteria for the ALWR, as required by 10 CFR 50.34, :hould be developed and included in Chapter 1. EPRI has responded that the ALWR must be in compliance with the current interpreta-tion of the GDC in Appendix A to 10 CFR 50. The only exception is GDC 4 regarding application of the leak-before-break methodology to the ALWR, which was the wbject of a plant optimization subject previously reviewed and commented upon by the NRC. EPRI has stated that a requirement will be added  ;

a to Chapter 1 of the Requirements Document to ensure that the ALWR design is ir compliance with the GDC as described above. Further, the plant designer will be required to establish a set of principal design criteria that will be docu-l mented in the safety analysis report used to support an ALWR design certifica-tion application. This requirement will be added to Section 10 of Chapter 1.

. Section 10 also must be expanded to describe how EPRI intends to use the

' Raquirements Document in the licensing process and what NRC approval of this  !

I document is intended to mean. It should be understood that NRC approval of the Requirements Document would imply general agreement with what is in the document, but such approval would not imply that the document is a comniete and adequate

, set of requirements for a nuclear power plant. As EPRI has stated, the Require-ments Document is intended for use with companion documents, such as utility procurement specifications, which would cover the remaining technical require-ments for a specific plant. Therefore, a thorough staff licensing review of an application for constructien and operation of an ALWR plant will be necessary.

] l CH 1, SEC 10 10-2 09/13/87 4

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However, the time and effort to perform that review can be reduced to the extent the issues involved have been resolved during the staff's review of the ALWR I 1

Utility Requirements Document. . '

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CONCLUSIONS ,

The Requirements Document recognizes that a specific ALWR plant will have to meet NRC's requirements. However, it includes some ut.ility requirements that are different from existing NRC requirements, based on the expectation that appropriate changes in the NRC requirements will be approved in the course of chis review. That is a logical approach for EPRI to take, based on the staff's initial, largely favorable reactions to EPRI's suggested resolutions of generic safety issues and its regulatory optimization proposals for Chapter 1 (Spets, June 6, 1986). While some changes in leak-before-break and source-term criteria have been made recently, it is not yet known to what extent all of these ALWR Programpositionswillbeapproved. To the extent they are not, appropriate changes will have to be made in the Requirements Document.

l Table B-1 in Appendix B to Chapter 1 lists most of the current regulatory l

criteria for design of LWR plants. However, in some disciplines, such as j instrumentation and control, they are not integrated sufficiently with the l general design requirements to allow the staff to conclude that a designer following the utility requirements in Chapter 1 would satisfy the NRC's criteria. Consequently, formulation of the staff's judoment in such areas must I await its review of subsequent chapters.

Notwithstanding the above, the staff finds that it is in general agreement with the objectives and overall requirements expressed in Chapter 1. Subject to confirmation that the ALWR Program has met its commitments to modify various l l

items in Chapter 1, the staff believes it will be possible to make the following 1 determination in the final SER:

I Chapter 1 of the Requirements Document contains requirements that, if properly translated into a design in accordance with the NRC regulations in force at the time of submittal, should generate a nuclear power plant that will have all the attributes required by the regulations to ensure that there is no undue risk to the peblic health and safety or to the environment.

CH 1, CONCL C-1 09/13/87

REFERENCES ,

1. Letter from C. Frederick Sears (Northeast Utilities), Chairman, ALWR Utility Steering Committee, to Harold R. Denton, Director, Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, December 3, 1985, enclosing topic papers on unresolved safety issues for Chapter 1.

2. Letter from Themis P. Speis, Director, Division of Safety Review and Oversight, ONRR, NRC, to C. F. Sears, Chairman, ALWR Utility Steering Committee, January 24, 1986, responding to Ref. #1.

3. Letter from C. F. Sears, Chairman, ALWR Utility Steering Committee, to Harold R. Denton, NRC, March 24, 1986, enclosing a report entitled "Advanced i LWR Plant Optimization Subjects for Chapter 1 of the Requirements Document."

4. Letter from T. P. Speis, NRC, to C. F. Sears, Chairman, ALWR Utility Steer-ing Committee, June 6, 1986, responding to Ref. #3.

5. Letter from C. F. Sears, Chairman, ALWR Utility Steering Committee, to H. R. Denton, NRC, July 8, 1986, enclosing Advanced LWR Requirements Document Chapter 1.

6. NUREG-1197, Advanced Light Water Reactor Program - Program Management and Staff Review Methodology, December 1986.

7. Letters to E. E. Kintner (GPU Nuclear Corp.), Chairman, ALWR Utility .

Steering Committee, from T. P. Speis, NRC, dated January 5, 1987, March 11, 1987, and March 18, 1987, requesting additional information concerning Chapter 1 of the ALWR Utility Requirements DocLnent.

8. Letter from E. E. Kintner, Chairman, ALWR Utility Steering Committee, to T. P. Speis, NRC, March 27, 1987, responding to hRC letters in Ref. #7. .;

9. NUREG-0800, Rev. 3, Standard Review Plan for the Riview of Safety Analysis Reports for Nuclear Power Plants, LWR Edition.

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CH 1, REF R-1 09/13/87

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DRAFT SAFETY EVALUATION REPORT 4

ON CHAPTER 2 0F EPRI's ADVANCED LIGHT WATER REACTOR UTILITY REQUIREMENTS DOCUMENT prepared by the Office c* Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission February 1988 l

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PREFACE ........................................................... v

1. INTRODUCTION .................................................. 1

2. COMMON REQUIREMENTS ........................................... 4

3. MAIN / EXTRACTION STEAM SYSTEM .................................. 5

4. FEEDWATER AND CONDENSATE SYSTEM ............................... 6

5. CNEMICAL ADDITION SYSTEM ...................................... 8

6. CONDENSATE MAKEUP PURIFICATION SYSTEM ......................... 9

7. AUXILIARY STEAM SYSTEM ........................................ 9

8. CONCLUSIONS ................................................... 10

9. REFERENCES ..................................................... 11 FIGURES t

1. Overview of PWR Power Generation Systems ...................... 2

2. Overview of BWR Power Generation Systems ...................... 3 l

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EPRI ALWR CH 2 DSER iii 02/17/88

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i PREFACE .

This Draft Safety Evaluation Report (DSER) is the second in a series the Nuclear Regulatory Commission (NRC) plans to issue on its reviews of the 13 chapters of the Advanced Light Water Reactor (ALWR): Utility Requirements Document (URD). The first DSER in this series, which was issued in September i 1987, addressed the URD Executive Summary and Chapter 1 regarding overall requirements of the ALWR program. Chapters 3, 4, and 5 are currently under-going staff review, and the present Electric Power Research Institute (EPRI) schedule calls for submitting the remaining chapters during 1988. On the basis of that schedule, the staff expects to complete the last DSER in the series by September 1989. EPRI plans to submit changes in the 13 chapters as a result of the comments by NRC and others. Those changes will be considered by the staff while it is preparing the final SER.

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EFRI ALWR CH 2 DSER v 02/17/88 l

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DRAFT SAFETY EVALUATION REPORT ,

ON CHAPTER 2, "POWER GENERATION SYSTEMS" J

1. INTRODUCTION On October 15, 1986 (Ref. 1), EPRI* submitted URD Chapter 2 entitled "Power Gen 6 ration Systems" for staff review. This chapter defines the utility-generated requirements for the main / extraction steam, feedwater and condensate, chemical addition, condensate makeup purification, and auxiliary steam systems.

These systems are depicted schematically in Figures 1 and 2 for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The utility requirements apply to both types of plants in sizes up to 1350 MWe; however, an 1100-MWe plant size with a sixaflow turbine was used in establish-ing some requirements that are based, in part, on economic evaluations.

On May 27, 1987 and June 12, 1987 (Refs. 2 and 3), the NRC staff asked EPRI to supply additional information. EPRI provided the information in its response ,

dated September 17, 1987 (Ref. 4). No topic papers addressing unresolved safety issues and no optimization proposals of technically supportable alternatives to existing regulatory requirements were submitted for consideration with Chapter 2.

The staff's approach to reviewing the URD is described in Sections 1.3 and 10 of the DSER for Chapter 1. As nated therein, the staff is using the NRC Standard Review Plan (SRP, Ref. 5) as guidance for the review. However, the primary emphasis of the URD is on the prevention of significant problems that have been experienced in existing plants and many details are lacking that will have to be provided later for specific designs. The staff, therefore, has considered the utility requirements at the level of detail presented in order

• For convenience, the term "EPRI" is used in this document to designate the Electric Power Research Institute and/or its ALWR Utility Steering Committee.

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e to identify any actual or potential conflicts with regulatory requirements, but not to determine their adequacy with respect to the NRC requirements.

2. COMMON REQUIREMENTS .

Paragraph 2.1 of Chapter 2 requires that the power generation systems "shall comply with the overall requirements of Chapter 1." Consequently, the resolu-tion of any open issues identified in the DSER for Chapter 3 (for example, the application of leak-before-break analyses) could result in associated changes in Chapter 2.

In response to a staff comment calling attention to an apparent deficiency in Chapter 2, EPRI stated that "Chapter 1 provides guidance to the Plant Designer for classifying and designing safety related portions of systems for seismic and environmental qualification." (Ese Item 3 of Reference 4.) Hence, the staff has re-examined the relevant paragraph and table in Chapter 1 (Para-graph 4.3, "Classification Requirements," and Table 4-1, "Structural Codes and Standards for Structures, Systems and Equipment") and found that the classifi-cation in/ormation is insufficient and too general to provide adequate guidance.

The staff, therefore, recommends that, for each system listed in Chapter 2, the corresponding design code or standard be specified for the piping and equip-ment. Also, the schematic diagram for each system should include the jurisdic-tional boundaries for the corresponding design codes and standards. The staff position on seismic and environmental qualification is stated in Section 4 of the DSER on Chapter 1.

In Paragraph 2.2.B. EPRI has provided substantial guidance to plant designers aimed at minimizing and simplifying the valving throughout the power generation 4 systems. The staff requested clarification of several items in its comments (Ref. 4) on valving and piping materials (Para. 2.2.C), but did not find any discrepancies with respect to current licensing requirements.

With respect to instrumentation and controls (I&C), Chapter 2 defines functional requirements that will affect their type, range, and location but notes that the actual design requirements will be idefitified in Chapter 10. The staff is, EPRI ALWR CH 2 DSER 4 02/17/88

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therefore, deferring its judgement on the I&C requirements in Chapter 2 until Chapter 10 is reviewed.

3. MAIN / EXTRACTION STEAM SYSTEM The main / extraction steam system is defined in Paragraph 3.1 of Chapter 2 as l l

censisting of the following for a BWR: main steam piping downstream of the second main steam isolation valves (MSIVs), extraction steam piping, turbine bypass system, and moisture separator / heater. (The BWR MSIVs are not included i because they are addressed in Chapter 3 of the URD.) For a PWR, the system consists of the main steam piping from the steam generators to the main tur-bine, the MSIVs, extraction steam piping, turbine bypass system, moisture separator / reheater, safety valves, and power-operated relief valves. (The i turbine generator requirements will be in Chapter 13.)

The performance requirements for the BWR specify that the minimum steamline i volume should be at least 2,850 cubic feet and the turbine bypass capability should be at least 33 percent of the full-load turbine steam flow rate at full-load steam pressure. Similarly, the performance requirements for the PWR  ;

specify that the maximum differential pressure between any two steam generator outlet nozzles should be less than 10 psi and the turbine bypass capability should be at least 40 percent of the full-load turbine steam flow rate at full-load steam pressure. Because the main steam piping will be designed so as to pass the full-rated flow of steam to the main turbine, it will have the {

capability to remove the residual heat from the reactor system in conformance l with 10 CFR 50, Appendix A, General Design Criterion (GOC) 34, "Residual Heat l

Removal" (Ref. 6).  ;

l Chapter 2 specifies that the extraction steamlines will be at least 1 inch in j diameter but sized to maintain a flow rate of less than 10 feet per second.

These lines will be routed continuously downward at 1 minimur. of 1/8 inch per i foot of pipe, and each line will contain one pneumatically operated fail-open valve and one manually operated locked-open valve.

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EPRI ALWR CH 2 OSER 5 02/17/88

i The above features are intended to improve the performance of the main steam system by improving its normal operating capability and, thereby, reducing the liinlihood of a steam system transient. ,

The MSIVs in a PWR will be fail-closed, bidirectional valves capable of stopping fully developed steamline break flows within 5 seconds in either direc-tion. The valves will be environmentally qualified for both normal operating i conditions and for the environment resulting from a steamline break. Thus, in the event of a main steamline break and a concurrent single active failure of one MSIVs, the remaining isolation valves will close and limit the blowdown to the one steam generator with the brcken main steamline. (The design of the MSIVs in a BWR is discussed in Chapter 3 of the URD.)

During its review of the material provided on the main / extraction steam system in Chapter 2, the staff did not identify any items that did not conform with the criteria of SRP Section 10.3, "Main Steam System." However, the utility ,

requirement in Paragraph 3.2.A.1.b of Chapter 2 may need to be reviewed with respect to specific plants. The requirement indicates that turbine trip without reactor trip is permissible up to a certain power level to prevent reactor trip following spurious turbine trips that sometimes occur during i startup (also see Item 19 of Reference 4 in this DSER).

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4. FEEDWATER AND CONDENSATE SYSTEM i

The feedwater and condensate system is designed to return high-quality feedwater from the condenser hotwell to the reactor vessel in a BWR or the steam genera-l* tors in a PWR, In addition, the system includes a number of stages of regenera-tive feedwater heating and provisions for maintaining feedwater quality. '

Chapter 2 of the URD specifies that this system will incl,ude three condensate pumps, three low pressure feedwater heating trains with four feedwater heaters j each, three feedwater pumps (plus three feedwater booster pumps in a PWR plant),

and two high-pressure feedwater heating trains with two feedwater heaters each.

In 3ddition, the BWR system will include a full-flow condensate domineralizer system and two feedwater heater drain pumps.

! EPRI ALWR CH 2 DSER 6 02/17/88 l

a The URD performance requirements state, in part, that the capability to operate the plant at full power should be maintained in the event of failure of a single pump. No low suction pressure trip of the main feedwater pump is to be provided. Waterhammer transient loads are to be below the system p'ressure boundary design limits. The system also will incorporate recirculation lines designed to permit system operation when the demand for feedwater flow is low, such as during startup and shutdown. The specification for overall pump flow capacity calls for a 3 percent margin above the design flow rates to account for wear. Thus, the system should provide a reliable supply of feedwater to the reactor (BWR) or the steam generators (PWR) for all modes of operation, including startup and shutdown. The above design guidelines are also intended to ensure the availability of adequate feedwater for off-normal transient events, thereby reducing challenges to safety-related systems. l Redundant main feedwater isolation is specified for the PWE. system; this will i be provided by the feedwater control valve and a separate feedwater isolation ]

valve upstream of the auxiliary feedwater isolation line. Both valves are l designed to fail closed on loss of actuating fluid and are specified to close  ;

within 5 seconds after the actuating signal has been received. These features l are intended to prevent excessive reactor coolant system cooldowns and/or con-tainment overpressurizations as a result of the addition of excessive feedwater  !

to the steam generators following a steam-line or feedwater-line break. They also will provide appropriate isolation of the non-safety-related part of the main feedwater system 'n order to ensure the decay heat removal function of the safety-related auxiliary feedwater system. In addition, a check valve will be provided at each steam generator to prevent reverse flow under acci-dent conditions (such as a feedwater-line break) and, thus, prevent the blow-down of more than one steam generator. The above isolation features ensure that postaccident decay heat removal functions can be provided in accordance with the requirements of GDC 44, "Cooling Water."

The feedwater and condensate system is also required to maintain water quality suitable for long-term power operation, startups, shutdowns, and extended out-ages. In the ALWR, properly designed condensate polishers will be provided to maintain water chemistry within specified limits. In the URO, the polishers i are sized to meet this requirement, assuming a condenser leak of 0.001 gpm EPRI ALWR CH 2 OSER 7 02/17/88

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l during continuous operation and 0.1 gpm during an orderly shutdown lasting not more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In addition, the polishers will provide adequate cleanup function during plant heatup and low power operation. No ion exchange resin regeneration will be provided in the system. In PWR plants, side-stream polishers will handle at least one-third of the full condensate flow; however, there will be a provision for installing full-flow polishers. In BWR plants, full-flow polishers will be installed. The polisher system will include flow controllers and instrumentation to ensure its efficient operation. In the feedwater and condensate system, connections will be available for injecting chemicals and taking samples as required for controlling system chemistry. In addition, there will be a provision for injecting chemicals into the condenser for biofouling control. The water chemistry in the system will be further controlled by deaerating the condensate during startups and during normal plant operation.

The staff reviewed the utility requirements in Chapter 2 for the feedwater and condensate system and found that they conform with the NRC criteria in Sec- i tions 10.4.6 and 10.4.7 and Branch Technical Position MTEB 5.3 in the SRP, and Regulatory Guide 1.56 (Ref 7).

5. CHEMICAL ADDITION SYSTEM The chemical addition system is designed to maintain feedwater, condensate, and offgas chemistry within required limits by the addition of suitable liquid l or gaseous chemicals. )

For the BWR, this system will consist of gas generation and/or storage  !

f acilities, piping, flow meters, instrumentation, and addition points on the j

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feedwater and offgas system. It should be able to introduce hydrogen into the feedwater at levels above 30 percent of reactor full powe'r and oxygen into the offgas system to recombine with excess hydrogen. A 3-cay supply of these gases will be stored on site.

For PWR, this system will consist of the chemical addition tanks, pumps, piping, instrumentation, and addition points on the feedwater and condensate system.

EPRI ALWR CH 2 OSER 8 02/17/88

It should be able to inject suitable amounts of hydrazine and ammonia or morpholine during plant operation and during plant layups.

The staff reviewed the utility requirements in Chapter 2 for the chtmical addition system and found that they conform with the NRC criteria in Sec-tion 5.4.2.1 and Branch Technical Position MTEB 5-3 in the SRP, and with EPRI's "Guidelines for Permanent BWR Hydrogen Water Chemistry Installations," (Ref. 8).

6. CONDENSATE MAKEUP PURIFICATION SYSTEM The condensate makeup purification system is designed to treat the raw makeup j water and store it for use in the feedwater and condensate system. It also will supply the PWR primary makeup system and the fuel pool. The purification system will consist of a domineralizer for rer,ving ionic impurities, a vacuum degasifier for removal of dissolved oxygen, a demineralizer water storage tank for sampling, and a condensate storage tank. In some cases, additional equip-ment may be needed to filter, clarify, and soften the makeup water. The system is designed with two 100-percent-capacity trains that have certain site specific features. Instruments and controls will monitor and control system performance.

The staff evaluated the design and operational requirements in Chapter 2 for the condensate makeup purification system and found them in general agreement with the NRC criteria in SRP Section 9.2.3.

7. AUXILIARY STEAM SYSTEM The auxiliary steam system is doigned to supply low pressure clean steam to various plant components when the main steam system is not available and to be the normal source of steam for the radwaste evaporators. The coxponents and functions for which this auxiliary system will provide steam are the following:

deaerator pegging (PWR) steam jet air ejectors

• turbine gland sealing

• boron recycle evaporator and batch tank (PWR)

EPRI ALWR CH E DSER 9 02/17/88 1

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• radioactive waste evaporator / concentrator

• reactor core isolation cooling turbine test (BWR) miscellaneous components ,

In Paragraph 7.2 of Chapter 2, the URD states that the auxiliary steam system will have no safety-related function and the staff confirmed that this is true.

Therefore, no further review of this section was conducted.

8. CONCLUSIONS On the basis of the evaluations summarized above, the staff concluded that the utility requirements included in Chapter 2 are in general agreement with the l NRC guidelines and regulatory requirements for the power generation systems involved. Subject to confirmation that EPRI has met its commitments to modify various items in Chapters 1 and 2, the staff believes it will be possible to make the following determination in the final SER:

Chapter 2 of the Utility Requirements Document contains requirements that, if properly translated into a design in accordance with the NRC regulations  !

in force at the time of submittal, should generate a nuclear power plant that will have all the attributes required by the regulations to ensure that there is no undue risk to the health and safety of the public or to  ;

the environment.

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EPRI ALWR CH 2 DSER 10 02/17/88

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9. REFERENCES ,

1. Letter from C. F. Sears (Northeast Utilities), Chaiman, ALWR Utility Steering Committee, to H. Denton, Director, Office of Nuclear Reactor Regulation, NRC, enclosing Chapter 2, "Power Generation Systems," of the ALWR Utility Requirements Document, October 15, 1986. j

2. Letter from P. H. Leech, NRC, to E. E. Kintner (GPU Nuclear), Chairman, ALWR Steering Committee, requesting additional information relative to Chapter 2 of the ALWR Utility Requirements Document, May 27, 1987. ,

3. Letter from P. H. Leech, NRC, to E. E. Kintner (GPU Nuclear), Chairman, ALWR Steering Committee, requesting additional information relative to Chapter 2 of the ALWR Utility Requirements Document, June 12, 1987.

4. Letter from E. E. Kintner (GPU Nuclear), Chairman, ALWR Steering Com-mittee, responding to References 2 and 3, September 17, 1987.  ;

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5. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan ter the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," Rev. 3, July 1981.

6. Code of Federal Reculations, Title 10 "Energy," U.S. Government Printing Office, Washington, DC, revised annually, l 1 1

7. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.56, "Maintenance of Water Purity in Boiling Water Reactors," Rev. 1, July 1978.

8. EPRI, "Guidelines for Permanent BWR Hydrogen Water Chemistry Installations,"

1987 Revision, t

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