ML20127G964

Summary of 921005 Meeting W/Industry Siting Group on Technical & Regulatory Review of Siting Issues Re Site Safety & Plant Parameter Envelope Methodology.List of Attendees,Agenda & Viewgraphs Encl

ML20127G964
Person / Time
Issue date:	11/12/1992
From:	Zalcman B Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9211180339
Download: ML20127G964 (32)

Text

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8[pueyjo, UNITED STATES NUCLEAR REGULATORY COMMISSION

'- ,'z WA6HING TON. D, C. 20665 Noveaber 12, 1992 t

ORGANIZATION: INDUSTRYSITINGGROUP(ISG) ,

SUMMARY

OF OCTOBER 5, 1992 MEETING WITH INDUSTRY SITING GROUP ON THE TECHNICAL AND REGULATORY REVIEW OF SITING ISSUES - SITE SAFETY AREA, AND PLANT PARAMETER ENVELOPE METHODOLOGY On October 5, 1992, the NRC staff met with representatives of the Indestry Siting Group (ISG)'to discuss the ISG review of the technical and regulatory requirements for an early site permit (ESP) application. Originally noticed as sequential meetings on site safety issues and on the plant parameter envelope (PPE) methodology, the meeting discussions were related and resulted in this single meeting summary. These public meetings were part of the series with the NRC staff to discuss the issues related to siting nuclear power plants. The ISG staff presented information on its PPE development activity that may be used by prospective applicants in developing their ESP applications. Related meetings were held on environmental protection issues on June 2,1992, on emergency planning issues on June 30, 1992, and on regulatory and legal issues on September 3,1992.

In a letter dated May 26, 1992, the Nuclear Management and Resources Council (NUMARC) submitted a document titled: "Early Site Permit Demonstration Program, Regulatory Criteria Evaluation Report," that provides the basis for the discussions on the site safety area. In a letter dated August 28, 1992, NUMARC submitted h document titled: " Plant Parameter Envelope Development, Overview of Approach and Work Product Example," that provided the basis for the discussions on the PPE. A list of the October 5,1992, meeting attendees and the meeting handout are included as Enclosures 1 and 2. res)ectively.

This meeting provided a forum for the ISG staff to interact wit 1 the NRC legal

• and technical staffs on ESP regulatory issues; no decisions were_made or positions taken during this meeting.

Prior to the ISG staff presentation, the NRC staff updated the audience on the status of the proposed revision to 10 CFR Part 100. " Reactor Site Criteria."

The staff anticipated that Part 100 would be published in the near future.

Therefore, inasmuch as it was still predecisional, the staff would limit its discussion to the information that was released to the public during

. presentations before the Commission and the Advisory Committee for Reactor Safeguards, i

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! In their opening remarks, the ISG staff outlined the industry's activities on

! the technical issues related to the siting rule, in the seismic and geologic i areas, there has been a continuing dialogue between the industry, the ad hoc

advisory committee (AHAC), and the staff. The NRC staff indicated that the l AHAC could serve a useful purpose in establishing the industry's viewpoints;

for example, in issue areas such as soil liquefaction potential and dam failure and flooding. The ISG staff indicated that it anticipated another i

AHAC would be constituted to develop the industry's comments on the non-l seismic issues.

The industry staff presented a brief discussion of its activities related to i the population density criteria and their impact on the safety goal and l societal risk. The staff presented the criteria to the Commission on June 24,

! 1992, as part of the Part 100 rulemaki'ng briefing. The ISG staff indicated

! that it could not find a technical justification for the criteria and

illustrated a series of dose-vs-distance curves (prompt individe.1 risk,

! latent individual risk, individual dose profile, population dose profile, and i percentage of safety goal) to support its position. The NRC staff interjected l and reminded the ISG that the proposed rule had not been published to date and

! that it was inappropriate to discuss the merits of the rule at this meeting, j The staff noted that this meeting would be limited to a discussion of the

methods and a>proaches that an applicant could consider to satisfy the siting ,

i criteria (eitler current or proposed) and not the basis for the proposed role.

The staff also offered that comments provided during the public comment period on the proposed rule include the technical bases to support positions and recommendations in areas of agreement and disagreement with the staff.

I The ISG staff described its proposal for-the format and content of the site

safety analysis report (SSAR) t1at would te submitted as part-of an ESP l a) plication. In particular, the SSAR would include a description of site claracteristics and a safety assessment of the site affecting facility design i

consistent with Chapter 2 of safety analysis reports (see: ' Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear

Powerplants). The SSAR would also include the relevant portions of Chapter 3

dealing with the major structures, systems, and components (SSCs) that bear i significantly on the acceptability of the site.

l The ISG staff discussed an a)proach, establishing PPEs, to define the reactor I characteristics that are to >e considered to assess site features and to

assess potential consequences from construction and operation. The approach outlir .d is intended to ensure that either a design for a specific plant, a design for a class of plant, .or a design for a plant that is nt,t yet specified

' could be characterized in sufficient detail to develop a design envelope. In all, the ISG planned to develop seven PPEs:- four for specific advanced light water (ALWR) reactor designs, a composite of the four evolutionary ALWRs, a passive reactor design, and-an overall composite of the composite evolutionary and passive designs. For the purposes of assessing site suitability, the ISG 180003 '

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3 staff suggested that an applicant could adopt a parameter set and corres-ponding design values as part of its application. In this manner, if an applicant was predisposed to a preferred design, it would include in its application the parameters and values for that design; if an applicant did not decide on a specific design, it would include in its application the parame-ters and values for one of the composite plants.

The NRC staff expressed a number of concerns regarding the PPE approach that

- was presented in the context of the site safety analysis report. These concerns included: 1) the ownership and control over the PPEs (for example, a vendor, an industry group, or the applicant); 2) the continuing significance of PPE values during the license proceedings (for example, adoption of a referenced set of PPEs or inclusion of, the specific parameters and values in the application); 3) the obligation to reassess PPEs after obtaining a license (for example, if PPEs were part of the decisionmaking on the license and a referenced design is changed, is there a mechanism for reanalysis?); and,

4) the generic sco>e of 15e PPEs against the site specific situation (for example, man-made lazards considerations that are peculiar to a site). A number of these concerns were addressed by the ISG staff in the detailed discussion that followed on the purpose for and methodology for developing PPEs. ,

The ISG staff outlined the methodology developed to construct the PPE sets.

On a reactor design-specific basis, the plant 'rameters were those that specify either 1) a functional or operational mquirement that consumes resources, 2) a design capability to withstand hazards, or I; a direct impact on resources from operations. The interface (plant / site) parameters were those not directly specified for a reactor design, but could be generically applicable and reasonably quantified. The site and detail parameters were those that would require consideration of cite specific conditions.

In all, 29 categories were established for the plant and plant / site parameters that comprise SSCs and other germane characteristics (i.e., plant, severe accident, and construction) for each design. Some of these category titles were redundant for systems that performed similar functions but were uniquely titled for a specific reactor design. When questioned by the NRC staff of the significance of the category binning process presented, the ISG staff suggested that the categories were merely organized in the manner presented because of convenience and logical groupings.

The ISG indicated that only the plant and plant / site parameters were necessary for inclusion in the PPE sets for use by ESP applicants. The site and detail parameters would need to be defined by the applicant for the specific site and, therefore, would be inappropriate for inclusion in the PPE sets. The 29 categories were then screened to determine applicability of the specific parameters to each of the reference ALWR designs. The values assigned to each of the applicable parameters were obtained from the ALWR Utility Requirements Document (URD) and the vendor specific design information documents. Absent a 4

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l referable value from the URD or the vendor, an engineering assessment would be performed to generate a value for the parameter. The NRC staff raised a concern regard ag the quality assurance aspects of compiling the PPEs and performing the engineering assessment. Finally, composite (evolutionary, passive, and overall) PPEs would then be generated by selecting the limiting condition for each parameter. The ISG suggested that the composite PPEs are useful to those prospective applicants that have not narrowed their selection j to a specific design.

j In stating the purpose of the PPE, a number of concerns rai ed earlier by the staff were addressed. The ISG staff and DOE staff indicateo that the PPE is a tool specifically intended to permit the evaluation of the plant's impact on the site and is useful only for the pu,rposes of obtaining an ESP. The ISG l staff indicated that there would be no ongoing significance of the PPE since a prospective combined license (COL) applicant either would need to demonstrate that the reactor fits within the parameters included in the ESP or would need to resolve the matter as part of the COL licensing process.

l The ISG staff emphasized that the PPE document should assist a prospective applicant in developing the set of information necessary to permit an assessmentofthesuitabilityofasitetohostoneormorenuclearpower i

plants. The ISG recognized that the design information necessary for an ESP application may not be universally available and, consequently, developed the set of PPEs as a resource. Also, the ISG staff emphasized that the PPE document cannot contain all of the information necessary for an ESP application and site specific tailoring will be necessary to assess the site's impact on the plant.

Additionally, the ISG staff indicated that an ESP application should be self contained. The ISG intended to outline the methodology it followed in creating the PPE document to make NRC and the public aware of its existence as a resource. The NRC staff reiterated from earlier public meetings that it was not seeking to review the PPE document. Nevertheless, the staff expressed interest in the plans for completion of the document (currently scheduled for late 1992) and its distribution to interested parties. At this time, EPRI is responsible for the control of the information contained in the PPE document; ownership of the document may be transferred.

1 In closing out the discussion of the ISG technical and regulatory review, the ISG outlined the opportunity for limited work to be conducted at a site. In order for an applicant to conduct the activities provided in i 50.10(e)(1),

the applicant could propose a redress plan for NRC consideration in the ESP application. The ISG staff described the purpose for this plan as achieving an environmentally stable and aesthetically acceptable site suitable for an alternate use if the authorized activities were performed and the site is not used to host a nuclear power plant, i

This meeting was the 1.)st of the targeted meetings under consideration )rior to the October 14 1992, siting conference Q at is being sponsored by tie ISG and the DOE early site permit demonstration program. The ISG and NRC staffs indicated that the targeted meetings were useful in raising issues that still need to be resolved.

No followup actions are 11anned on the referenced documents submitted by the ISG in preparation for tie targeted meetings.

BarryZbcman,SeniorProjectManager 4

License kenewal and Environmental Review Project Directorate Associate Directorate for Advanced Reactors and License Renewal

Enclosures:

l. Attendee's List *

2. Meeting Handout DISTRIBUTION:

Central File NRC PDR PDLR-R/F TMurley/FMiraglia,12G18 JPartlow, 12G18 DCrutchfield, llH21 WTravers, llH21 FAkstulewicz, llF23 BZalcman, llF23 LLuther, 11F23 'GC

. EJordan, MNBB3701 ACRS (10) -MTaylor, 17G21 LSoffer, NLS324 WRussell, 12G18 JCraig, llF23 FCongel,10E2 LCunningham, 10D4 GGrant, 17G21 JMoore, 15B18 DCleary, NLS314 AMurphy, NLS217A PSobel, 7H15 TEssig, 1004 RRothman 7H15 TKing, NLS007 CAder, NLS324 GBagchi, 7H15 GMizuno, 15B18 JNWilson, 11H3 MLopez-Otin, 3D23 SDroggitis, 3D23 FKantor, 9H15 RErickson, 9H15 PDLR:LA . PDLR: -

LLuthec48- BZ n JCry 5 11// o/92 -11/l0/92 II/{ W 2 10/5/92 Meeting summary

ENCLOSURE 1 NRC HEETING WITH THE INDUSTRY SITING GROUP i

• TARGETED SITING ISSUES - SITE SAFETY AREA AND PLANT PARAMETER ENVELOPE HETH000 LOGY ONE WHITE FLINT NORTH 16-B.11 October 5, 1992 1:00 P.M.

NAME AFFILIATION B. 2aleman NRC/NRR/PDLR 4

i.

F. H. Akstulewicz NRC/NRR/PDLR j

G. S. Hizuno NRC/0GC l R. W. Bishop NUMARC J. P. Ronafalvy , NUMARC S. T. Gray EPRI D. Powell EPRI H. H. Finkelstein NRC/0GC J. Schmitt NUMARC

! W. Pasedag DOE L. Soffer NRC/RES/SAIB J. Love Bechtel S. J. Cereghino Bechtel E. F. Fox NRc/NRR/PEPB J. Y. Lee NRC/NRR/PRPB i D. J. Zannoni NRC/OSP P. C. Cota NRR/PHAS i

J. N. Wilson NRC/NRR/PDST R. L. Rothman NRC/NRR/ESGB C. E. Ader NRC/RES/SAIB J. J. Hayes NRC/NRR/PRPB G. Bagchi NRC/NRR/ESGB C. Snyder NUS-Halliburton

, C. Marco NRC/0GC H. Newsome NRC/0GC

l ENCLOSURE 2 s

Agenda NRC/ Industry Meeting on Early Site Permh Safety Analysis Report Requirements October 5,1992 Introduction and Opening Remarks John Schmitt ESP, Safety Analysis Report John Ronafalvy e Regulatoiy Bases .

• ESP Application Requirements ESP, Site Redress Plan t

Plant Parameters Envelope Susan Gray Summary John Schmitt y

i i

EARLY SITE PERMIT SAFETY ANALYSIS l

REPORT REQUIREMENTS OCTOBER 5,1992 John Schmitt Nuclear Management and Resources Council, Inc.

l l

EARLY SITE PERMIT FOCUSED MEETINGS I

O ajective:

• To demonstrate to October 1992 con"erence participants that the ,

industry las acquired t1orough and accurate understanding of the NRC's requirements for the Early Site Permit process l

i

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! EARLY SITE PERMIT 1

SITE ANALYSIS REPORT FOCUSED MEETING 1

i

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1 l

I e Discuss mutual understanding of site l and plant interface safety analysis report needs for ESP applications.

! e Discuss the use of plant narameters envelope (PPE? for the preparation of an ESP application.

4 l .

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l' i-I EARLY SITE PERMIT FOCUSED MEETING i

j 1

5 l Proposed 10 CFR Part 100 revision

• Status of industry efforts to date:

l l

Seismic criteria and methodology l

Non-seismic area:

l -

population density criteria l -

5 year reporting of external l hazards Appendix' Q, ESP renewal

! language definition l

Exclusion zone criteria

i 4

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4 EARLY SITE PERMIT SITE SAFETY ANALYSIS REPORT I OCTOBER 5,1992 i

i i John Ronafalvy i Nuclear Management and l Resources Council, Inc.

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l EARLY SITE PERMIT i SITE SAFETY REPORT l

Heaulatory Bases '

l Public Law: Atomic Energy Act

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NRC Regulations: 10 CFR Part 50 10 CFR Part 52

! 10 CFR Part 100 4

NRC Regulatory Guides: R.G. - 1.70 4

R.G. - 4.7 Standard Review Plans: NUREG-0800 4

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EARLY SITE PERMIT SAFETY ANALYSIS REPORT REQUIREMENTS I

• An ESP application will contain safety analysis report (SAR) technical information pertaining to site characteristics (Chapter 2) and the assessment of site features affecting the plant design; major systems, structures, and components that bear significantly on site acceptability (portions of Chapter 3).

If the applicant wishes to ensure that any ALWR design could be licensed for the site at COL, the relevant plant parameters envelope (PPE) could be used to verify site suitability.

l This industry document contains surrogate plant design information, obtained from actual ALWR designs, supplied by the NSSS vendors. The PPE presents this data for use in performing conservative analyses for determining site suitability. The objective is to determine with finality at ESP stage that a site could successfully l host a standard ALWR(s).

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Early Site Permli Demonstration Program A5 _

Plant Parameter Envelope Development Process

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Tre she is approved for verity the etabon parameters tJture constructen and ' are wthin the PPE sp+cded operston of statene eth in the ESP and consider any parameters that are wihin ,,. ,,. . . , ognifcant new iniormaton.

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SITE REDRESS PLAN i

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• If applicant wishe's to perform site

activitie's allowed by 10 CFR 50.104

eX1D, the applicant shall

propose in the ESP application a plan

for redress 6f the site.

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• The redress plan will achieve an environmenDly stable and

! aesthetically acceptable site suitable l for whatever non-nuc ear use may

conform with lo' cal zoning laws.

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Plant Pararneters Envelope for Early Site Permitting

S.T. Gray, EPRI l

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i il Early Site Permit Demonstration Program ^E -

i Technological Plant / Site Interfaces 4 '

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Number of Geography Reactors

Demography Type of Reactor

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Level Geology "

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I l Summary of Methodology 4

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• Compile list of qualitative parametets i characterizing the plant's:

! -functional or operational needs from the .

! site's natural / environmental resources l capability to withstand the natural or i man-mac e environmental hazards inherent

! to a site l - direct impact on the site's natural /

l environmental resources ,

• Develop quantitative parameters lists by using:

) - ALWR Utility Requirements Documents l - Vendor design .information

- Engineering Assessments l

• Analyze Individual parameter lists to create

generic envelopes by selecting limiting j values for

- Evolutionary Plant

- Passive Plant

- Composite i

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PLANT PARAMETERS SITE PARMiETERS Plant functional or operational Site's capacity to satisfy needs (e.g. maximum normal raw < plant needs (e.g. raw water

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water now rate to the circula- available from site or near-ting water system by water sources)

Plant capability to withstand Site specine hatards (e.g.

natural or manmade envirqn- > site specine carthquake mental hazards (e.g. design ground acceleration) basis safe shutdown earth-quake ground acceleration)

Plant's direct impact on the Site's capacity to accommo-site's natural /ensironmental 3 date plant operation (e.g.

resources (e.g. circulating federal, state, and local water system thermal efnuent) discharge limit ations l

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l Qualitative Plant Parameters List i

Plant Those parameters that firmly specify the reference plants' requirements for a to resources, i design capacity to withstand site conditions or i affluents to the site and environs. (Example l tornado wind speed)

Plant / Site Those parameters that describe typleal nuclear

(P/S) plant features interfacing with the site that are not specified !n the reference plant designs and are not of such site variability as to sreclude development

of reasonable quantitles for nelusion in the plant e

alope. (Examplet cooling tower height) l 1 Site Thosrs parameters specifying alte conditions which have a direct significant impact on plant SSC's but

! are of such site specific varlability that customized analyNs or design features are necessary to ensure suitability with the ALWR plants.

I Detall Those parameters specifying design details of plant i SSC's that generally can be designed to accommodate a variety of site specific conditions, or whose requirements or impacts can be stated in terms of another SSC with a more direct intadace with the site PLANT and P/S represent the parameters necessary to describe the ALWR plant designs for an Early Site Permit Application l

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l l PARAMETERS APPLICABILITY MATRIX LIST

3. STRUCTURES

2. NORMAL PLANT HEAT SI!1K

3. ULTIMATE HEAT SINK

4. CONTAINMENT HEAT REMOVAL SYSTEM (POST-ACCIDENT)

5. POTABLE WATER / SANITARY WASTE SYSTEli '

6. ' DEMINERALIZED WATER SYSTEM

7. TIRE PROTECTION SYSTEM

8. MISCELLANEOUS DRAIN

9. LTNIT VENT / AIRBORNE ETTLUENT RELEASE POINT

10. LIQUID RADWASTE SYSTEM i

11. GASEOUS RADWASTL SYSTEM

12. SOLID RADWASTE SYSTEM

13. REACTOR COOLANT SYSTEM

14. RCS CLEANUP SYSTEM

15. CVCS LETDOWN SUBSYSTEM

16. CVCS PURITICATION SUBSYSTEM

17. CVCS SHIM / BLEED SUBSYSTEM

18. SPENT FUEL STORAGE

19. STEAM GENERATOR BLOWDOWN SYSTEM

20. STANDBY-GAS TREATMENT SYSTEM

22. AUXILIARY BOILER SYSTEM

22. CONDENSATE CLEANUP SYSTEM

23. GAS STORAGE SYSTEM

24. HEATING, VENTILATION AND AIR CONDITIONING SYSTEM

25. ONSITE/OTFSITP. ELECTRICAL POWER SYSTEM

26. STANDBY POWER SYSTEM

27. SEVERE ACCIDENT TEATURES

26. PIJJ4T CHARACTERISTICS

29. CONSTRUCTION l

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l Plant Appi cab lity Matrix l .

i i The alant ap alicability matrix uses the

! qual tative p ant parameters list to Identify the applicability of the

parameters to the reference ALWR designs.

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• For each PLANT and P/S parameter, a

. Parameter Quantity Summary (PQS) sheet is

larepared. Each PQS provides the following j .

ntc,rmation: ';

j - Structure, System, Component l - Parameter identification

! Datameter Definition

% meter Units

, meter Design Envelope Parameter ss I

"' ca a parameter value is defined by the UHD or vendor Information, then that value is included.

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• When no URD or vendor value exists, engineering assessments are performed.

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l 2. Assign a responsible engineering discipline.

3. Collect data - potential sources are:

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- FSARs and ERs

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- NUREG-1437

4. Evaluate data and determine an approach

5. Peer review and pro;ect review.

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lable 33 PARAMETER QUANTITY SU}OtARY h*VAC SYSTEMS Structure, System, Component PRIMARY PARAMETER: Ambient Air Requirements SECONDARY PARAMETER: Vent System Max. Ambient Temp. (5% exceed) (24.1.6)

Vent System Min. Ar.bient Temp. (5% exceed) (24.1.7)

's DEFI NI'.'I ON : The maximum and minimum ambient temperature for design of ventilation systems (no air conditioning) based upon 5% exceedance values.

UNITS: Degrees Fahrenheit i

DESIGN ENVELOPE:

BASIS: URD Value a u nit:.ecs ere w ee sy: ce e:

S ev n tet" $ y; Ca*t-

i Quantitative Parameter Envelopes 9

The following " rules" were used in

determining the appropriate values to Insert in the envelopes

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• For any parameter with a value taken from the URD, the URD value is always used asi the envelope value

• In cases where not allvendors l provided values, engineering

! assessments are performed. When l both an engineering assessment and vendor value exist, the vendor value is used as the plant envelope value.

• For any parameter with neither an URD value nor vendor value, the value I develo 3ed by engineering assessment i

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i

EARLY SITE PERMIT 4 SAFETY ANALYSIS REPORT

SUMMARY

?

• The Plant Parameters Envelope (PPE) l document, developed by the Early Site Permit Demonstration Program, is an i industry tool to assist an ESP l applicant b providing plant design information that can be used in ESP

! stage site safety and environmental suitability determinations required by NRC regulations. The PPE was developed to ensure future viability to license (COL) fo'r the construction and l operation of ALWRCs? on the site  :

where an ESP had been previously obtained.

1

, EARLY SITE PERMIT 10 CFR PART 100

SUMMARY

"2

• The NRC is encouraged to adopt an interactive approach, yvith opportunity for dialogue, in obtaining public comments for the proposed siting-related regulatory revisions.

Codification of guidance criteria provided for

in the revision could establish far reaching i and restrictive'public policy on nuclear plant sitings in the future, apparently without technical bases. Such regulations, if adopted, will increase the public's perception of risk 4

(already much higher than actual risk) '

i associated with future and operating nuclear power plants.

• The industry, through NUMARC, will file written comments regarding the siting-related regulatory revisions currently issued for public comment.

., - - - - - -