ML20077J505

Proposed Initial Decision,Findings of Fact & Conclusions of Law Re Cp.Certificate of Svc Encl

ML20077J505
Person / Time
Site:	Clinch River
Issue date:	08/15/1983
From:	Edgar G, Luck W ENERGY, DEPT. OF, JOINT APPLICANTS - CLINCH RIVER BREEDER REACTOR, PROJECT MANAGEMENT CORP.
To:
References
NUDOCS 8308160465
Download: ML20077J505 (173)

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agy' AW 15igp/15/83 g7 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISS Thw[5%.f,

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In the Matter of )

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UNITED STATES DEPARTMENT OF ENERGY )

PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537 TENNESSEE VALLEY AUTHORITY )

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(Clinch River Breeder Reactor Plant) )

APPLICANTS ' PROPOSED INITIAL DECISION (CONSTRUCTION PERMIT)

George L. Edgar Thomas A. Schmutz Frank K. Peterson Morgan, Lewis & Bockius 1800 M Street, N.W.

Washington, D. C. 20036 Attorneys for Project Management Corporation Leon Silverstrom William D. Luck U. S. Department of Energy Office of General Counsel 1000 Independence Avenue, S .W.

Room 6B-2 56 - - Forrestal Bldg.

Washington, D. C. 20585 Attorneys for United States Department of Energy Herbert S . Sanger, Jr.

Lewis E. Wallace W. Walter LaRoche James F . Burger Edward J. Vigluicci DATED: August 15, 1983 Tennessee Valley Authority 400 West Summit Hill Drive i

Knoxville, Tennessee 37902 Attorneys for the Tennessee Valley Authority

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GLOSSARY. -

ACRS Advisory Committee on Reactor Safeguards AEC Atomic Energy Costmission A Exh Applicants' Exhibit A'SME American Soci.ety of Mechanical Engineers AW Applicants' Witness CE Commonwealth Edison Company CDPL Core Disruptive Accident .

CP Construction Permit i

CRBRP Clinch River Breeder Reactor Plant CS Containment System DBA Design Basis Accident DHRS Direct Heat Removal Service DOE Department of Energy DRP Developmental Reprocessing Plant EBR-II Experimental Breeder Reactor II EPA Environmental Protection Agency

.EPZ Emergency Planning Zone i

ERDA Energy Research and Development Administration l

i ETEG East Tennessee Energy Group l

FES Final Environmental Statement FFTF Fast Flux Test Facility ,

FMEA Failure Mode and Effects Analysis Final Supplement to the Final Environmental Statement l

FSFES

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HCDA Hypothetical Core Disruptive Accident l

HTS Heat Transport System l

IEEE Institute of Electrical and Electronic Engineers IHTS Intermediate Heat Transport System IHX Intermediate Heat Exchanger J JCAE - Joint Committee on Atomic Energy LMFBR Liquid Metal Fast Breeder Reactor LOF Loss of 71ow LOF-d-TOP Loss of Flow Driven Transient Overpower LOHS Loss of Heat Sink LNA Limited Work Authorization LWR Light Water Reactor MJ Megajoules NEPA National Environmental Policy Act NRC Nuclear Regulatory Commission NRDC Natural Resources Def ense Council, Inc.

NSSS Nuclear Steam Supply System OL Operating License ORGDP Oak Ridge Gaseous Diffusion Plant ORNL Oak Ridge National Laboratory l

PAG Protective Action Guide j PHTS Primary Heat Transport System PID Partial Initial Decision l PMC Project Management Corporation l PPS- Plant 'Prote'ction System l

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PRA Probabilistic Risk Assessment ~

PSAR Preliminary Safety Analysis Report 1 psia pounds per square inch absolute psig pounds per square inch gauge PSST Primary Sodium Storage Tank

• QA ' Quality Assurance QC 'Cdality; Control RAF Reliability Assurance Program RCB Reactor' Containment Building

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RSS Reactor Shutdown S? stems '

SER Safety Evaluation Report , -

S Exh Staff's Exhibit SGAHRS Steam Generator Auxiliary Heat Removal System SHRS Shusdown Heat Removal System SMBDB Structural Margin Beyond the Design Base SR Savannah River ~ '=

SSR Site Suitability Report SSST Site Su'ltability Source Term SW Staff's Witness i SWRPRS Sodium Water Reaction Pressure Relief System TMBDB Thermal Margin Beyond the Design Base TMI Three Mile Island .

TOP TransientOv$rpower S TR Transcript Page TREAT Transient Readtor Test TVA Tennessee Valley Authority, 3 TWRA Tennessee Wildlife Resources Agency

TABLE OF CONTENTS

.i Page I. INTRODUCTION ......................................... 2 1

II. OPINION ......................................... 15 A. WHETHER AN HCDA SHOULD BE A DBA.................. 16 B. THE ADEQUACY OF APPLICANTS'

. . AND STAFF'S HCDA ANALYSES........................ 22

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C. BOARD AREAS OF INTERE3T.......................... 26 Board Area of Interest 1...................... 27 Board Area of Interest 2...................... 27 Board Area of Interest 3...................... 28 Board Area of Interest 4...................... 28 Board Areas of Interest 5 and 6............... 29 Board Area of Interest 7...................... 32 Board Arca of Interest 8...................... 33 Board Area of Interest 9...................... 36 Board Area of Interest 10..................... 37 Board Area of Interest 11..................... 37 l Board Area of Interest 12..................... 38 Board Area of Interest 13..................... 39 Board Area of Interest 14..................,.. 40 Board Area of Interest 15..................... 41 Board Area of Interest 16..................... 41 Board Area of Interest 17..................... 42

Page D. OTHER MATTERS.................................... 42

1. NRDC's Statement.............................. 42

2. Emergency Planning at Nearby DOE Facilities................................ 43

3. Incorporation of PRA Results.................. 43

4. Additional Matters............................ 43 III. CONCLUSION . . . . . . . . . . . . . ....................,....... 43 - '

IV. ORDER . . . . . . . . . . . . . ............................ 46 V. FINDINGS OF FACT...................................... F-1 THE CRBRP DESIGN SAFETY APPROACH...................... F-2 WHETHER AN HCDA SHOULD BE A DBA....................... F-5 Reactor Shutdown Systems......................... F-15 Shutdown Heat Removal Systems.................... F-16 Means to Prevent Rupture of Primary Heat Transport System Inlet Piping.............................. F-18 Features and Capabilities to Prevent Local Imbalance Between Heat Generation and Heat Removal................. F-20 THE ADEQUACY OF APPLICANTS' AND STAFF's HCDA ANALYSES......................................... F-26 l

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Board Area of Interest 1......................... F-34 Board Area of Interest 2......................... F-37 Board Area of Interest 3......................... F-39 Board Area of Interest 4......................... F-41 Board Area of Interest 5......................... F-42 Board Area of Interest 6......................... F-48 Board Area of Interest 7......................... F-49 Board Area of Interest 8......................... F-51 Board Area of Interest 9......................... F-61 Board Area of Interest 10........................ F-63 Board Area of Interest 11........................ F-64 Doard Area of Interest 12........................ F-65 Board Area of Interest 13........................ F-67 Board Area of Interest 14........................ F-69 Board Area of Interest 15........................ F-70 Board Area of Interest 16........................ F-72 Board Area of Interest 17........................ F-72 OTHE R MATTE RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F -7 6 Thyroid Doses.................................... F-80

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HCDA Initiation Probability...................... F-82 Savannah River and Fort St. Vrain Reactors......................................... F-83

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Page Emergency Planning at Nearby Facilities................................ F-84 PRA Results...................................... F-87 Additional Matters............................... F-88 VI. CONCLUSIONS OF LAW.................................... C-1 APPENDIX A - BOARD AREAS OF INTEREST -

APPENDIX B - EXHIBIT LIST APPENDIX C - WITNESS LIST .

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I i l l 8/15/83 i

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD t f

In the Matter of )

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UNITED STATES DEPARTMENT OF ENERGY )

)

PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537

)

TENNESSEE VALLEY AUTHORITY )

)

(Clinch River Breeder Reactor Plant) )

APPLICANTS' PROPOSED INITIAL DECISION (CONSTRUCTION PERMIT)

The United States Department of Energy (DOE) and Pro-ject Management Corporation (PMC), for themselves and on behalf I

of the Tennessee Valley Authority (the Applicants), hereby file this Proposed Initial Decision for a Construction Permit pursuant to 10 C.F.R. $ 2.lO4(b)(2) and I 50.35. This Proposed Initial Decision is presented in six parts: I) Introduction, II) .

Opinion, III) Conclusion, IV) Order, V) Findings of Fact, and VI) 1 Conclusions of Law. /

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1/ Enclosed also' is a set of Appendices to the Proposed Deci-sion consisting of: A) Board Areas of Interest, B) an Exhibit List, and C) a Witness List.

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l I. INTRODUCTI'N O This Initial Decision concerns the application filed with the United States Nuclear Regulatory Commission (the Commission or NRC) by the United States Department of Energy j

(DOE), Project Management Corporation (PMC), and the Tennessee Valley Authority (TVA) for a Construction Permit (CP) and Operating License (OL) for the Clinch River Breeder Reactor Plant (CRB RP ) .

The CRBRP is a Liquid Metal Past Breeder Reactor (LMFBR) demonstration plant with a rated output of 350 megawatts of net electrical power, to be located on the Clinch River in Oak Ridge, Tennessae. A Exh 86; S Exh 26 at 1-4, 1-12.2 / On February 28, 1983, this Board issued a Partial Initial Decision (PID) addressing the portions of the application for a Construction Permit which are necessary for Limited Work Authorization (LWA) findings under 10 C.F.R. $ 50.10 (e ) ( 2 ) ,

2/ Citations to the record hereia are in the following form:

a) Applicants' Exhibit - A Exh; Staff's Exhibit - S Exh; b) Applicants' Witness - A W: Staff's Witness - S W:

c) Transcript - TR d) Citations to prefiled written testimony will include citations to exhibit number, page number, and transcript page.

e)

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, The Board's February 28, 1983 Partial Initial Decision (Limited Work Authorization), United States Department

-LBP of Energy (Clinch River Breeder Reactor Plant),

, NRC (1983) - PID.

f) Board Exhibit - B Exh l

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i namely, findings on all pertinent radiological site suitability and environmental issues. This Initial Decision addresses the remaining portions of the application which are necessary for grant of a Construction Permit.

4 The CRBRP was first authorized by the Congress in 1970 as a cooperative effort between industry and government to design, construct, and operate the Nation's first demonstration-scale LMFBR. Pub. L. No.91-273, Section 106. In early 1972, the Atomic Energy Commission (AEC) accepted a joint proposal by the Commonwealth Edison Company (CE) of Chicago and the TVA to under-take the design, construction and operation of the demonstration plant as part of the TVA electric system. Under this proposal, PMC, a non-profit corporation organized and existing under the laws of the District of Columbia, had the overall lead management responsibility for the CRBRP, TVA would operate it, and the AEC had lead technical responsibility for the nuclear reactor systems. More than 750 electric systems in the United States 3/ Pub. L. No. 92-84. See Joint Report of the House Committee on Science and Technology and the Joint Committee on Atomic Energy (JCAE) , 94th Cong., 1st Sess., H. Rep. No.92-294 at 32-35 (1975) [ hereinafter, Joint Report]; JCAE Authorization Report, 94th Cong., 1st Sess., S. Rep. No.94-104 at 17-20 (1975) -[ hereinafter, JCAE Report].

4/- See Report on Hearings before the Joint Committee on Atomic Energy on the Basis for the Proposed Arrangement for the LMFBR Demonstration Plant, 92d Cong., 2d Sess. (Sept. 7, 8, and 12,1972) [hereinaf ter, JCAE Hearings] at IV-V. See also Report on Hearings before the JCAE.to Consider Proposed

, Changes in the Basis for-the Cooperative Arrangement for Design, Construction, and Operation of the LMFBR Demonstra-i- tion Plant, 93d Cong., 1st Sess. (Feb. 28 and May 4, 1973).

c have pledged more than $250 million in financial payments which are applied to the project by PMC.

l In October 1974, PMC and TVA jointly filed an appli-L

cation with the AEC for a Construction Permit and Operating License for the CRBRP pursuant to Section 104(b) of the Atomic Energy Act of 1954, as amended, 42 U.S.C. S 2011 et_ seq. After the Energy Reorganization Act of 1974, 42 U.S.C. S 5801 eji j0tg. , .

transferred the developmental and regulatory functions of the AEC to the Energy Research and Development Administration (ERDA) and the NRC, respectively, the NRC assigned the application to its docket for review on April 11, 1975.

On June 18, 1975, receipt of the application and pro-ceedings before the Atomic Safety and Licensing Board (the Board) were noticed. A timely joint petition for leave to intervene was filed by the Natural Resources Defense Council, Inc., the Sierra Club, and the East Tennessee Energy Group (Intervenors),

I and on October 9, 1975, the petition was granted by this Board.

Af ter the East Tennessee Energy Group (ETEG) had become defunct, the Intervenors requested the withdrawal of ETEG as a party on February 8, 1982, and on February 11, 1982, the Board granted the request. The Natural Resources Defense Council, Inc. and the Sierra Club remained as joint intervenors in the LWA proceedings.

5/ Id.

jij/ 40 Fed. Reg. 25708 (June 19,1975) .

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t The State of Tennessee Attorney General filed a timely petition for leave to intervene and was admitted as a party on October 9, 1975.

On March 29, 1982, the State of Tennessee Attorney General filed a Motion to Withdraw as a party under 10 C.F.R. $ 2.714, but requested leave to continue participating as an " interested state" under 10 C.F.R. $ 2.715. The motion was granted on March 31, 1982.

The City of Oak Ridge filed a petition for leave to intervene on July 17, 1975, amended that petition on January 22, 1976, and was admitted as a party on March 4, 1976.7 / On August 20, 1982, the City of Oak Ridge requested leave to withdraw as a party to the proceeding and to continue participating as an

" interested municipality" under 10 C.F.R.

I 2.715(c). On September 7, 19 82, the Board granted the motion.

On May 6,1976, pursuant to authorization contained in the 1976 amendments to Pub. L. No.91-273, as amended, the appli-cation was amended to include ERDA as a co-applicant (with PMC and TVA),

and to reflect the realignment of the respective Project participants' roles. Under this realignment, ERDA ass umed 7/

~~ Roane County, which was admitted as a party by the Board's Order, dated October 9,1975, was granted leave to withdraw

from1976.

13, all participation by the Board's Order, dated December counties and municipalities was denied by the Board onThe l

August the Appeal 26,Board.

1976, and on appeal, the denial was af firmed by River Breeder Reactor Plant), Project Management Corporation (Clinch l

af f ' d, ALAB-354, 4 NRC 383 (1976). LBP-76-31, 4 NRC 153 (1976),

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the lead management role in the integrated CRBRP Project Office, which included PMC and TVA personnel, and TVA remained as the operator. See Joint Report at 35; JCAE Report at 19; 122 Cong.

Rec. S10613-22 (June 25,1976); 122 Cong. Rec. H5835-5898 (June 8

15, 1976). DOE is the successor-in-interest to ERDA. /

Commencing in November 1975, extensive prehearing activities ensued,9 / and by March 1977 the NRC Staf f had issued a

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Site Suitability Report (SSR) and Final Environmental Statement  ;

l (FES). S Exh 23. On March 28, 1977, the Board issued an Order

for commencement of LWA hearings in Oak Ridge on June 14, 1977, which hearings were to run continuously until completion.  !

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On April 20, 1977, the previous Administration announced its decision to cancel the project. On April 22, 1977, ERDA filed a motion to suspend the proceedings, and on April 25, 4

19 77, the Board issued an Order granting that motion. In 8/ See 42 U.S.C. $ 7101 et seq.

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9/ Intervenors filed fif teen sets of interrogatories, seven j sets of requests for admissions, and four requests for pro-l duction of documents against the Applicants. Intervenors filed twenty-two sets of interrogatories, seven sets of requests .for admissions, and three requests for production of docu.ments against the NRC Staf f. An appeal arose con-cerning the admissibility of two Intervenor contentions, and the Commission held that certain programmatic issues pre-viously considered in ERDA's LMEBR Program Environmental Statement would not- be reconsidered in the CRBRP licensing

. proceedings. See United States Energy Research and Development Administration (Clinch River Breeder Reactor Plant), CLI-76-13, 4 NRC 67 .(1976) .

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November,1977, the NRC Staf f suspended its review of the application.

In the ensuing four year period, the project continued design, research and development, and procurement activities, while licensing activities remained suspended. In each of those years, Congress acted to preserve the project by providing sub-stantial f unding.

In August 1981, the President signed the Omnibus Budget Reconciliation Act of 1981, Pub. L. No. 97-35, which expressed the intention that the project be expeditiously completed. 11 / In a Nuclear Policy Statement of October 8,1981, the President directed that " government agencies proceed with a demonstration of breeder reactor technology, including completion of the Clinch River Breeder Reactor." 17 Weekly Comp. of Pres. Doc. 1101-1102 (October 12, 1981).

On January 11, 19 82, the Applicants filed a motion to lif t the suspension of hearings, and on January 19, 1982, the Board granted this motion and issued a Notice of Prehearing Con- I ference. On February 9-10, 1982, the Board held a prehearing conference, and on February 11, 19 82, issued an Order 10/ Pub. L . No.95-240, March 7, 19 78 ; Pub. L . No. 9 5-482, October 18, 1978; Pub. L. No. 96-8 6, October 12, 1979 ; Pub.

L . N o. 9 6-3 67 , October 1, 19 80; Pub. L . No. 9 6-5 36, December 16, 1980; Pub. L. No. 97-12, June 5, 1981.

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""~ See H. Rep. No.97-208, 97th Cong. 1st Sess. (1981);' 127 155sg. Rec. S8998 (1981); 127 Cong. Rec. H5817-18 (1981).

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establishing a schedule for all activities necessary for commencement of evidentiary hearings concerning LWA matters on

- August 23, 1982.

Pursuant to the Board's February 11, 1982 Order, all contentions related to the CP application were identified. The Intervenors restated and/or revised their original contentions, and filed additional contentions based upon new information.

Upon censideration of the pleadings filed by the parties and two sets of prehearing meetings with the parties, the Board issued two Orders Following Conference with Parties, dated April 14 and April 22, 1982, which ruled upon the admissibility, scope, and applicability (LMA vs. CP) of Intervenors' contentions.

Extensive discovery ensued, and on June 11, 1982, i

the NRC Staf f issued its updated SSR (NUREG-0786), which con-cluded that the Clinch River site was suitable for a reactor of the general size and type described in the application from the standpoint of radiological health and safety. S Exh 22. On July 13, 19 82, the Advisory Committee on Reactor Safeguards (ACRS) issued a letter which supported the NRC Staf f's site suitability conclusion. S Exh 4.

12/ By April 30, 1982, Applicants and Staf f had updated their responses to Intervenors' 1975-77 discovery. As of the close of discovery on June 30, 19 82, Intervenors had also filed an additional four sets of interrogatories, four sets of requests for admissions, and three requests for production of documents, and had deposed five persons from the NRC Staf f and eleven persons from the Applicants.

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9-On July 19, 19 82, the Board issued a Notice of Eviden-tiary Hearing and Prehearing Conference, which ordered that hear-ings commence in Oak Ridge on August 23, 19 82, and continue until completion of taking evidence on the issues and contentions admitted for the purpose of a limited work authorization hearing pursuant to 10 C.F .R. { 5 0.10 (e ) . On July 19, 1982, the NRC Staf f issued and served upon all parties to the proceeding its

update to the 1977 FES. In issuing that document, the NRC Staf f determined that it should be issued as a Draf t Supplement to the 1977 FES, and that it should be recirculated for public comment before issuance as a Final Supplement.

As a result of the decision to recirculate the FES, the schedule for hearings contemplated by the Board's February 11, 1982 Order could not be met. Upon consideration of motions filed by Applicants and Intervenors, and af ter hearing extensive argument during a conference with the parties, the Board issued

, an Order dated August 5,1982, which scheduled hearings on radiological site suitability issues (portions of Intervenors' contentions 1, 2, 3, and Intervenors' contentions 2e)/11 d)1) and j 2)), and ruled that hearings on environmental issues would await issuance of the Final Supplement to the FES.--- 13/

13/.

~"" In regard to this Order, Intervenors filed with the Appeal Board a Petition for Directed Certification, which was

. denied on August 25, 19 82. -United States Department of i

Energy (Clinch River Breeder Reactor Plant), ALAB-688 t

(Memorandum and Order, August 25, 19 82).

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l Site suitability hearings were completed during the  !

l week of August 23-27, 1982. The Board then reopened discovery on 14 all environmental issuer- / and set hearings for November 16-19, 1982, and December 13-17, 1982 to take evidence concerning the remaining environmental issues. Board Order dated August 31, 1982.

On Nover.ber 1, 1982 the NRC S taf f issued the Final Supplement to the FES (FSFES) . S Exh 24. Evidentiary hearings on the subjects encompassed by Intervenors' environmental conten-tions were held in Oak Ridge on November 16-19, 1982, and on 14/ Pursuant to this Order, Intervenors filed one additional set of interrogatories against the NRC Staff (27th Set), and took eight depositions of more than twenty Staf f witnesses.

15 Neither the State of Tennessee Attorney General nor the City of Oak Ridge participated actively in the LNA evidentiary hearings. By Order dated March 31, 1982, the Board granted the Attorney General's motion to withdraw as a party and F2.715( c ),rticipate and byasOrder an interested dated September State pursuant 7, 19to82, 10 C.F.R.

granted the City's motion to withdraw as a party and participate as an interested municipality pursuant to 10 C.F.R.

I 2. 715 (c ) . The Board received the " Position Paper of the Tennessee Attorney General on Socio-Economic Impact Matters and Other Matters Relating to the Clinch River Breeder Reactor Plant," dated November 10, 19 82, and "The City of Oak Ridge's Statement Relative to the Socio-Econceic Impact of the Clinch River Breeder Reactor Plant," dated November 12, 1982. At the direction of the Board (TR 33 56-58; TR 7104), the Applicants and Staf f filed, on January 11, 19 83, i 4

Responses to the Attorney General's Position Paper and the I City's S tatement. Neither the Attorney General nor the City conducted cross-examination, presented witnesses, or introduced documentary evidence concerning the socio-economic matters raised by their respective Position Paper

, and Statement. The Board's February 28, 1983 PID resolved the issues raised in the Position Paper and Statement. PID at 200-202.

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i December 13-16, 1982. Limited appearance statements were received from members of the public in Oak Ridge during the hearing sessions held August 23-27, 1982 and November 16-19, 1982.

Presentation of evidence on all LWA issues extended over the three hearing sessions for a total of thirteen days, and was completed on December 16, 1982. On December 16 and 17,1982, and on January 4 and 5, 19 83, the Board heard closing arguments from all parties, specifically addressing the record evidence and disputed issues as to all LNA contentions.

Af ter receipt of Proposed Findings of Fact from all parties on January 24, 19 83, the Board issued its February 28, 1983 PID, which addressed all pertinent radiological site suitability and environmental issues, and concluded, inter alia, that:

1) the Clinch River site is suitable for a reactor of the general size and type proposed in the CRBRP application from the standpo, int of radiological health and safety; 2) the contents of the Final Environmental Statement and the Final Supplement to the Final Environmental Statement (Staf f Exhs 23 and 24) were af firmed; 3) the requirements of NEPA and 10 C.F.R. Part 51 had been complied with in the proceeding; and 4) an LNA should be issued for the CRBRP pursuant to 10 C.F.R. $ 5 0.10(e ). PID at 198, 202. .

4 16/ By order of March 28, 19 83, the Commission itself determined that it would conduct the "immediate effectiveness" review of the PID. On May 5, 19 83, the Commission found that there was no reason to stay the eff ectiveness of the PID.

' Commission Order dated May 5, 1983.

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c.

By Order dated March 10, 1983 the Board opened discov- ,

ery on all remaining contentions in preparation for CP eviden-I tiary hearings. On March 11, 1983, the Staff issued its Safety Evaluation Report ( SER ) for the CRB RP . S Exha 26-28. On March 29, 19 83, the Board issued a Construction Permit Scheduling Order which established dates for the close of discovery, for filing motions of summary disposition, and for filing written direct testimony on all CP issues.

j During the discovery period, Intervenors filed responses to Applicants' and Staf f's discovery requests which I

indicated that they wished to withdraw their contentions 2f), g) -

and h ) , 9a), b), d ), and e ), 10, and lla ) . On May 17, 1983, the _ Board issued an Order granting Applicants' unopposed motions to dismiss those contentions.

On April 19, 19 83, the ACRS issued its report on the CRBRP CP application. The report concluded that, if the matters noted therein and the open items described in the SER were 17/ During this period Intervenors filed five sets of inter-rogatories and one document re quest. Applicants filed four l sets of interrogatories and reqpests for admissions, and conducted one deposition. The Staff filed five sets of interrogatories and requests for admissions, and conducted one deposition. . In addition, af ter the close of discovery, the Board granted Intervenors' request for additional discovery on the Staf f's HCDA dose calculations. TR 7188-7202.

i 18/ See Intervenors' April 19, 19 83 Response to Applicants' YTghth Set of Interrogatories and Intervenors' April 22, 1983 Response to NRC Staf f First Set of Construction Permit

-Interrogatories.

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resolved in a satisf actory manner, the CRBRP can be constructed i with reasonable assurance that it can be operated without undue risk to the health and safety of the public. S Exh 31. On May 2 and May 20, 1983, the Staf f issued its Supplements 1 and 2 to the CRBRP SER, respeccively, which resolved all open issues identified in the SER and ACRS Report. S Exhs 29 and 30.

On May 19, 19 83, Applicants moved for summary disposi- '

tion on Intervenors' contention 9 9), or.d for partial summary disposition on Intervanors' contentions 9c) and 9f). On June 29, 1983, the Board granted those Motions. TR 73 06.

On May 24, 1983 the Board issued a Notice of Construc-tion Permit Evidentiary Hearing, which provided, inter alia, that CP hearings would commence in Oak Ridge on July 18, 19 83 and consider: a) Intervenors' remaining contentions (1,3 and 9c) and f)); and b) seventeen specifically defined Board areas of inter-est. 48 Fed. Reg. 23944 (May 27,1983); see Appendix A hereto.

On June 21, 1983, Intervenors moved to withdraw all of their remaining contentions f rom consideration at the CP hear-ings, and reqpested permission to submit a written statement. At a June 29,1983 Conference with Counsel, the Board granted Inter-venors' notion and request, and dismissed Intervenors as parties to the proceeding. TR 7333. Conse quently, this proceeding became and remains uncontested, and only the Applicants and the NRC Staf f are parties to the proceeding. The State of Tennessee Attorney General and City of Oak Ridge remained as an " interested l

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state" and " municipality", res pectively, under 10 C.F .R.

$ 2.715(c), but neither participated in the CP proceeding.

On July 13, 19 83, the Board issued a Notice of

, Postponement of Hearing, and on July 19, 1983 issued a Notice rescheduling the CP evidentiary hearings for the week of August 8, 1983. S ee 48 F ed. Reg. 32711 (July 18, 1983); 48 Fed.

Reg. 39784 (July 25, 1983). The CP evidentiary hearings were held and completed in Oak Ridge during August 8 - 11, 1983.

The record of the CP hearings has focused upon the following areas of inquiry: a) Whether a hypothetical core dis- -

ruptive accident (HCDA) should be a design basis accident (DBA);

b) the adequacy of Applicants' and Staf f's HCDA analyses; c) the seventeen areas of interest identified by the Board in the 19 May 24,1983 Notice of Construction Permit Evidentiary Hearing- /

and reproduced in Appendix A hereto; and d) all other matters necessary to the Board's decision as to Whether a CP should be authori zed.

The decisional record in this proceeding consists of:

a) the Notice of Hearing; 19/ In-addition, the Board also considered: a) the matter of l evacuation of nearby DOE industrial f acilities in the event l of an accident at CRBRP (see PID, Finding 52); and b) the f easibility of implementing design and operational changes, if any, resulting from completion of Applicants' probabilis-L tic risk assessment (P RA) af ter ' CP issuance (TR 73 40-41) .

o-l 1

b) the material pleadings filed herein, and the Orders issued by the Board during the course of the proceeding; c) the Exhibits received into evidence as indicated in Appendix B hereto; and d) the transcript (witnesses who testified are listed in Appendix C hereto).

In making its findings in this proceeding, the Board i

considered the entire record and all of the proposed findings submitted by the Applicants and NRC S taff. Each of the proposed findings which is not incorporated directly or inf erantially in this Initial Decision is rejected as being- unsupported in f act or in law or as being unnecessary to the rendering of this Initial Decision.

i II. OPINION i This opinion will address: a) whether an HCDA should be a DBA; b) the adeqpacy of Applicants' and Staf f's analyses of HCDA's; c) each of the previously identified seventeen Board areas of interest; and d) other matters related to the Board's decision as to whether a CP should be authorized.

A. WHETHER AN HCDA SHOULD BE A DBA l The CRBPP has been designed with particular emphasis on i

prevention of core damage events. To this end, CRBRP includes -

two fast-acting reactor shutdown systems, each capable of l

l 1

e shutting the reactor down, and redundant and diverse shutdown heat reasoval systems carible of removing all decay heat through any one of four indeper. dent heat removal paths. Further, CRBRP includes measures in the design to preclude large leaks in primary system piping, and to prevent any local imbalances in core heat generation and removal from progressing to significant core-wide involvement. The Applicants and Staff have considered these preventive features and concladed that core damage events, or HCDA's, should not be DBA's for CRBRP. Findings 6-7.

Notvlthstanding these preventive features, CRBRP has l also been designed with specific features to accommodate and mitigate core damage events, including energetic HCDA's. In short, the CRBRP design incorporates not only the Commission's traditional three-level defense-in-depth concept, but has also placed particular emphasis on design for core damage prevention, and incorporates an additional level of protection censisting of i

l design features for mitigating the risk of HCDA's. Id.

i During the LWA proceedings, the Board placed limitations on the scope of inquiry as to whether HCDA's should be DBA's for the purposes of site suitability evaluation under 10 l C.F.R. S 50.10(3). Board Order dated April 22, 1982. In its February 28,.1983 decision the Board found that the consequences associated with the site suitability source term release represent an upper, or conservative, bound to all DBA's in

20 CRBPP. PID at 21, 81.-- / The Board also considered the evidence concerning the four categories of major CRBRP design features and characteristics to prevent the progression of an accident beyond design basis conditions to initiation of an HCDA. The Board found that these features can inhibit the initiation of an HCDA, and that these f eatures lend credibility to the proposition that HCDA's need not be included within the envelope of DBA's for CRBRP. PID at 19, 66-71.

In accordance with its previous rulings, however, the Board erarcised considerable caution to avoid attaching any preclusive or final ef fect to its findings on whether an HCDA should be a DBA, and most impor tantly, to avoid any prejudice to the rights of all parties to finally litigate that question at the CP proceedings upon completion of the full safety review.

PID at 22; see Board Order dated April 22, 19 82 at 4-7. While recognizing that there were no threshold matters which militate against exclusion of HCDA's from the design base, the Board foresaw a " heavy burden upon these parties at the construction permit evidentiary hearings to provide sufficient evidence to permit a resolution of this question." PID at 22.

20/

-~~ Fur ther, the Board found that the CRBRP containment /

confinement has been shown capable of performing its ,

intended f unction to accommodate all DBA's, and to hold doses below the applicable 10 C.F.R. I 100.ll(a ) site

. suitability dose guideline values. PID at 22, 81-82.

w, -*

• I

1. 4

-j We now consider the question of whether an HCDA should l be a DBA, af ter completion of the NRC Staf f safety review and on the basis of full detail. We do so in the face of a record which reflects an extraordinary depth of review by the NRC Staf f, and an unusually advanced level of design information for a CP proceeding. Further, we do so having had the opportunity to personally question and observe a substantial number of key 1

technical experts, and to assess the care and attention to detail which has characterized both the Applicants' analysis and the .

Staf f's independent review and analysis.

l Initiation of an HCDA would require multiple f ailures in the four types of preventive f eatures, even though such failures would be highly unlikely in light of the redundancy, diversity, and independence provided in those f eatures and l mandated by the CRBRP design criteria. The Applicants presented i

extensive analyses of the pathways for progression to HCDA conditions and the manner in which those features will limit, terminate, and mitigate progression along those pathways well short of HCDA conditions. The Staf f's independent analyses and review, which were extended beyond the Applicants' analyses to include review of LWR experience, domestic and foreign liquid metal fast reactor experience, and the available f ailure mode and ef fects analyses and initiator studies, verified that the Applicants' selection of events was suf ficiently comprehensive to envelop all credible HCDA initiators and sequences, and that no

., l 1

initiators and sequences of importance had been overlooked.

These analyses lend substantial insight into the tolerant physical response of the plant to a bounding set of postulated of f-normal conditions, and considerable confidence to the judgment that progression to HCDA conditions will be prevented.

Findings 9-15.

The evidence ccncerning ths foregoing analyses and reviews focused the inquiry on the four key f eatures which are necessary to prevent progression to HCDA conditions, namely, a) the reactor shutdown systems (RSS), b) the shutdown heat removal system (SHRS), c) features to prevent large primsry heat trans-port system pipe ruptures, and d) f eatures to prevent local imbalances in heat generation and heat removal from progressing to core-wide involvement. Findings 16-21. Each of these features embodies technologically proven design concepts and sound physical principles of operation or function, and the design of each is based upon tested criteria, appropriately modified to account for unique technological considerations, thereby assuring that the reliability of their respective j functions is inherently high. Id.

The RSS consist of two f ast-acting shutdown systems j (rather than one as in LWR's ), each of which is -independently capable of shutting down the reactor. The SHRS provides four heat removal paths, each independently capable of removing all decay heat. The RSS and three heat removal paths of the SHRS

-~ . .. . . - . - - . . - . .- - - -_ _ -. -

f unction automatically, and operator response consists merely of confirming and monitoring those functions. Findings 17-18.

The features for prevention of large pipe rupture rely .

upon four successive levels of protection, which are supported by extensive analytical and experimental evidence, and by worldwide operating experience. The four levels are: 1) stringent quality i etandards limit the potential for crack initiation from material l flaws; 2) even if flaws exist, piping toughness has been shown to limit growth of cracks to sizes well below that necessary to )

l penetrate a pipe; 3) even if a crack should grow and penetrate the pipe, it would be detected by sensitive leak detection well before any rupture could occur; and 4) even if a crack should

grow undetected, the crack would have dimensions well below those at which a pipe rupture would occur. In addition, the CRBRP can

- accommodate pipe leaks substantially larger than a design basis leak without a significant reduction in heat removal. Finding 19.

The features to prevent progression of local imbalances l

between heat generation and heat removal to core-wide involvement-incorporate passive mechanical interlocks to assure proper fuel subassembly positioning, and a multiplicity of redundant inlet flow paths to assure that flow cannot be blocked to any subassembly. Steel hexagonal subassembly ducts house each_ fuel rod bundle to inherently limit propagation of local imbalances between subassemblies, and extensive analyses, experimental data,

- ~ . . , . . .. ._ . . . _ _ _ . . _ . _ , - _ . _ _ _ . - _ . _ . _ . _ . _ . . . _ . _ . _ _ _ . _ . _ . _ . . _

=.

l .

4 4

f and domestic and worldwide operating experience all show that propagation beyond a single subassembly is highly unlikely. Any localized fuel failures can be detected by independent systems at levels well below those that could result in a significant local imbalance, and pending completion of testing at EBR-II, the Staff has imposed operating restrictions precluding any real possibility of local irabalsnce shich could pr' ogress to an HODA.

l j Finding 20.

Bcth Staff and Applicants have based their positions en a deterministic engineering approach, which is the most tested, mature, and comprehensive methodology available. Neither have used probabilistic assessments or reliability analyses as a decisive basis, and both have eschewed the use of numerical thresholds in arriving at their conclusions. The Board believes that this approach is in accord with the realities of the state of technology, and the. Commission's regulatory practice and policy. Findings 8 and 22-23.

4 Probability-based methodologies are not sufficiently mature to be employed in an absolute sense at this time. The appropriate role of probabilistic risk assessment (PRA) and reliability analyses should be to provide enhanced assurance that the inherent reliability of the CRBRP will be realized, and to i 21/ See 48 Fed. Reg. 10772 (March 24,1983) ; 48 Ped. Reg. 16014 (April 13, .1983) .

provide an assessment of the relative importance of systems and components to risk and safe operation. Id.

Consistent with this role, the Applicants are committed to complete a PRA by December 1984, and to implement a formal reliability assurance program (RAP), which will 2ddress all of the important plant safety features, with emphasis on the four features that are necessary to prevent CCDA's. The RAP will be conducted over the entire plant lifetime, and its implementation and results will be fed back into CRBRP design and operation.as x

part of a continuing risk management program. Findings 22-23.

The record shows that the Applicants'and Staff have properly applied proven engineering techniques and principles, with appropriate regard for quantitative probabilistic and reliability analyses, to show that it is highly unlikely that an HCDA will occur in CRBRP. The Applicants' and Staff's positions are well supported by reliable, probative evidence, and in light of that evidence, there is reasonable assurance that HCDA's need not be included within the CRBRP design basis.

B. THE ADEQUACY OF APPLICANTS' AND STAFF'S HCDA ANALYSES Although CRBRP has been designed so that HCDA's are properly excluded from the design basis, the NRC Staff has required the CRBRP to provide additional features and capabilities'in the design to maintain containment integrity and

,2_2/ Cf. 10 C.F.R. 3 50.34 (f) (1) .

,1 control releases of radioactivity, thereby limiting the residual risk of HCDA's.to acceptable levels. Finding 7. In the short ,

1 term (minutes or less), the containment could be challenged if an HCDA occurred and produced sufficient mechanical work energy to breach the reactor coolant boundary, and either overpressurize the containment due to sodium leakage and burning or generate minulles having sufficient energy to penetrate the containment.

Thus, a primary focus of the HCDA analysis has been the question of whether accident energetics are likely to exceed the structural capability of the reactor coolant boundary and thereby pose a short-term challenge to containment integrity. In the longer term (hours to months), an HCDA could result in thermal penetration of the reactor vessel (core melt-through), and challenge containment by overpressurization due to sodium and hydrogen burning and decay heat, or by non-condensible gas buildup. In this regard, the HCDA analysis has focused upon the capability of the CRBRP design features to accommodate longer-term challenges from containment overpressurization, and to control radiological releases to the environment. Findings 25-27, 30.

l l The Applicants' analyses and the Staff's independent l

l review and analyses considered the appropriate range of HCDA l initiators, with particular emphasis on the unprotected (i.e.,

failure to scram) loss-of-flow (LOF) and transient overpower (TOP) accidents, applied analytical tools which properly

s ,

represent the physical phenomena governing HCDA energetics, and determined that the likely outcome of HCDA sequences would be non-energetic. The Staff further concluded that an energatically induced reactor vessel head failure is physically unreasonable and does not pose a significant risk in CRBRP. Findings 27-29.

In spite of the evidence that HCDA sequences would be non-energetic, CRBRP has been designed with substantial margin to accommodate energetic events. Specific dynamic load, leakage, and geometric requirements have been imposed upon the reactor coolant boundary so that the .CRBRP can Isaintain the integrity of that boundary under conditions representative of an energetic HCDA. The dynamic loadings imposed by the Applicants were derived from an assumed set of conditions that produce an ulti-mate work potential of the core fuel of about 660 megajoules (;MJ) if expanded to one atmosphere (about 100 MJ if the expansion occurred within the free volume of the reactor vessel). Id,.

The Staff independently determined that HCDA's having an ultimate work potential up to 1130 MJ would produce only minimal energetic release on the primary system boundary, taking proper account of the attenuating effects of the reactor vessel internal structures. An accident h'aving an ultimate work potential of 2550 MJ was found to be necessary to challenge the structural capability of the~ boundary, taking proper account of

.the attenuating effects of the reactor vessel internal structure. These levels of energetics, 1130 MJ and 2550 MJ, I

p

- - - - - - - - - - - - - - - - - - ~ - - - -

k 94 I

correspond to disassemblids from the two-phase regime driven by reactivity ramp rates of 100 dollars per second (S/sec. ) and 200

$/sec., respectively, which are greater than any reactivity insertion rate estimated for any HCDA in CRBRP. Finding 28.

The record therefore shows that HCDA energetics, and consequent short-term challenges to containment integrity, do not pose a significant risk in CRBRP. The focus of the inquiry must then shift to consider the capability of the design to accommo-date longer-term challenges, and to control radiological conse quences .

In this regard, the CRBRP design features include a containment annulus cooling system, a reactor cavity vent system, and a containment vent / purge and cleanup system for accommodating longer term containment challenges and controlling radiological releases. Both Applicants and Staf f gave extensive consideration to, and presented detailed testimony addressing, the response of CRBRP to core melt-through phenomena, and the capability of these design f eatures. Findings 30-31.

The Applicants' analyses and the Staf f's independent review and analyses considered the phenomena of importance to core melt-through behavior (including sodium-concrete reactions, aerosol behavior, hydrogen generation, and containment structural response), took proper account of variations and uncertainties in these parameters, and yielded three conclusions of primary importance: a) core melt-through phenomena have been adequately

- , ,, , ..,,e- - - , - a e. , - - - , - - , - v ,v

I analyzed, b) long times (about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) would elapse af ter HCDA initiation before containment integrity would be challenged, and c) in the event of such challenge, the containment vent / purge and cleanup system will limit doses to values within the 10 C.F.R.

Part 100 dose guideline values. Id. The long times for the i

occurrence of containment challenges suggest that there is a significant margin of time ~ available for operator interdiction (repair or recovery of equipment or systems) or emergency action to further mitigate the risk of HCDA's.

Based upon the extensive evidence in the record, the Board finds that adequate attention has been directed to HCDA's, that HCDA energetics present an insignificant risk in CRBRP, and that the radiological risk of HCDA's in CRBRP is acceptably 1.ow.

C. BOARD AREAS OF INTEREST During the course of these proceedings, the Board noted a number of areas of interest and requested the parties to provide responses to each area. Based on these responses and the Board's inquiry during the evidentiary hearings, the Board is satisfied that each area has been properly addressed by the -

' Applicants and Staf f.

l L

23/ See 10 C.F .R. I 100.ll(a), footnote 2; see PID at 21, 74-80.

1 i

/

Board Area Of Interest 1 The Board requested the Staf f to advise the Board l

l whether information on new radiological source terms resulting from the Commission's Safety Goal Development Program would be evaluated for any impact in this proceeding. Bas ed on 'the evidence in the record, it is highly unlikely that any findings i by the Accident Source Term Program Of fice would affect the ,

source terms used in this proceeding, or affect the Board's conclusion that the Clinch River site is suitable from the standpoint of radiological health and safety or the assessment of i

environmental impacts of accidents. Findings 32-35. In the unlikely event that any changes in the SSST or environmental source terms are required, it is expected that such changes could i

be accommodated by the CRBRP design. In addition, the Staf f has committed to ensuring that the conclusions reached by the Accident Source Term Program Office receive appropriate j consideration during the OL stage of review. Finding 35.

l Board Area Of Interest 2 f

Applicants were requested to discuss the anticipated performance of CRBRP fuel during normal operation and in the event of f uel f ailure. The Applicants have committed to impose

! design requirements on the CRBRP fuel which will ensure that even in the case of extremely unlikely transients, the fuel.will be maintained in a coolable configuration. Findings 36 -37 . The analyses of operating and transient conditions and cladding

~

properties, which are supported by an extensive experimental data base, confirm that the overall design requirements will be met by the CRBRP fuel. Finding 38.

Board Area Of Interest 3 Applicants were requested to present a technical summary of how the coolant characteristics of sodium will result in a reduced likelihood of pipe rupture in CRBRP primary heat transport system piping. In terms of piping integrity, the high boiling temperature of sodium allows operation of CRBRP at near atmospheric pressure; thus reducing the primary membrane stress on the piping walls to one-third of the allowable stress under normal operating conditions and approximately one-sixth of the allowable stress for accident conditions. Finding 39. Because of the low primary stress, the CRBRP piping has a critical crack length which is very large compared to the length at which a developing crack would be detected by the leak detection systems, thus permitting timely and ef fective corrective action. Finding 41.

Board Area Of Interest 4 Applicants were requested to explain the capability of CRBRP to remove decay heat by natural convective circulation of the sodium coolant. . Bas e'd on Applicants' and Staff's responses, the Board believes there is reasonable assurance that natural convective circulation of sodium will be available to transfer

decay heat. Findings 42-43. The convective circulation capa-bility results from placement of the thermal centers of the heat exchanging components at successively increasing elevations in the plant, thus providing the necessary thermal driving head for adequate loop flows. Convective circulation will occur without any operator action and will remove decay heat in the event of loss of all of f-site power and the loss of all on-site diesel generators. Finding 42. Convective circulation has been verified by both Applicants' and Staf f's analyses and by evaluations of data from tests conducted at EBR-II and FFTF. In addition, convective circulation will be demonstrated in CRBRP during the initial start-up testing program. Finding 43.

Board Areas Of Interest 5 and 6 Applicants were requested to explain their quality assurance (QA) program, including how divisions of authority have been assigned between contractors and the Applicants, how differing functional levels of QA will be applied, how current as-built plans and specifications will be maintained, how QA and quality control (OC) activities will be integrated, and how the Staf f will monitor those activities. Applicants have established a comprehensive QA-QC program using a disciplined management system of checks and balances. The CRBP@ Project Office, which has central management and control over project activities, has clearly assigned responsibilities to each of the major contrac-tors, including the responsibility' to plan, implement, and manage j

I

. )

l 1

1 integrated quality assurance programs over the particular contractcr's contractual scope of work. Finding 44.

In order to ensure that the program QA functions are coordinated and integrated, the project has established three levels of control: 1) the system, component, material and service supplier programs, which are primarily quality control programs concerned with direct control and verification through analys is, review, inspection, examination and testing; 2) program participant programs (those having direct or indirect interfaces with each other and the Project Of fice), which are management-type programs with responsibilities for quality assurance f unctions such as surveillance, audit, interface coordination, and lower-tier program integration f unctions, including overview of the lower-tier quality control processes; and 3) the Project Office program, which is a management-type program with audit and surveillance activities for verification of participant performance, interf ace coordination and program integration functions, including the coordination of fabrication and construction ef forts for the project. Finding 45.

The coordination of interf acing systems is controlled through a formal review and approval cycle that provides the necessary safeguards for proper system integration. Interface control is also a f undamental part of the design control process. Through a four-tier design control system, which includes design reviews and configuration management, the

l interf ace requirements between systems are defined, translated into project specifications, and enforced as project require-ments. Finding 46.

The quality assurance program is applied, in a gr'aded manner, to all systems, structures and components in CRBRP regardless of their safety classification. In implementing this graded approach to quality assurance, nine levels of program reqpirements have been developed which are applied based on the importance of items or services to the plant's f unction. To guide the selection of the appropriate level of quality assur-ance, a matrix has been developed which takes into account such factors as service f unction, plant application, saf ety category or classi fication, temperature application and production category, and provides recommended levels of quality assurance to apply to any particular plant item. For each plant item, the recommended quality assurance programs and activities derived from the matrix are specifically reviewed and examined by the cogni zant technical and quality assurance disciplines to assure that the appropriate level of QA ef fort is applied. Findings 47-48.

The project will, through its configuration management system, ensure that accurate as-built plans and specifications will be available when needed. Configuration management formally approves and establishes the reference plant design on a system l

basis. Once formally approved, i.e., " bas el ined" , any changes to l

that design require review and approval before the change is implemented. The baselined documentation will be required to reflect the as-built configuration of that structure, system or component. This documentation is maintained and stored in a quality records system, which provides the necessary documentation retrieval system for using the baselined documentation. Finding 49.

Both Applicants and Staf f have treated quality control as an integral part of quality assurance, and the Staff has and will closely monitor quality assurance program implementation and results for CRBRP, with particular attention to the special attributes of CRBRP. Findings 50-51.

Board Area Of Interest 7 Applicants were requested to discuss control of river traf fic during of f-normal plant conditions and also to discuss the potential for hazardous cargo on the river to pose a threat to CRB RP . Through cooperative arrangements with the Tennessee Wildlif e Resources Agency, the Coast Guard, and the Army Corps of Engineers, Applicants will implement ef fective control measures over commercial and recreational traf fic in the Clinch River within both the 10 mile plume exposure pathway emergency planning zone and the portion of the Clinch River immediately adjacent to the exclusion boundary. As to shipnent of hazardous material, it is highly unlikely that such shipments would take place. In the past, no hazardous materials have been transported by barge past

l .

t the site. In the f uture, coal and steel are the only materials likely to be shipped by barge past the site. In the unlikely event that a new industry would ship hazardous materials past the site, the Corps of Engineers permit procedure would require identification of hazardous material shipments, thus allowing precautions, if needed, to be implemented at that time. Findings 52-53.

Board Area Of Interest 8 The Board requested Applicants to discuss possible challenges to the steam generators and the containment /

confinement structure arising from transient-induced overpressure and overtemperature conditions.

The structural design of the steam generator modules accounts for both steady state and transient loads. The most severe thermal transient imposed on the steam generator is that predicted to result from postulated water side isolation and dump of an evaporator with concurrent failure of the water inlet isolation valve. Because of its " hockey stick" configuration, the steam generator module will accommodate the differential

! expansion between the tube and shell resulting f rom this transient without interference or excessive stresses. The ability of the eteam generators to withstand thermal transients will be confirmed by testing prior to fab-ication of the steam i

generator modules for the plant. Finding 54.

\ -

I 1 .

Since neither the intermediate heat transport system

,_ (IRTS) sodium nor the steam generator water / steam are significantly radioactive, the primary nuclear saf ety consideration related to the steam generator modules is mitigation of the ef fects on the intermediate heat exchanger (IHX) from a sodium-water reaction. The steam generator modules have been designed with special features to both prevent tube ,

leaks and to mitigate the eff ects of any leaks which might occur. In the area of prevention, special attention has been given to the design, choice of materials, inspection procedures and construction techniques to ensure the integrity of the steam generator tubes. Three levels of protection are provided to mitigate steam generator tube leaks and to ensure the integrity of the IHX: 1) a sensitive leak detection system capable of

~

detecting leaks as small as 2x10 lb. water / second; 2) a rupture disc on the cover gas space of the IHTS expansion tank to relieve the slow pressure rise associated with a postulated intermediate-5 size tube leak, and to cause automatic plant shutdown and water I

i side isolation of the steam generator modules in the af fected loop; and 3) an engineered safety f eature, the Sodium Water Reaction Pressure Relief System ( SWRP RS ) , which will limit loadings in* the IHX to an acceptable level in the event of a 1

design basis leak. Findings 55-59.

The bounding DBA with respect to containment vessel temperature and pressure retaining capability is the postulated I

_ = - - - . - _ , -.- .. .-

i failure of the primary sodium storage tank (PSST) during main-tenance. This DBA assumes a series of highly unlikely events, including the postulated instantaneous f ailure of the PSST while it contains 35,000 gallons of sodium - even though the PSST will contain this volume of sodium only a few times during plant life. The PSST f ailure is accommodated by 1) actuation of the containment isolation system and 2) passive accommodation of the pressure and temperature resulting from the sodium burning. The maximum containment pressure and temperature predicted for the PSST f ailure DBA are well within the containment vessel design pressure and temperature. Findings 60-63. The effectiveness of the containment system to deal with the radiological conseq;uences of design basis accidents was analyzed using the CRBRP site suitability source term (SSST). The SSST was postulated for site suitability evaluation and is identical to that used for light water reactors (LWR's), except that the CRBRP source term contains 1 percent plutonium (whereas none is included in the LWR source term). All SSST doses were well below the site suitability dose guideline values. The maximum dose resulting from the release to containment of any design basis accident is many times less than the corresponding SSST dose. Findings 63-64.

The design features which mitigate the consequences of beyond design basis accidents (by maintaining containment integrity and controlling radiological releases) include the

~. j containment annulus cooling, vent / purge, and containment cleanup systems, and instrumentation to follow the course of an accident. These features will be ef fective in reducing the residual risks f rom HCDA's in CRBRP to acceptable levels. See Section II B, supra. There is considerable margin in the pressure capability of the CRBRP containment to accommodate HCDA's. The highest pressure calculated for any of the sensitivity studies was approximately 30 psig, as compared with a containment pressure capability of about 40 psig. Findings 65-71.

Board Area of Interest 9 The Board asked the Staf f to discuss the need for revision of Protective Action Guidelines (PAG's ) for the CRBRP.

Additional PAG's, beyond those already established by the Environmental Protection Agency (EPA) for nuclear incident response planning, are not needed for the CRBB@. The EPA has established the range of PAG's for the plume exposure pathway emergency planning zone (EPZ) as 1 to 5 rem for whole body and 5 to 25 rem for thyroid exposure. The HCDA radiological dose consecaences for whole body and/or thyroid are more limiting than those for other organs. While PAG's for other organs can be derived by scaling from the existing whole body r thyroid PAG's, because the whole body and thyroid doses are controlling, it is unlikely that any specific PAG's for other organs would be either necessary or usef ul 'in CR3RP em?rgency planning. Findings 72-73.

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37 - .

Board Area of Interest 10 The Board asked the Staff to explain the significance of DOE's goals for material control and accountability at the Developmental Reprocessing Plant (DRP) to the effectiveness of CRBRP fuel safeguards measures. Research and development activities on measurement capabilities for material control and accounting are not necessary for the effectiveness of safeguards at the DRP. Primary reliance for protection against theft of nuclear material at the DRP is placed on physical protection systems. The primary role of material control and accounting is to provide assurance that the protective systems are working eft;ctively. Physical security and material control and accounting do not have to be considered independently. While rapid material accounting may augment safeguards measures to prevent unauthorized diversion of fuel at the DRP, the DOE commitments for DRP safeguards provide a level of protection equivalent to that embodied in current NRC regulations without the need for a rapid material accounting system. Finding 74.

Board Area of Interest 11

1 The Board asked the Staff to explain the significance of its statement concerning "isentropic expansion yield to one

atmosphere" in the context of the HCDA energetics analysis , and whether such values contribute to the conservatism in the analysis. The concept of an "isentropic expansion yield to one atmosphere" is used as a convenient reference point to indicate

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38 - .

the relative potential severity resulting from disrupted core conditions. It has no physical application in that such yields  :

cannot be realized in real systems. It is not directly used in j analyzing the capability of the system to accommodate HCDA loads, and does not make a real contribution to the conservatism in the analys es. Finding 75-76.

Board Area of Interest 12 The Board asked the Staf f to comment on the potential impact of items identified in the SER for resolution at the OL review stage. The items identified in the SER as requiring review at the OL stage, and which have the potential for resulting in substantive changes of a costly or time-consuming nature, f all into the following areas a) fuel design limits, methodologies, and bases; b) high temperature mechanical design limits and methodology; c) reactor vessel head structural capability; f d) PRA/ reliability analysis; and

(

e) natural circulation.

It is unlikely that any of these will result in a significant impact on cost or schedule. Staff and Applicants have agreed on l a course of completion for each item and are developing a timetable for their review and final resolution. In any case, yve-- y ,e ,-e-- o we

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l-l the process of confirmation or resolution of these items will not 24 l

result in any compromise of safety.- / Findings 77-78.

Board Area of Interest 13

)

l The Board asked the Staf f to discuss the potential '

impact of the fuel system f allback positions on CRBRP programmatic objectives. The f uel system fallback positions consist of restrictions on CRBRP operation which can be imposed -

on the f uel system design basis, limits and methodology if future analytical and experimental data do not substantiate the Applicants' proposed design. It is unlikely that any of the f allback positions would have to be implemented. Applicants have committed to address the Staf f's concerns through experimental and analytical programs, including programs at EBR-II and FFTF to confirm fuel design performance predictions. Even if a f allback position were implemented, redesign of reload fuel can be l

! 24/

~~~ It should be noted that Appendix B of the SER (S Exh 27, App. B) comprehensively addressed all unresolved generic

saf ety issues for the CRBRP. Gulf States Utilities Co.

(River Bend Station, Units 1 and 2), ALAB-444, 6 NRC 760 (19 77 ) .- Appendix B shows that in all cases the problem has either 1) already been resolved for CRBRP; or 2) there is a

reasonable basis for concluding that a satisf actory solution will be obtained before the reactor is put in operation.

, Id. See also Virginia Electric and Power Co. (North Anna Fuelear Power Station, Units 1 and 2), ALAB-491, 8 NRC 245, 248 (1978). Thus, there is reasonable assurance both a) that there will be a satisf actory resolution of the out-standing saf ety questions prior to operation ' of the f acil-ity, and b) that operation will not present undue risk to the public health and safety. 6 NRC at 777-778; Finding 132.

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accomplished and it is unlikely that CRBRP programmatic objectives would be compromised. Findings 79-81.

Board Area of Interest 14 The Board asked the Staf f to comment on the ef fect of operation with leaking fuel pins on CRBRP fuel performance. Bo th the Staf f and Applicants have considered the ef fects of leaking f uel pins on of f-normal performance of the f uel. S odium in contact with f uel inside a pin may result in increased pellet-to-clad gap conductance but would not adversely affect fuel performance. The Staf f is concerned that continued operation with f ailed fuel rods might cause local swelling with the potential for flow restriction. Staf f and Applicants have agreed to operational restrictions on CRBRP which would require removal of any f uel assembly containing f ailed pins at the next reactor shutdown or upon exceeding a predetermined delayed neutron signal .

Those restrictions will be reviewed based on the results i of tests being conducted at EBR-II to investigate the limits of steady-state and transient operation with f ailed f uel rods. '

Findings 82-83.

i Board Area of Interest 15 The Board asked the Applicants -to discuss the I

relationship of the reliability assurance program (RAP) with the i l

quality assurance (QA) program. 'The RAP is not a formal part of the QA program. However, certain QA program activities are

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I applied to f ulfill requirements of the RAP. For instance, the RAP f ailure evaluation activities are considered part of the QA program. As part of the quality assurance program for the CRBRP as a whole, surveillance and auditing of RAP activities and documentation will be performed to assure that RAP procedures are being met and that overall RAP requirements 'are satisfied.

-Finding 84.

Board Area of Interest 16 The Board asked the Staf f to comment on the ef fect of variations in concrete aggregate composition on aerosol behavior and its relation to containment shell cooling. The results of sodium-concrete testing with both calcitic and dolomitic limestone aggregate demonstrated no detectable difference in sodium-concrete reactions and depth of penetration of concrete.

The type of aggregate would not af fect the sodium aerosol behavior in the reactor containment building (RCB). No significant difference in the physical characteristics of the reaction products was observed in tests of the two types of concrete. Virtually all of the sodium aerosols in the RCB would be generated in the RCB by burning of the sodium vapor being vented from the reactor cavity, and would be unaf fected by the reaction occurring in the reactor cavity. Finding 85.

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Board Area of Interest 17 The Board asked the Staff to comment on eight areas of concern which were identified at the initial stages of the Staff's review of HCDA energetics. All of those concerns have j 2

been resolved to the Staff's and the Board's satisfaction. l Findings 86-94.

D. OTHER MATTERS The following additional matters of interest were included within the Board's inquiry: 1) matters arising from a limited appearance statement filed by NRDC eti al.; 2) matters relating to emergency planning at nearby DOE industrial facilities; 3) the matter of whether the results of the CRBRP PRA can be incorporated in a timely and effective manner; and 4) a series of additional matters relating to the Board's decision here. We now address each of these areas of inquiry.

1. NRDC's Statement The substance of NRDC's statement consisted of an argument previously presented at closing argument during the LWA proceedings. The Board has found no basis to credit the argument in the LWA record, and'none of the new matters raised by the statement were found to have merit. Findings95-105.

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2. Emergency Planning at Nearby DOE Facilities 4

As noted in Finding 52 of the PID, the Board considered the matter of emergency planning at nearby DOE industrial facili-ties (the Oak Ridge National Laboratory, the Oak Ridge Gaseous Diffusion Plant, and the Y-12 Plant). All of these facilities a have existing-emergency programs in effect, and there are no significant impediments to evacuation or protection of workers at these facilities in the event of an accident at CRBRP. Findings 107-111.

3. Incorporation of PRA Results The PRA for CRBRP will be completed in December 1984.

Based upon the existing engineering analysis of CRBRP, the reliability inherent in the major safety-related systems and components, and the projected schedule for construction and operation, there is reasonable assurance that the results of the PRA will be implemented in a timely and effective manner.

! Findings 112-115.

4. Additional Matters During the Conferences with Counsel held on May 13 and June 29, 1983, the Board requested that the Staff's conclusions relative to the seismic qualification of small diameter piping be furnished to the Board and offered into evidence in this proceeding. TR 7270-71, 7345. Applicants evaluated the seismic margins available in both large and small diameter Heat Transport

~

System (HTS) piping and conservatively calculated the margins to be 1.45 for both. S Exh 35 at 2. The Staff's independent evaluation determined that the margin in actual cases could be much higher and that the seismic margins for both small and large piping are generally comparable and significant; this can be translated into a " reserve margin earthquake" of 0.605g. (in comparison the CRBRP safe shutdown earthquake (SSE) is 0.25g.).

Id. The Staff concluded that a) the approach used by Applicants to assess the HTS piping seismic margin is acceptable and b) the seismic margin in small HTS piping is significant and comparable to the margin in large HTS piping. Id.

III. CONCLUSION Based on the foregoing, and pursuant to 10 C.F.R.

3 50.35 and 10 C.F.R. S 2.104 (b) (2) , the Board finds that the record of the proceeding contains sufficient information and the review of the application by the NRC Staff has been adequate to conclude that the Applicants have described the proposed design of the facility, including, but not limited to, the principal architectural and engineering criteria for the design, and have identified the major features or components incorporated therein for the protection of the health and safety of the public. In addition, such further technical or design information as may be required to complete the safety analysis, and which can reasonably be left for later consideration, will be supplied in the final safety analysis report. Moreover, the safety features-

or components, if any, which reqpire research and development i

have been described by the Applicants and the Applicants have identified, and there will be conducted, a research and development program reasonably designed to resolve any safety questions associated with such features or components. On the basis of the foregoing, there is reasonable assurance that such saf ety questions will be satisf actorily resolved at or before the latest date stated in the application for completion of con-struction of the proposed f acility, and taking into consideration the site criteria contained in 10 C.F.R. Part 100, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. The Board f urther concludes that the Applicants are technically qualified to design and construct the CRBRP, and that issuance of a permit for construction of the f acility will not be inimical to the common defense and security or to the public health and safety.

The matters examined during the evidentiary hearing which are not discussed in this opinion were considered by the Board and found either to be without merit er not to affect our decision herein. The findings of. fact and conclusions of law which are annexed hereto are incorporated in the opinion by reference as if set forth at length. In preparing its findings of f act and conclusions of law, the Board reviewed and considered the entire record and the findings of f act proposed by the Appli-l l

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cants and NRC-Staff. Further, the Board has reviewed its pre- i vious Partial Initial Decision in light of the entire record, and except as otherwise indicated, hereby affirms and incorporates that Partial Initial Decision and its annexed findings and con-l clusions, including its determinations that the requirements of the National Environmental Policy Act and 10 C.F.R. Part 51 have been. complied with, that the final balance among conflicting factors contained in the record of the proceeding has been inde-pendently considered, and that, subject to the conditions set forth in the Partial Initial Decision and this Initial Decision, the Construction Permit should be issued. Those proposed find-ings not incorporated directly or inferentially in this Initial Decision are rejected as being unsupported by the record of the case or as being unnecessary to the rendering of this decision.

The Board, having made the findings and determinations required by 10 C.F.R. 5 50.35 and S 2.104 (b) (2) , concludes that the Director of Nuclear Reactor Regulation, upon making requisite findings with respect to matters not embraced in this Initial Decision, may issue the Applicants a Construction Permit for the l

Clinch River Breeder Reactor Plant consistent with the terms of the Partial Initial Decision and this Initial Decision.

IV. ORDER Wherefore, it is ordered that the Director of Nuclear

[

Reactor Regulation is authorized, in accordance with the Commis-sion's regulations, to issue Applicants a permit to construct the l

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Clinch River Breeder Reactor Plant, consistent with the terms of l

the Partial Initial Decision and this Initial Decision. -

It is further ordered that this Initial Decision shall constitute the final action of the Commission forty-five (45) days after the issuance thereof, subject to any review pursuant to 10 C.F .R. { { 2.76 0, 2.762, 2.764, 2.785, and 2.786.

Exceptions to this Initial Decision may be filed within ten (10) days af ter its service. A brief in support of the exceptions shall be filed within thirty (30) days thereaf ter and forty (40) days in the case of the Staf f. Within thirty (30) days of the filing and service of the brief of the Appellant, and forty (40) days in the case of the Staff, the other party may file a brief in support of, or in opposition to, the exceptions.

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,l IT IS SO ORDERED.

FOR 'IHE ATOMIC SAFETY AND LICENSING BCARD Marshall E . Miller, Chairman l ADMINISTRATIVE JUDGE i

Dated at Bethesda, Maryland This day of , 1983 Gustave A. Linenburger, Jr.

ADMINISTRATIVE JUDGE Dr. Cadet H. Hand, J r.

ADMINISTRATIVE JUDGE WHEREFORE, on the basis of the foregoing and the annexed Proposed Findings of Fact and Conclusions of Law, j Applicants respectfully request that the Board adopt this i

Proposed Decision.

Respectf ully submitted, At y fo Project Management Corporation

/ i H d ~K:o . (s I ' !' t .

William D. Luck Attorney for United States Department of Energy l

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DATED: August 15, 19 83 OF COUNSEL:

l Thomas A. Schmutz Herbert S. Sanger, Jr.

Frank K. Peterson Lewis E. Wallace

)

Morgan, Lewis & Eockius W. Walter LaRoche 1800 M S treet, N .W. James F. Burger Washington, D. C. 20036 Edward J. Vigluicci Tennessee Valley Authority Leon Silverstrom 400 West Summit Hill Drive U.S. Department of Energy Knoxville, Tennessee 37902 Of fice of General Counsel 1000 Independence Avenue, S .W.

Room 6B-256 -- Forrestal Bldg.

Washington, D. C. 20585 J

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V. FINDINGS OF FACT

1. The Applicants in this Construction Permit (CP) proceeding are the United States Department of Energy (DOE), Project Management Corporation (PMC), and the Tennessee Valley Authority (TVA). S Exh 26 at 1-1, A Exh 86.

2. The other party to the proceeding is the Staff of the United States Nuclear Regulatory Commission (Staf f or 'NRC Staff ) .

The Natural Resources Defense Council, Inc. (NRDC) and the Sierra Club (Intervenors) participated actively during the Limited Work Authorization (LWA) stage of the proceedings, but were dismissed as parties after all of their contentions-were dismissed or withdrawn. TR 7333.- The Tennessee Attorney General and the City of Oak Ridge have status as an interested state and municipality, res pecti ve3.y, undec 10 C.F.R. $ 2.715(c), but neither participated in the CP proceedings.

3. Applicants seek a CP and Operating License (CL), pursuant to

.Section lO4(b) of the Atomic Energy Act, for the Clinch River Breeder Reactor Plant (CRBRP ) . S Exh 23 at 1-2 7. S Exh 22 at 1-1; A Exh 86. These findings concern issuance of a CP. '

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4. The CRBRP is a Liquid Metal Fast Breeder Reactor (LHFBR) l l

demonstration plant with a rated output of 350 megawatts (net) electrical which is proposed for location on the Clinch River in Oak Ridge, Tennessee. S Exh 26 at 1-4, 1-12.

5. This is an uncontested proceeding and there are no disputed issues before the Board related to the issuance of a CP for the CRBRP. The Board, however, has raised seventeen -

specific areas of interest in the Notice of Construction Permit Evidentiary Hearings. 48 Fed. Reg. 23944 (May 27, 1983); see Appendix A hereto. In addition, at~its June 29, 1983 Conference with Counsel, the Board affirmed that the

-parties should also address 1) whether a hypothetical core disruptive accident (HCDA) should be a design basis accident (DBA) for the CRBRP and 2) whether the analyses of HCDA's for the CRBRP were adequate. The Findings which follow address: a) the issue of whether an HCDA should be a DBA,

^

b) the adequacy of the HDCA analyses; c) the areas of Board interest; and d) other matters related to a decision as to

, whether a CP should be authorized.

L THE CRBRP DESIGN SAFETY APPROACH

6. The proposed design approach to consideration of accidents in CRBRP is similar in most respects-to that normally

. applied in the Commission's reactor licensing process.

Three levels of safety, incorporating the traditional l defenselin-depth concept, have been defined. S Exh 26 at

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l-2; see 10 C.F.R. Part 50, Appendix A. The Staff has established the objective that CRSBP will achieve a level of '

safety comparable to that of Light Water 3eactors (LWR's).

S Exh 32 at 14. Major emphasis has been placed upon the prevention of accidents which could l$ad to core melt and disruption and loss of containment integrity. To this end, features tnd characteristics are incorporated in the design to assure that the likelihood of core disruptive accident initiatiobisverylow, and that the identified accident sequences are terminated within acceptable (design basis)

limits -- that is, before they progress to core melt and l

disruptive conditions. S Exh 32, TR 8036-8101; A Exh 87, TR 7378-7'594. On this basis, the Applicants and Staf f have concluded that core melt and disruptive accidents can be i

excluded from the CRBRP design basis. A Exh 87, TR 7378-7594;;5{., Exh 3 2, TR 8036-8101.

7. Notwithstanding the foregoing, the CRBBP design approach is I

unique insofar $s the NRC Staff has required the CRBRP to

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provide additional features and capabilities in the design to assure that there is a low likelihood of containment failure and other unacceptable con' sequences associated with core melt and disruptive accidents beyond the design.

1/

bas ie ;i- A Exh 89, TR 7763-7916; S Exh 41, TR 8270-8442.

1/

- The Staff determined that these requirements should be imposed based upon their judgment concerning more limited experience with CRBRP relative to Light Water Reactors (LWR's). S Exh 23 at I-4. s.

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The reactor coolant boundary has been strengthened and seals have been added to the reactor vessel closure head to accommodate loadings from core disruption and to limit the leakage of sodium, gases and vapors to the containment. A Exh 89 at 5-9, TR 7767-7771; A Exh 90, 5 5.4. In addition, the design includes a vent / purge and cleanup system to relieve containment pressure and to limit hydrogen and radionuclide concentrations, a containment annulus cooling system, pressure relief vents for the reactor cavity, and associated instrumentation. A Exh 89 at 11-24, TR 7773-7786; A Exh 91, $$ 2.1-2.2. These design provisions have been evaluated against core disruptive accident conditions, including all pertinent nuclear, thermal, structural and radiological considerations, to ensure that the residual risks of such accidents in CRBRP are acceptably low and made comparable to those in LWR's. S Exh 41, TR 8270-8442; 5 Exh 27, App. A; S Exh 30; A Exh 89, TR 7763-7916; A Exhs 90-

93. Taken as a whole, the proposed design safety approach thus reflects a full consideration of the range of potential accidents, with special emphasis on prevention of core disruptive accidents and assurance that the residual risk associated with such accidents is acceptably low. In this context, we r:oceed to consider the question of whether an HCDA should be a DBA, and the adequacy of Applicants' and Staf f's HCDA analyses.

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WHETHER AN HCDA SHOULD BE A DBA 8, Both Applicants and Staff have based their position on j whether an HCDA should be a DBA on deterministic judgments, criteria, analyses, and applicable experience. S Exh 32 at 7-8, TR 8042-8043; A Exh 87, TR 7378-7594. Although probabilistic assessments and reliability analyses will play a role in the ultimate design and operation of CRBRP (see Findings 22 - 23 below), the results of such assessments and -

analyses have not played a decisive role in either the Applicants' or Staff's position that HCDA's should not be DBA's. S Exh 32 at 13, TR 3048; A Exh 87 at 175-177, TR 7552-7554. Neither Applicants nor Staff have established or j used numerical probability thresholds as the bases for their positions. S Exh 32 at 7-8, TR 8042-8043; A W Clare, TR 7742. . Probability-based methodologies are not sufficiently mature to be used as a decisive basis, and the deterministic approach is the most tested, mature, and comprehensive methodology available for this purpose. S Exh 32 at 13, TR 8048; A W Clare, TR 7749. The Staff has established the objective that CRBRP will achieve a level of safety compara-

, ble to LWR's, and has employed the deterministic methodology

^

l developed for LWR's, modified to account for technological differences, as its basis for excluding HCDA's from the CRBRP design basis envelope. S Exh 32 at 14, TR 8049.

.9. Initiation of an HCDA requires multiple failures of mitigat-ing safety systems. S Exh 32 at 7, TR 8042. NRC regulatory

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practice has traditionally applied deterministic criteria, such as systems redundancy, diversity, and independence, to determine that such multiple failures are not credible, i.e., need not be postulated as part of the design base.

S Exh 32 at 7-8, TR 8042-8043. Based upon these principles, relevant experience with LWR's and the Fast Flux Test Facility, and the NRC Staff's Standard Format and Content of Safety Analysis Reports for Liquid Metal Fast Breeder -

Reactors, the Applicants developed an extensive initial list of design basis events for detailed analysis and review by l the Staff. S Exh 32 at 8, TR 8043; S Exh 26 at'15-1 -- 15-78; A Exhs 71-72, 33 15.0-15.7. Through extensive analyses and 4

review of the CRBRP safety systems, engineered safety features, and the prescribed DBA's as compared to the design, the Staff verified that the DBA spectrum for CRBRP is sufficiently comprehensive, and that no initiators and/or sequences of importance to HCDA initiation have been

, overlooked. S Exh 32 at 8-9, TR 8043-8044; S Exhs 26-28, l

passim; S Exh 26, SS 6 and 15. The Staff extended and confirmed this verification through a comparison of CRBRP DBA's to those in LWR's and domestic and foreign liquid metal fast reactors, and an examination of the available failure mode and effects analyses (FMEA's) and initiator studies for CRBRP. S Exh 32 at 9, 36-42, TR 8044, 8071-8077; S'Exhs 33 and 34.

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- F 10. The Staf f's review, which was based upon the traditional l l

daterministic criteria and methodology (with modifications 4

to reflect technological dif ferences), disclosed that the safety functions necessary for prevention of HCDA's are not fundamentally different for LWR's and LMPBR's. S Exh 32 at 14, TR 8049. The Staff concluded, however, that in light of i

the technological differences, a set of major additional requirements were necessary to assure that HCDA's are sufficiently unlikely that they need not be treated as DBA's in CRB RP . These are: 1) two f ast-acting, redundant, diverse, and independent reactor shutdown systems; 2) redundancy, diversity and independence in decay heat removal, such that after an initiating event and a single failure, at least two heat removal paths would be available;

3) means for prevention and/or timely detection of local l imbalances in heat generation and removal; 4) provision of suf ficient sodium flow and inventory maintenance for heat removal; and 5) provisions for accommodating sodium 2

leaks .- / S Exh 32 at 15-16, TR 8050-8051. These requirements are reflected in four major CRBRP design features or sets of measures which are of controlling importance to prevention of HCDA's:

1) the reactor shutdown 2/

The Staf f also specified that there should be: 1) provi-sions for accommodating sodium fires; and 2) a commitment to a formal reliability assurance program. S Exh 32 at 15-16, TR 8050-8051. See Finding 19, note 14 and Finding 22, respectively, infra.

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systems (RSS); 2) the shutdown heat removal system (SHRS);

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3) means for preventing large leaks in primary system 1

piping; and 4) means for preventing local imbalances in heat l 1

generation and removal from progressing to significant core,- i I

wide involvement. )

11. The Applicants undertook extensive mechanistic analyses to examine the potential for progression to HCDA conditions at a f undamental physical level. At this level, it was shown that all initiators and sequences of importance to initia-tion of HCDA conditions - irrespective of their individual causes or details - must involve one or both of two conditions in the reactor core: a) reduced heat removal, or b) excessive heat generation.-3 / A Eih 87 at 4-5, TR 7381-7382; S Exh 32 at 10, TR 8045; PID at 66; A Exh 1 at 14-15, TR 2003-4. The potential pathways by which an imbalance in heat generation and removal could occur are depicted as follows:

3/

The Staff's approach was substantively the same, and differed only as a matter of convention. The Staff defined four basic l conditions or failures of importance to HCDA initiation: a) I f ailure to shut down the nuclear chain reaction during an overpower or flow reduction transient (including whole core and local events ); b) f ailure to maintain suf ficient coolant j inventory; c) f ailure to maintain sufficient coolant flow; and  ;

d) ' f ailure to extract suf ficient heat from coolant. S Exh 32 at'lO-ll, TR 8045-8046. Failure a) involves excessive heat generation or reduced heat removal, while f ailures b), c), and d) involve reduced heat removal.

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Whole Core Pressesteen embaience d

4 I I g incrossed Reduced G tion Remo

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1 I I I I I g., p,., Core Control Rod g,,,,,, Loss increased

%,g Compaction Wnhdrewel of Flow iniet Temp, 6

+ l l l l l l l Rohseling Enr6chment Error ggeeg,,, ,,, Lose PHTS n ,e Y reorn Erro' of Pump resures Leah 73,pwys n 4 h $

I l l l l l l l Inlet in-Core Outset riesion Gas moce Molten ruel Loss of Mechaniced SOP tHTS or SGS macn.se mocasse R.i e R.i e. ro e e.isu,. resi .. F.aur.

A Exh 87 at 5, TR 7382.

The Applicants' testimony examined in extensive detail the capability of the design to terminate, limit, and mitigate each of these potential pathways, and showed that the CRBRP i

l design will prevent the progression of core conditions beyond the design basis to HCDA initiation. A Exh - 8 7 at 4-54, TR 7381-7 431.

12. The record contains voluminous detail concerning Applicants' analyses and the Staf f's independent review and analyses of design basis accident sequences for CRBRP. A Exh 87 at 4-54, TR 7381-7431; A Exhs 71 and 72, $ 15; S Exh 26, $ 15; S

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Exh 32 at 36-42, TR 8071-8077. These analyses lend considerable insight into the mechanisms and pathways necessary for progression to HCDA conditions, and a high degree of confidence that CRBRP has been thoroughly l

engineered to preclude HCDA initiating conditions. Id.

13. i The analyses commence with definition of the two fundamental ,

l core conditions of interest to HCDA initistion: 1) reduced i

, heat removal, and 2) increased heat generation. A Exh 87 at 4-5, TR 7381-7 382. The Applicants selected the criterion that no DBA will result in loss of coolable geometry in order to test the ability of the CRBRP design features to prevent core conditions from approaching HCDA initiation.

This would, in turn, preclude core meltdown or any reactivity insertion from core material relocation. By maintaining the sodium coolant temperature below its boiling point, cladding melting is precluded. Therefore, cladding support to the fuel material and coolant channels is ass ur ed, and loss of core coolability is prevented. On this basis, the Applicants have translated the no-loss-of--  !

I coolable geometry criterion into a surrogate criterion which I

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reqpires that no DBA will result in sodian boiling.4/ -

A Exh 87 at 6-8, TR 7383-7385.

14. Accidents involving reduced whole core heat removal result from reduced capacity or f ailures in the overall heat transport system (NTS) .-5 / A Exh 87 at 9, TR 7386. Such reduced capacity will be reflected as reduced primary coolant flow through the core or increased primary coolant temperature at the core inlet. A Exh 87 at 18-19, TR 7395- ~

7396. Bounding events involving reduced core flow (instantaneous loss of pumping power to all three primary HTS pumps, and instantaneous seizure of a primary pump) have been imposed on the design, and either the secondary or primary RSS and the SHRS will f unction to preclude sodium boiling in the core for these events. A Exh 87 at 19-29, TR 7396-7406. Leakage from PHTS piping would also result in reduced core flow. The features to prevent rupture of primary heat transport system piping ensure that the effects on the core from such leakage will be insigificant. A Exh 87 at 27-29, TR 7 404-7 406. A bounding DBA involving 4/

~~ As a result- of a matter raised by the NRC S taf f in relation to the core compaction DBA (A Exh 87 at 43-51, TR 7420-7428; see Finding 15, infra), Applicants have committed to impose an additional quantitative limit on internal fuel pellet melting for the OL analysis and review. A Exh 87 at 8, TR 7385 ; S Exh 26 at 4-44--4-45 ; S Eth 38 at 4-5, TR 8211-8212.

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~~ See A Exh 87 at 9-18, TR 7386-7395, for a description of the overall' HTS and its major components.

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increased core inlet temperature (complete loss of heat  ;

removal through an Intermediate Heat Exchanger (INX)) have been imposed on the design, and either the secondary or primary RSS and the SHRS will preclude sodium boiling in the core for these events. Iji. at 29-34, TR 7406-7411. Even in the event of a f ailure of heat removal from all three IHX's, which envelops the loss of all main and auxiliary feedwater, all three steam generator loops, and all three intermediate '

heat transport system (IHTS) loops, the primary or secondary .

RSS and the direct heat removal service (DHRS) will preclude sodium boiling. Id. at 34-36, TR 7411-7 413. Thus, the RSS and SHRS will function to limit, terminate, and mitigate all physically realizable reduced whole core heat removal t

sequences in CRBRP.

15. Significant increases in whole core heat generation can result only from either control rod withdrawal or compaction of fuel geometry. Id,l . at 36, TR 7413. Inherent features of the control rod drive mechanisms will physically limit i

withdrawal rates to less than 73 inches per minute, and withdrawal will be terminated by the- RSS without sodi um boiling.6/ -

Id. at 37-4 3, TR 7 414-7 420. The core and fuel i

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~~ Extensive testing has shown a maximum observed rate of 45 inches per minute. The system will be subjected- to inservice tests to assure that the withdrawal rate is well within the design basis ' of 73 inches per minute. A Exh 87 at 39, TR 7416.

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assembly mechanical design will limit core compaction and the resulting reactivity insertion, but even under assumed conditions of a) a sudden compaction of conservatively specified core clearances, b) a design basis reactivity insertion caused by instantaneous compaction closing all such clearances, and c) a safe shutdown earthquake causing retardation of control rod insertion and coastdown of all three primary pumps, no sodium boiling will occur. Ifl. at 43-51, TR 7420-7428. In addition, even assuming that the fission gas were released from an entire 217 rod subassembly, no sodium boiling will occur. Id. at 50-51, TR 7427-7428. Thus, the RSS and SHRS will limit, terminate, and mitigate all physically realizable increased whole core heat generation sequences in CRBRP.

16. The Applicants' detailed mechanistic analyses of the potential pathways for HCDA initiation confirmed that four systems or sets of design features are necessary to maintain core conditions so that reduced heat removal or excessive heat generation will remain within design basis conditions and not progress to HCDA conditions. A Exh 87 at 5-54, TR 1/ Some centerline fuel melting may occur, but no loss of coolable geometry would result. A Exh 87 at 43-51, TR 7420- l 7428. '

8/ The features which will prevent local, as distinct from whole core, imbalances in heat removal and heat generation from progressing to core-wide involvement are addressed in Finding 20, infra.

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7382-7431; see PID at 66-71. These are: a) the RSS; b) the SHRS; c) means to prevent rupture of Primary Heat Transport System (PHTS) inlet piping; and d) features to prevent local imbalance between heat generation and heat removal f rom progressing to significant core-wide involvement. Id. The manner in which each such system or set of f eatures will function to prevent, limit, terminate, and mitigate each pathway for progression to HCDA conditions is depicted as '

follows: ,

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The record evidence shows that each of these four sets of features has been designed to function with high l reliability, so that HCDA's can be excluded from the CRBRP l

design base.

+. ,

- F ,

Reactor Shutdown Systems i

17. To assure that HCDA's caused by a whole core imbalance between heat generation and heat removal can be prevented, l l

CRBRP has proposed a design consisting of two (rather than one, as in LWR's ) f ast-acting reactor shutdown systems (the primary and secondary shutdown systems), either of which by itself can reduce the reactor power level (and hence restore the balance between heat generation and heat removal) and ~

shut down the reactor when required. A Exh 87 at 9-5 3, TR 7386-7430; S Exh 32 at 21-24, TR 8056-8 059. The design of the systems will conform to all applicable criteria in the CRBPP Principal Design Criteria, NRC Standard Review Plan and Regulatory Guides, IEEE Standards, and the ASME Code,Section III. S Exh 32 at 21-22, TR 8056-8057 ; A Exh 87 at 79-81, TR 7456-7458. The two shutdown systems are based upon proven technology, are fast-acting, and are redundant, diverse, and independent in regard to sensors, logic, con-trol rod drive mechanisms, and control rods. A Exh 87 at L 55-85, TR 7432-7462; A Exhs 64 and 65, 5 4. 2. 3 ; A Exh 67, I 7.2; S Exh 26, $ 7. Both shutdown systems f unction automatically, and the role of the operator is to confirn RSS action. A Exh 87 at 81-82, TR 7458-7459. The major components and the appropriate integrated systems are being tested beyond the number of event cycles expected during the n,-, , . . , , . , , , - - , , , ,-.e-,- - . - - , . , , - - .- -,-,.-.v..-- ,--e.,-.., ,, -cc.-

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plant lifetime,9/ and both systems will be subject to periodic on-line functional testing during plant operation. A Exh 87 at 71, 81-84, TR 7448, 7 458-7461.

Shutdown Heat Removal Systems

18. After reactor shutdown is achieved, the SHRS proposed for i CRBRP will remove reactor core decay heat (and hence maintain the heat generation-heat removal balance) through four separate heat transport paths: a) through each of three PHTS and IHTS and steam generator system (SGS) loops with the Steam Generator Auxiliary Heat Removal System '

11 (SGMIRS ); / and b) by means of a diverse DHRS. Any one of the four paths is capable of removing all short and long

~

9/ The testing included testing with def ects deliberately imposed on the control rod drive mechanism bellows. These tests showed no adverse effects upon operation. A W Clare, TR 7650.

10/ The Applicants and NRC Staf f have considered the implica-tions of the recent Salem event, involving the failure of SCRAM breakers. A Exh 87 at 65-66, TR 7442-7443; S Exh 32 at 24, TR 8059; A W Clare, TR 7756. The CRBRP uses a different, enclosed SCRAM breaker design in the primary RSS which is not likely to degrade, and which has been tested without failure for over nine times the expected plant l cycles. A Exh 87 at 65-66, TR 7442-7443; A W Clare, TR 7756. Fur ther, unlike the Salem breakers, the CRBRP breakers can and will be periodically tested at power.

Id. In any event, the secondary RSS, which does not use TURAM breakers, will shut down the reactor even if the primary RSS should f ail due to SCRAM breaker failure. Id.

11/ Alternatively, heat can be removed through the turbine-l ge ne rator, condenser, and cooling towers. A Exh 87 at l 86-111, TR 7463-7488.

- F ~,

1 term decay heat from the reactor core. A Exh 87 at 86-111, TR 7463-7488; A Eih 67, $$ 5.0-5.7; S Exh 26, $$ 4, 5, 7, and 15. These systems, which conform to CRBRP Principal Design Criterion 35 and incorporate redundant, divers e, and independent features, are safety grade systems. Id.; S Exh 26, $ 3; see 10 C.F.R. Part 50, Appendix A, General Design Criterion 34. Any one of the PHTS/IHTS/SGS paths in conjunction with SGAHRS can remove reactor decay heat without the need for operator action. A Exh 87 at 94-9 6, l 104-110, TR 7471-7473, 7481-7487. All three paths have the diverse capability to remove decay heat via natural circulation even in the event of station blackout. 12/ A Exh 87 at 97-99, TR 7474-7 47 6; S Exh 32 at 26, TR 8061; S Exh

! 37, TR 8192-8196. Maintenance of suf ficient primary system inventory to enable adequate decay heat removal in the event of a leak is assured by a highly reliable passive means --

the use of guard vessels around the major primary system i

piping and elevated piping between those components. A Exh 87 at 99-102, TR 7476-7479; S Exh 32 at 27-28, TR 8062-8 06 3; S W King, TR 8148.

12/ See also Findings 42-43, infra, relating to Board Area of T3Ier_est 4. .

- F .

Means to Prevent Rupture of Primary Heat Transport System Inlet Piping

19. The proposed CRBRP design has incorporated a series of measures to prevent a pipe rupture in the PHTS reactor vessel inlet- piping, prevent sodium boiling in the core, and prevent HCDA initiating conditions. Since the operating temperature of the sodium coolant (700-1000* F) is below the atmospheric saturation temperature (1623' F), the sodium coolant is pressurized only to the extent necessary to pump the coolant through the primary system. Thus, there is no potential for flashing of the coolant into vapor due to loss of systen pressure, as in an LWR. A Exh 87 at 112-14, TR 7489-7491. The CRBRP primary coolant system piping and components are housed in a series of nitrogen-filled cells, which provide protection and a less hostile environment (from the standpoint of degradation processes) than in LWR service. A Exh 87 at 118-119, TR 7495-7496. The CRBRP can accommodate leaks substantially larger than a design basis leak without a significant reduction in heat removal. A 13/ See also Findings 40 -41, infra, relating to Board Area of Tnterest 3.

14/ The Applicants and Staff have given extensive consideration to sodium leaks and fires in the CRBRP design and design review, respectively, to assure that all necessary saf ety functions can be performed under those conditions. S Exh 32 at 33-35, TR 8068-8070; S Exh 26, $ $ 3.8. 3, 9.13, 13.2, 15.6.2; A Exh 72, $ $ 6. 0, 6. 2, 6.4-6. 5, 15.6; A Exh 87 at

, 200-204, TR 7577-7 581. Equipment has been properly quali-fled for a sodium aerosol and combustion product environ-(Continued)

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Exh 87 at 27-29, TR 7404-7406; S Exh 32 at 29, TR 8064; S l Exh 26, { 15. 3. 2. The CRBRP design incorporates a series of measures to assure that very small pipe leaks can be detected, and that piping leaks in excess of design basis values can be prevented. These measures include: 1) the use of tough ductile stainless steel, proven for high temperature service, as a piping material; 2) a material surveillance program; and 3) highly reliable redundant, -

diverse leak detection systems which can detect a leak several orders of magnitude below design basis values. A Exh 87 at 114-17, 119-122, TR 7491-7494, 7496-7499; S Exh 32 at 29, TR 8064. In addition, the CRBRP hast a ) speci fied stringent limitations on flaws in the piping materials to assure that the potential for crack initiation is minimized; b) conducted fracture mechanics analyses to show that if a flaw up to the material specification limits existed, it would not propagate to create a crack in the piping wall; c) completed a technology program to show that even if a crack did grow significantly, it would penetrate the pipe and be detected as a small leak prior to developing the potential for a large pipe break; and d) conducted additional crack propagation analyses and testing to show that any crack ment, the primary system is enclosed in inerted cells with liners designed as engineered safety f eatures, and catch pans and fire suppression decks have been provided in the air-filled cells containing non-PHTS sodium piping and i compone nts. Id. l l

I l l

- F ,

which would develop into a leak (even if undetected) would have dimensions well below those at which pipe rupture could

, occur. A Exh 87 at 122-129, TR 7499-7506; A Exh 88; S Exh l

32 at 29-30, TR 8064-8065.

Features and Capabilities to Prevent Local Imbalance Between Heat Generation and Heat Removal

, 20. The CRBRP design incorporates f eatures and inherent capa-bilities to assure that local imbalances between heat '

generation and heat removal do not progress to involvement of a significant portion of the core (and hence HCDA initi-ating conditions ) . A Eih 87 at 131-157, TR 7 508-7 534.

These features and capabilities are of two types: a) those to preclude mispositioning a fuel assembly in a location where it would receive inadequate flow; and b) those to preclude blockage of flow to an individual subassembly. I dl.

at 131, TR 7 508. Features of the former type consist of lower inlet module discriminator inserts and outlet nozzle identification notches, manual and computerized inventory systems, and a monitoring and detection capability provided by Source Range Flux Monitors. A Exh 87 at 131-135, l

TR 7 508-7 512. Features of the latter type include two cate-gories: a) features to preclude a reduction in flow to a limited region of the core; and b) f eatures to ensure that local fuel failures do not propagate to core wide involve-ment. A Exh 87 at 135, TR 7 512. In the first category, the

• . N l

- F .l CRBRP design provides a multiplicity of redundant flow paths in each core subassembly inlet, in the inlet modules that hold groups of subassemblies, and in the core support struc-ture that holds and supports the inlet modules. A Exh 87 at 136, TR 7 513. These redundant flow paths, which have been subjected to extensive scale model testing, provide an inherently reliable, passive means of preventing flow reduc-tion to a fuel assembly caused by foreign objects. Id. at 136-137, TR 7 513-7 514. These features are complemented by the capability to strain and remove particulates so that the potential for formation of blockages is minimized. Id. at 138, TR 7 515. Even so, the eff ects of assumed inlet blockages have been analyzed, and the design has margin to accommodate a substantial blockage without sodium boiling.

A Exh 87 at ~ 38-139, TR 7 515-7 516. Analyses of potential blockages within the fuel rod bundle similarly show that sodium boiling will not result. A Exh 87 at 139-140, TR 7516-7517. Experimental and analytical work based upon EBR-II experience, world-wide LMFBR operating experience, and analyses of the CRBRP design characteristics, all show that

{

propagation of local fuel rod f ailures beyond their immediate vicinity is highly unlikely. A Exh 87 at 143-147,  ;

TR 75 20-7524 ; A Exh 72, { 15. 4; S Exh 32 at 32, TR 8067 ; S Exh 26, { 15.4; S W King, TR 8149, 8151-8153. In addition, steel hexagonal subassembly ducts house each f uel rod bundle i

i

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- F .

and provide passive, inherent physical protection against propagation from one subassembly to another. A Exh 87 at 143, TR 7520. Finally, fuel failures can be detected by fission gas detectors and delayed neutron detectors at

, levels well below those that could result in significant imbalances in local heat generation and removal. A Exh 87 at 147-152, TR 7524-7529; S Exh 41 at 38, TR 8312. The fea-tures and capabilities to prevent local imbalances will be complemented by a comprehensive test program to establish the capability of CRBRP to operate with local fuel failures. Pending completion of that program, the NRC Staff has imposed operating restrictions to preclude significant local imbalances. A Exh 87 at 132-156, TR 7509-7533; S Exh 32 at 33, TR 8068; S Exh 26, S 4.2.1.3.2.6; S W King, TR 8154.

21. The criteria applied in consideration of whether HCDA's should be DBA's in CRBRP rest upon two sound basic prin-ciples: a) the design approach should incorporate the well established defense-in-depth concept, and b) the risks associated with accidents in CRBRP (including beyond design basis accidents) should be comparable to those for LWR's.

E.g., S Exh 26 at 1-2; S Exh 32 at 14, TR 8049. In applying

\ -

I the first of these principles, requirements have been i

established for redundancy, diversity, and independence of

, .. l

- F ,

systems for prevention of HCDA initiation that go beyond the equivalent requirements for LWR's. A Exh 87, TR 7378-7594; l

A Exh 89 at 2, TR 7764. In applying the second of these l basic principles, requirements have been established for design features aimed at limiting the residual risks of HCDA's. A Exh 89, TR 7763-7916; S Exh 23 at I-4. The Staff has developed and applied a set of CRBRP Principal Design Criteria which are similar to those in 10 C.F.R. Part 50,

• Appendix A, with appropriate modifications to reflect technological differences. S Exh 26, S 3; S Exh 32 at 17-19, TR 8052-8054; S W King, TR 8165; see 10 C.F.R. Part 50, Appendix A. These general criteria will be coupled with the specific criteria embodied in the existing generic NRC regulatory requirements (e.g., 10 C.F.R. Part 100, 10 C.F.R.

Part 73, 10 C.F.R. Part 50, Appendices B and R, Standard Review Plan). Ijl. _

22. Even though it is not an existing NRC regulatory require-ment, the Staff has imposed an additional requirement on CRBRP for a formal reliability assurance program. S Exh 32 at 50-54, TR 8085-8089; S Exh 27, App. C. This program is designed to enhance the safety-related reliability inherent in the CRBRP design. Id.; A Exh 87 at 158, TR 7535. All major CRBRP safety systems will be addressed, with substantial emphasis on the four sets of features necessary l

.. . _ - _ . - . . . . _ . - _ - . . . . - .- -. . - . . - . - = , - , - - . . - . - . - - - .

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for prevention of HCDA conditions. S Exh 32 at 57-59, TR 8092-8094; A Eih 87 at 158-159, TR 7535-7536. The program will employ a f ull range of analytical techniques, experience, and data and will provide for traceable information gathering, documentation, and feedback into the design and operation of CRBRP. S Exh 3 2 at 55-62, TR 8090-8097; A Exh 87 at 159-169, TR 7536-7 546. The program will be conducted throughout the entire CRBRP plant lifetime, and its results and implementation will be reviewed by the NRC Staf f. S Exh 3 2 at 62-64, TR 8097-8099; S W King, TR 8157.

The reliability program is considered by the Staf f to be a valuable means of assuring enhancement of the reliability inherent in the CRBRP design, but it has not been used as a present basis for exclusion of HCDA's from the CRBRP design base. S Exh 32 at 52-6 0, TR 8087-8 09 5. It is not i

anticipated that numerical goals or thresholds will be used l l

for determining feedback into the CRBRP design, since the analytical methods are not suf ficiently advanced to warrant application in that manner. S Exh 32 at 59-6 0, TR 8094-8095.

l

23. A comprehensive probabilistic risk assessment (PRA), com-parable in scope to WASH-1400, will be performed by the Applicants. S Exh 32 at 43-46, TR 8078-8081; A Exh 87 at 170-178, TR 7547-7555. The objective of the PRA is to define the relative importance of systems and components to

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reliability and safety, to identify system weaknesses, and to identify specific preventive or mitigative actions to l l

reduce risk.

15/

l S Exh 32 at 46-47, TR 8081-8082; A Exh 87 at 170, TR 7547. While the PRA is a useful tool or adjunct to assuring the safety of CRBRP, the state of the art is not suf ficiently mature to use or require a PRA as a decisive basis for determining the CRBRP design basis envelope at this time. S Eih 32 at 13, 44-46, TR 8048, 8079-8081; A Exh 87 at 176-177, TR 7 553-7 554; A W Clare, TR 7749; S W King, TR 8168-8169.

i 24. The potential for human error and common cause failure has been considered and actions to minimize them have been implemented in the design to assure that the likelihood that system interdependencies or human error could cause an HCDA is made extremely low. The general design characteristics of CRBBP include the use of: a) redundant, i ndepe ndent, di vers e, and automatically actuated or passive safety systems, and b) inherent physical characteristics which assure that rapid operator action will not be necessary in responding to accidents at CRBRP, and that the potential for 15/ Based upon the experience with the CRBPP design and prior PRA studies, and in light of the schedule for CRBRP con-str uction, it is expected that the results of the PRA can and will be f actored into the design, without signi ?icant impact on cost and schedule and without compromise of safety. S Exh 3 2 at 47-50, TR 8082-8085; A Exh 87 at 176-77, TR 7553-7554; see also Findings 112-113, infra, relating l to the Board Area cl~ Interest identified at TR 7340-41.

1

~ 1 F .

human error will be minimized. See Findings 17-20, supra.

The Staf f's review of the design accounted for system inter-1 dependencies and common cause f ailures by reliance upon well established principles enunciated in the CRBRP Principal Design Criteria, the Standard Review Plan, arid applicable Regulatory Guides. S Exh 32 at 40, TR 8075; S Exh 26,

$ 3.1. The Applicants have undertaken and will continue studies to assure that human error, system interdependencies -

and common cause f ailures will not compromise the reliabil-ity inherent in the redundant, diverse and independent systems of importance to the prevention of HCDA's. A Exh 87 l at 160-61,164, 168, 174, TR 7537-7538, 7541, 7545, 7551.

The reliability program and PRA will include consideration of human error and common cause f ailure, and f eed appropriate corrective action back into CRBRP design and operation. Id.; see Findings 22-23, supra. The post-TMI CP requirements relating to human factors have been reviewed by the Applicants and Staff and implemented for CRBRP. S Exh 26, { 18; A Exh 74, App. H; Cf . 10 C.F .R. $ 5 0. 34 ( f ) .

THE ADEQUACY OF APPLICANTS' AND STAFF'S HCDA ANALYSES

25. ~ Although the Applicants have designed the CRBRP so that HCDA's.are beyond the design basis, features and i

capabilities have been included in the design to provide additional margin to mitigate HCDA's and thereby limit the l residual risk of accidents beyond the design base to w, Tsw--eup--yms'-- w --- w -ea'-- w -wm- -iy-* -- *v---ww -Ne - *

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acceptable levels. A Exh 89 at 2, TR 7764. These features l l

and capabilities are designed tot a) maintain containment  !

integrity and b) control releases of radioactivity, in the ,

unlikely event that an HCDA should occur. Id.

26. Containment integrity might' be challenged in two ways: 1) internal missiles and 2) internal pressure. A Exh 89 at 3,

! TR 7765. The Applicants have provided two types of featur.es and capabilities to accommodate these challenges: 1)

Structural Margin Beyond the Design Base ( SMBDB ) , and 2)

Thermal Margin Beyond the Design Base (TMBDB). I d. The former addresses short-term (minutes or less) challenges, while the latter addresses longer-term (hours to months) challenges. Id.

27. In the short term, a challenge to containment could result from: a) a large prompt sodium release through the reactor 4

vessel head into the containment, sodium burning, and a resultant overpressurization of containment; or b) internal t

missiles from the reactor vessel head area with suf ficient energy to penetrate the containment. A Exh 89 at 4, TR 7766. Either of these challenges could result only if an

( HCDA occurred which imparted suf ficient energy to the reactor coolant boundary (specifically the reactor vessel head) to exceed its structural capability. I d. In either c as e, the objectives of the Applicants' analyses and the NRC Staf f's extensive independent review and analyses were to I

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assess: 1) the likelihood of energetic behavior, and 2) the l

capability of the CRBRP reactor coolant boundary to l l accommodate any energetic behavior. A Exh 89 at 4-5, TR l 7766-7767; S EWh 41 at 11-12, 16-17, TR 8282-8283, 8287-

! 8288. Specific dynamic load, leakage, and geometric I 1 requirements have been Laposed on the CRBPP reactor coolant i

boundary to assure that short-term challenges from HCDA's i can be accommodated. A Exh 89 at 5-9, TR 7767-7771. The l Applicants' analyses showed that the likely outcome of an HCDA sequence would be well within the structural capability

! of the reactor coolant boundary. A Exh 89 at 4-5, TR 77 66-

! 7767. The Staf f's independent analyses and review concluded i that HCDA's with energetics suf ficient to fail the reactor i

l vessel head are physically unreasonable and not a l

I significant saf ety concern for CRBP@. S Exh 41 at 6, TR i

8275.

28. The Staf f's review and analysis defined three generic cases i

! which represented a bounding range of initiating I

conditions: 1) the unprotected (without scram) loss-of-flow I

(LOF) accident; 2) the unprotected transient overpower (TOP)

accident; and 3) the loss of heat sink (LORS) accident.

1 S Exh 41 at 21-23, TR 8292-8294. The Staf f's analysis f followed these initiators through the disruption phase until l

termination of neutronic activity (achievement of permanent l

subcriticality by f uel removal), and included consideration

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of an appropriate range of phenomena which significantly influence energetics and termination. S Exh 41 at 24-26; TR 8295-8297. The Staf f determined that accident energetics characterized by values of work done in isentropic expansion  !

to one atmosphere of up to 1130 megajoules (MJ) would produce minimal dynamic loadings on the reactor coolant system boundary because of mitigation of 'the fuel expansion due to the core barrel, upper internal structure, and core -

support structure. Fur ther, the Staf f determined that acci-dent energetics characterized by 2550 NU work done in an j isentropic expansion to one atmosphere would be necessary to i approach the structural capability of the reactor head (75 M7 slug impact kinetic energy), taking proper account of the internal structure of the reactor.

17/ S Exh 41 at 28-34, TR 8300-8307. These levels of energetics, 1130 MJ and 2550 MJ, 16/ During the initial stages of the Staf f energetics review, eight speci fic areas of concern were developed. Each has been resolved to the Staf f's satisfaction. S Exh 41 at 51-60, TR 8325-8334; A Exh 89 at 51-6 0, TR 7813-7822. See l Findings 86-94, infra, relating to Board Area of Interest 17.

17/ The Staf f did identify a potential kinematic f ailure mode, involving the rotating plugs in the reactor head, which could challenge the structural integrity of the reactor head. S Exh 41 at 34-36, TR 8307-8309; - A Exh 89 at 142-143; TR 7904-7905 ; A W Strawbridge, TR 7939-7942. The Applicants have committed to f urther analysis and additional testing, and have developed a f easible design modification. Id; S W Butler, TR 8447. The Staff has reviewed the modification and test program, and concluded that this  !

matter can be satisf actorily resolved. Id.

1

• 1

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correspond to disassemblies driven by reactivity ramp rates of 100 dollars per second ($/sec. ) and 200 $/sec.,

respectively, occurring in the two-phase regime. The expected ramp rates during HCDA's are significantly lower than these val ues . S Exh 41 at 36-49, TR 8309-83 23. The Staff concluded that HCDA induced reactor vessel head failure is physically unreasonable.-1 9/ -

' S Exh 41 at 47-49, TR 8321-8323; S W Theofanous, TR 8446. Based on the projected .

absence of signi ficant energetic events, the Staf f has concluded that the Applicants' SMBDB energetics level of 661 MJ (75 MJ slug impact kinetic energy), as applied by Applicants in the SMBDB evaluation, is adequate. S Exh 27, App. A at A . 2-ll .

29. While the Applicants' approach to the analysis differed from the S taf f's, their ultimate conclusions were similar. The 18/

Similarly, the Staf f review indicated that during the LOF sequence plenum fission gas induced compaction could create a ramp rate in the order of 50 $/sec, and a potential for Saugmentatien Exh 41 at of reactivity in the unvoided core regions.

l TR 7854-7855,38-40, TR 8312-8314; A Exh 89 at 92-93, 107-108, 7869-7870. Applicants have committed to l

conduct further analyses to resolve the concern, I

l implement a simple, f easible design modificationorto limit the energetics potential of this phenomenon. Id; A W Fauske, TR 7968; S W Theofanous, TR 8457.

19/

The Sta'f f's evaluation was based upon the 75 MJ slug impact

kinetic energy value stated by the Applicants, and did not consider lower values. Although a lower value might be ade quate, at 75 MJ the Staff concluded that f ailure of the reactor vessel head is physically unreasonable. SW Theofanous, TR 8444-8446.

- F ~

Applicants analyzed an appropriate range of initiating events, accident phenomenology, and accident regimes, and determined that the: likely outcome of HCDA sequences was non-energetic. A Exh 89 at 61-128, TR 7823-7890.

Notwithstanding this conclusion, dynamic load requirements have been -imposed on the CRBRP to assure that structural margin exists to accommodate a wide range of HCDA conse quences . The load calculations were based en an '

assumed accident having energetics characterized by about 660 MJ work done in isentropic expansion to one atmosphere. In performing the calculation of loading on the reactor coolant boundary, the Applicants conservatively neglected the ef fects of the upper internals structure in mitigating the expansion. A Exh 89 at 129-130, TR 7891-7892. The structural analyses based upon the conservatively calculated loads showed that the CRBRP has substantial margin in the reactor coolant boundary to accommodate HCDA e nerge tics. Based upon the Applicants' and Staf f's analys es , it feilt s s nat short-term containment integrity i

would be maintAinea 1.< the event of an HCDA. In that regime, the design basis containment / confinement system would limit any radiological releases to acceptable levels. A Exh 89 at 5, TR 7767 ; sea A Exh 87 at 19 7-211, TR 20 / But see note 17 supra. . i

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1 F-3 2 - ,

j 7574-7588. With this result, the focus of inqpiry shifts to l the capability of the CRBRP design to maintain containment integrity and control of any longer-term radiological releases.

30. Even though there is no short-term threat ~ to containment

{

posed by energetic HCDA's, in the longer t- erm thermal pene-tration of the reactor vessel and guard vessel could occur, with a resultant release of fuel, fission products, and '

sodium into the reactor cavity and with resultant longer-term challenges to containment integrity of two types: a) overpressurization due to sodium and hydrogen burning and decay heat; and b) overpressurization by non-condensable gases (principally hydrogen, if it does not burn). A Exh 89 at 10, TR 7772. In either case, the objectives of the Applicants' analyses and the NRC S taf f's extensive indepen-dent review and analyses were to assess the capability of the CRBPP design features to a) avoid challenges from over-pressurization, and b) control radiological releases to the environment. A Exh 89 at 10-11, TR 7772-7773. Speci fic design f eatures have been provided in CRBRP, including a reactor cavity vent system, a containment annulus cooling 1

system, a containment vent system, a containment purge

)

system, and a containment cleanup system, to avoid longer-

- term overpressurization and to control the radiological consequences of HCLA's. A Exh 89 at 11-24, TR 7773-778 6.

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The Applicants' analyses and the Staff's independent review and analyses showed that the containment would not be challenged by overpressurization for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation of an HCDA, and that even then, venting the containment through the cleanup system to maintain contain-ment integrity and control releases would not, based upon 4

realistic assessments, result in doses exceeding the 10 C.F.R. Part 100 dose guidelines. A Exh 89 at 10-60, TR -

7772-7822; S Exh 41 at 61-115, TR 8335-8393; S W L. Bell, TR 8472; S Exh 30 at A.5-4.

31. The Applicants' analyses and the Staff's independent review and analyses extensively examined the parameters of impor-tance to post-core melt-through HCDA behavior, including sodium-concrete reactions, aerosol behavior, heat transfer effects, hydrogen generation, and containment structural 1

response, and evaluated the performance of the specific CRBRP features designed to mitigate core melt-through events and radiological consequences. S Exh 41 at 61-115, TR 8335-8393; A Exh 89 at 10-60, TR 7772-7822; S Exh 27, App A 21/ The Staff's best calculations indicate that the base mat will be not be breached in the event of core melt-through.

S W Long, TR 8494. Notwithstanding this, the Staff reviewed the hydrology of the site of assure that even in the event of such a breach, the radiological consequences would be acceptable. Id. The potential for post melt-through criti-cality was examined by the Staff and determined to be negli-gible in its effect. S W Long, TR 8495-8497; S W C. Bell, TR 8497-8499.

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at A.4-1 -- A.5-17; S Exh 30; A Exh 91. The Staff concluded that the studies of possible variations at each juncture had established conservative bounds to accident behavior, that the design features were acceptable and feasible, that the uncertainties in radiological consequences were such that i

only uncertainties related to iodine isotopes presented a significant potential for exceeding the 10 C.F.R Part 100 dose guidelines, and that if those uncertainties prevail at j the OL stage, additional filtration could be provided.

S Exh 41 at 98-99, 110-111, TR 8376-8377, 8388-8389. The Applicants also considered extensive variations in parameters and concluded that the radiological consequences were below the 10 C.F.R. Part 100 guidelines or equivalent

, values, that the results were relatively insensitive to short term releases of materials into the containment or to the time at which venting occurs (variations in vent time from a base case of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to a bounding case of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> showed doses within the Part 100 dose guidelines'or equivalent), and that the consequences of an HCDA in CRBRP are acceptably low. A Exh 89 at 58-60, TR 7820-7822; A W l

Strawbridge, TR 7954-7956; A Exh 91, App K.2.

Board Area of Interest 1 In its Safety Goal Development Program announcement (48 Fed. Jygl.10772, March 14, 1983) the Commission stated that during the 90-day period (ending June 8,1983) for public comment on the proposed evaluation

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plan "it is expected that preliminary 1 information on new radiological source terms i will become available . . . " (Id., at '

10778). The Staff is requeste3 to advise whether that information will be evaluated for any impact on this proceeding, and the reason for its answer.

'_ 32. The Accident Source Term Program Office plans to address the severe accident source term for LWR's, and not' for LMFBR's, since the latter involve different coolant, fuel, and design. S Exh 41 at 115, TR 8393. -

33. Two source terms were used to evaluate the CRBRP design from a safety perspective: a) the source term used for site suitability, and b) the source term for HCDA's in which the beyond design basis mitigating f eatures are operating. In addition, source terms were developed to evaluate the environmental consequences of accidents at CRBRP for the CRBRP FES Supplement. The Site Suitability Source Term (SSST) methodology parallels that used for LWR's and is based on TID 14844, a ref erence document footnoted in 10 C.F.R. Part 100. The SSST was used to bound DBA's for the site suitability determination in combination with

, engineered saf ety f eatures. A change in the SSST or environmental source terms in a more conservative direction is not likely to result f rom the efforts of the Accident Source Term Program O ffice. If the Accident Source Term Program Of fice efforts do indicate that the CRBRP SSST or 1 environmental source terms may need to be modified in a :nore l

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conservative direction, it is unlikely that the Staf f's l conclusions would change with respect to the suitability of i

the Clinch River site or the environmental consequences of CRBBP operation. Howevor, changes in the SSST or environmental source terms may require modifications to the design of certain engineered safety features, such as the annulus filtration system. The Staf f expects that such changes co6 4 be accommodated by the CRBRP design. S Exh 41 at 116-117, TR 8394-8395; S W Hulman, TR 8510-8514.

34. The source term used by the Staff to evaluate HCDA doses for CRBRP has no parallel in LWR's. The CRBP@ source term accounts for fuel configuration, aerosol behavior and deposition within the reactor cavity and vent ducts, etc.

Such considerations were not explicitly evaluated for LWR's, but are expected to be factored into the development of the LWR severe accident source term by the NRC Accident Source Term Program O ffice.

Since such considerations have generally been evaluated for the CRBRP, a re-evaluation of them in light of a revised LWR source term would not be expected to produce larger source term estimates for the CRB RP .

It is unlikely, therefore, that the Accident Source Term Program Of fice findings will appreciably alter the Staf f's HCDA source term or af f ect the S taf f's conclusion that the calculated HCDA doses are below the 10 C.F.R. Part 100 guidelines. S Exh 41 at 117, TR 8395.

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35. The Accident Source Term Program Office deliberations should not result in findings that would substantially affect the  !

CRBRP construction permit proceeding. However, in the unlikely event that a design change in the Annulus Filtration System and/or the Vent / Purge Filtration System (such as by increasing filtration ef ficiency) would be needed as a result of the Accident Source Term Program of fice findings, it is expected that such changes could be -

accommodated by the CRBRP design. The Staf f will ensure the conclusions reached by the Accident Source Term Program office receive appropriate consideration during the OL stage of review. S Exh 41 at 117-118, TR 8395-8396.

Board Area of Interest 2 As regards fuel performance, to date the use of the term "f ailed f uel" has not consistently permitted delineation of the various f ailure modes that might have been alluded to (e. g. , clad perforation, fission product leakage, clad bulging or rupture, melting of fuel pellets, etc.). The Applicants are requested to summarize the anticipated performance of the CRBR fuel associated with normal operation and accidental transients, describe various failure modes that must be dealt with, identify any operational limits (e . g. ,

maximum linear heat generation rates, maximum cladding hot spot temperatures, etc.) to be imposed, and to review the basis for confidence (e . g. , supportive evidence) that the proposed fuel behavior characteristics will be realized.

l l 36. Failed f uel ref ers to any loss of cladding integrity resulting from either 1) unpredictable conditions such as

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f abrication faults resulting from weld porosity, and 2) mechanistic f ailures from excessive strains caused by internal gas pressure or fuel-cladding-mechanical interaction. A Exh 87 at 182, TR 7559. The f ailure modes  !

may range from a thin small crack to a larger longitudinal rupture and may allow sporadic or continuous release of fission gas or the release of delayed neutron precursors resulting from direct fuel-sodium contact. A Exh 87 at 182, TR 7559.

37. Applicants have imposed overall design requirements on the CRBRP fuel to deal with four levels of reactor conditions:

1) normal operation, 2) anticipated transients, 3) unlikely transients, and 4) extremely unlikely transients. A Exh 87 at 185, TR 7 56 2. Design limits and acceptance guidelines for the fuel for each level of reactor conditions were established to facilitate demonstration that the CRBB@ fuel design requirements will be met. A Exh 8 7 at 183-184, TR 7560-7561. Because the structural capability of the cladding is a function of its ductility-limited strain and cumulative damage f unction, the f uel performance predictions took into account operating conditions such as temperature and pressure and damage mechanisms such as creep, irra-diation eff ects and f atigue damage. A Exh 87 at 184, TR 7561. The application of these design requirements, design limits and acceptance guidelines ensure that even in the 1

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case of extremely unlikely transients, the f uel will be maintained in a coolable configuration. A Exh 87 at 183-185, TR 7 560-7 56 2.

38. Analyses of operating conditions and cladding properties as well as a large experimental data base from tests conducted at the EB R-II and TREAT reactors demonstrate that the overall design requirements will be met by the CRBRP fuel.

Additional data and experience will be obtained from FFTF -

before operation of the CRBRP. Moreover, during operation of CRB RP , the reactor mixed mean inlet and outlet temperature, and reactor power, as well as outlet coolant temperature of most of the fuel and blanket assemblies, will be monitored to estimate cladding temperatures and predict the capability of the fuel. A Exh 87 at 186, 'TR 7 563.

Board Area of Interest 3 Avoidance of primary coolant pipe rupture seems to depend in part upon the f act that coolant temperature is well below its boiling temperature and that coolant pressure is near atmospheric pressure ( 10 atmos.).

Applicants are requested to present a tech-nical summary of how these coolant charac-teristics will result in a reduced likelihood of pipe rupture in piping designed for CRBR use.

39. The principal cause of sudden, burst-type pipe f ailures is primary stresses on the piping walls. Primary stresses include membrane stresses and bending stresses. Piping internal pressure is the principal contributor to piping i

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primary membrane stress. Because sodium is a coolant with a high boiling temperature, thus allowing operation at near atmospheric pressure, the primary membrane stress is only one-third of the allowable stress under normal CRBRP operating conditions and approximately one-sixth of the 22 allowable stress for accident conditions.-- / A Exh 87 at 113, TR 7490; A W Clare, TR 7622-7625.

40. The relationship between primary stresses and the potential for pipe rupture has been quantified ex.perimentally for CRBRP using the concept of " critical crack length." The critical crack length is the length of a through-wall crack which will rapidly open or grow as a result of the applied load. The critical crack (circumferential or longitudinal crack) length for the CRBRP ' inlet pipe for the PHTS under normal operating pressures is 30 inches. A Exh 87 at 113-114, TR 7490-7491; A W Clare, TR 7632; A Exh 88, $ 4. 3.

41. The critical crack length of 30 inches is large compared to the length at which a developing crack would be detected. A Exh 87 at 114, TR 7491. The leak detection system is capable of detecting a 100 grams per hour leak, which is l 22/

--- The CRBRP piping will be subjected to extensive quality 1

assurance tests for chemical composition. A W Clare, TR 7629-7630; A Exh 87 at 124, TR 7501.

23/ The Applicants' analyses properly included consideration of the fact that flaws are more likely to occur in weldments than in PHTS piping base metal. A W Clare, TR 7 626-7628; A Exh 88, $ 4. 2. 2.12.

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orders of magnitude below the design basis leak, and which in turn, is orders of magnitude below a double-ended guillotine rupture. A Exh 87 at 119, TR 7496.

Board Area of Interest 4 Applicants a're requested to explain how the CRBR will be configured to assure that convective circulation of the sodium coolant will be available to prevent fuel damage, if ne ed ed. This explanation should reference

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any supportive experimental or operational evidence. The Staff is requested to advise the Board whether it accepts convective circulation as a viable mechanism for fuel protection, and the reason for its answer.

42. The three HTS flow paths are designed to transfer decay heat from the reactor to SGAHRS by natural circulation, if necessitated by loss of off-site and on-site power. This inherent capability results from placement of the thermal centers of the heat exchanging components at successively increasing elevations in the plant, thus providing the necessary thermal driving head for adequate loop flows. A Exh 87 at 97, TR 7474; S Exh 37 at 2-3, TR 8193-8194.

Removal of decay heat via natural circulation will be initiated without any operator action. S Exh 37 at 3,TR 8194. Because of the natural circulation capability of CRB RP , decay heat removal can be maintained in the event of loss of all of f-site power and loss of all three on-site 1

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diesel generators. A Exh 87 at 99, TR 7476; S Exh 37 at 3-4, TR 8194-8195; S W King, TR 8198-8204; S Exh 26, { {4 and

5. 6 . 3.

43. The natural circulation capability of CRBRP has been i

verified by both Applicants' and Staf f's analyses, which demonstrate that in-core sodium temperatures during natural circulation cooling would be well below the sodium boiling tempe ratures. A Exh 87 at 98, TR 7475; A W Clare, TR 7631; ~

S Exh 37 at 5, TR 8196. In addition, evaluations of data from EBR.-II natural circulation tests and of predictions (both pre-test and post-test) and data from FFTF natural circulation tests confirm the Staf f's and Applicants' i analys es. Id.; S W King, TR 8198-8201. Additional verification is planned to support f urther refinement of these analyses. A Exh 87 at 98-99, TR 7475-7476. Natural circulation will be demonstrated in CRBPP during the initial start-up testing program. S Exh 37 at 5, TR 8196.

Board Area of Interest 5 In the area of quality, the Applicants are requested . to explain whether (and/or how)

!. differing f unctional levels of ef fort will be applied, depending upon whether a component or system is necessary for safety, important to safety, or not safety related. The divisions of authority and f unctional responsibilities for quality assurance and quality control amongst the various contractors and the Applicants should be dis-cussed with emphasis on how the management of the various CRBR contractor f abrication and construction ef forts will be coordinated to

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i assure the minimizing of QA and QC over-sights, especially where interfacing is involved. Applicants are also requested to describe what efforts will be undertaken to insure that accurate as-built plans and specifications will be available when needed, j if the CRBR is constructed. ,

44. Applicants have established a comprehensive quality assur-l l

ance (QA)-quality control (QC) program using a mr.nagement system of checks and balances. A Exh 95 at 5, TR 8629. The CRBRP Project Office, which has central management and control over project activities, assigned responsibilities to each of the major contractors, including the responsi-bility to plan, implement, and manage integrated qutlity assurance programs over the particular contractor's contrac-tual scope of work. A Exh 95 at 4-5, TR 8628 - 8629.

45. In order to ensure that the program QA functions are coordinated and integrated, the project has established three levels of control. The first level of control includes the system, component, material and service suppliers. Their quality assurance programs are primarily quality control programs concerned with direct control and verification through analysis, review, inspection, examina-tion and testing. This level requires the performer of an 24/ Within the CRBRP Project Office and the participants' organizations, the QA officials have been granted the neces-sary independent authority to assure effective program implementation. A W Karr, TR 8750; A W Hedges, TR 8751; A W Anderson, TR 8752-8753.

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- F .l activity to implement a system of checks and balances that provides direct control over his work process. The second level of control includes the program participants that have i direct or indirect interfaces with each other,and the l Project Office. The nuclear steam supply system (NSSS) supplier and the constructor are examples of this level of participation. These portions of the overall program are

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management-type programs with responsibilities for the quality assurance functions such as surveillance, audit, interface coordination, and lower-tier program integration functions, including overview and audit of the lower-tier quality control processes. The Project Office portion of

the program is the third level of control. The Project

, Office is responsible for the overall program and its ade-quacy. The Project Office program is a management-type program with audit and surveillance activities for verifica- .

tion of first and second level participant performance, l

interface coordination and program integration functions, including the coordination of fabrication and construction l

efforts for the project. A Exh 95 at 11-12, TR 8635 - 8636; A W Karr, TR 8669 - 8670. The three level control system provides the necessary inspections and review functions, the verification of those functions and the checks and rechecks necessary to assure the quality required for the plant. A

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Exh 95 at 12, TR 8636 ; A W Hedges, TR 8745 ; A W Clare, TR 8748.

t 46. The entire quality assurance program is a major part of the management control systems which cut across all levels of project activities. A W Hedges, TR 8737. A strong Project Of fice organization to coordinate and integrate the design, f abrication and construction ef fort serves to mini-mize problems with quality, especially Where interfacing is involved. The coordination of interfacing systems is con-trolled through a formal review and approval cycle that pro-vides the necessary safeguards for proper system integra-tion. A Exh 95 at 12-13, TR 8636 - 8637; A W Hedges, TR 8673 - 8674, 8679; A W Anderson, TR 8675 - 8677. Interface control is also a fundamental part of the design control process. Through a four-tier design control system, Which includes design reviews and configuration management, the interf ace requirements between systems are defined, translated into project specifications, and enforced. A Exh 95 at 14-17, TR 8638 - 8 641. The quality assurance program

, assures that throughout the process of design, manuf actur-ing, construction and operation, procedures are adhered to

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25/ The requirements of the Management Policy and Requirements (MPR) document are implemented by all project participants through formal procedures and training programs. A W Clare, TR 8717; A W Hedges, TR 8718-8719. The MPR and QA system have been part of the project since its inception. AW Anderson, TR 8719-8721.

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and documentation is both traceable and complete. A Exh 95

, at 15, TR 8639.  !

47. During the design control process, each system and component is characterized as to its plant function. The charac-terization of f unction provides appropriate consideration for the importance of the function relative to safety. The project has defined four categories for systems and components based upon the following functions: 1) Permanent Plant-Saf ety Related, 2) Permanent Plant-Operationally Signi ficant, 3) Permanent Plantdion-Operationally Related, and 4) Non- Permanent Plant. A Exh 95 at 17-20, TR 8641 -

8644.

48. The quality assurance program is applied, in a graded manner, to all CRBRP systems, structures, and components.

In implementing the project's graded approach to quality assurance, nine levels of program requirements have been developed which are applied based on the importance of items or services to the plant's function. These nine levels range from total quality assurance management (10 C.F .R. 50, Appendix B , RDF F 2- 2 ) to the standard manufacturing process i

of a supplier. Within the nine levels, both the level of quality assurance management and administration and the degree of quality control inspection, testing and documenta-tion are varied. The selection of the appropriate level of quality assurance is made using the technical judgment of l

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the design engineer, the cognizant Project O ffice engineer, and the quality assurance engineers in the design and Project Office organizations. A W Anderson, TR 8728 -

8729. To guide the selection, a matrix has been developed which takes into account such f actors as service f unction, plant application, safety category or classification, t emp-erature application and production category. Based on these criteria, the matrix provides recommended levels of quality -

assurance to apply to the plant item. For each plant item, the recommended quality assurance programs and activities derived fion the matrix are specifically reviewed and examined by the cognizant technical and quality assurance disciplines to assure that the appropriate level of QA effort is applied. A Exh 95 at 20-26, TR 8644 - 865 0.

49. The project has, through its configuration management system, ensured that accurate as-built plans and specifica-tions will be available when needed. Configuration manage-ment formally app. oves and establishes the reference plant design on a syscem basis. Once formally approved, i . e. ,

"baselined," any changes to that design require review and approval before the change is implemented. A Exh 95 at 17, 27, TR 86 41, 86 51; A W Clare, TR 8684 - 8686; A W Karr, TR 8688 - 8691. The configuration management system has been implemented and imposed as a requirement on all project participants. A W Anderson, TR 8697. The baselined docu- ,

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mentation will be required to reflect the as-built configuration of that structure, system or component. AW Karr, TR 8732-8733. This documentation is maintained and stored in a quality records system which provides the neces-sary documentation retrieval system for using the baselined documentation. A Exh 95 at 27, TR 86 51.

Board Area of Interest 6 The SER discussion of quality seems to emphasize quality assurance and the various separate contractor organizations that will implement it. Does the Staff consider that QC responsibilities and activities are separate from QA or an integral part thereof? The Staf f is requested to discuss its answer to this question and to explain ~

briefly how it~ will monitor QA and QC ef forts for adeqpacy.

50. Both Staf f and Applicants have treated quality control as an integrLA part of quality assurance. S Eth 44 at 2-3, TR 87 61 - 8762; A W Karr, TR 8669. Quality assurance consists of all those planned and systematic actions necessary to provide adequate confidence that a structure, system or component will perform satisfactorily in service. Quality control is included in quality assurance and consists of those quality assurance actions i

26/ The project has also implemented systems to assure that problems or concerns identified by those working on the

project are freely surf aced, ventilated and resolved. AW Anderson, TR 8738; A W Hedges, TR 8739; A W Karr, TR 8740; A W Clare, TR 8741.

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related to the physical characteristics of a material, structure, component, or system, which provide a means to control the quality of the material, structure, component, or system to predetermined requirements. A Exh 95 at 4, TR 8628; S Exh 44 at 2, TR 8761.

51. The NRC Staff will monitor Applicants' quality assurance and quality control activities throughout construction of CRBRP. S W Ignatonis, TR 8770 - 8771. The Staff has developed inspection procedures for pre-construction permit work. Inspections started in the first half of 1983, and a l resident inspector will be assigned to the site. SW Ignatonis, TR 8770. The Staff is in the process of develop-ing a construction inspection program for CRBRP based on the Staf f's IE Manual for Construction Inspection of LWR's, with l

appropriate modifications for CRBRP. S Exh 44 at 3-4, TR 8762 - 8763; S Exh 26, S 17.5; S W Brownlee, TR 8775 -

8778. The Staff's review will involve review of documenta-tion by Applicants and their contractors as to procedures, inspection and test reports, manufacturing data, specifica-tions, drawings and the like. In addition to the CRBRP, the Staff will also inspect CRBRP fuel fabrication activities.

l S Exh 44 at 4, TR 8763.

Board Area of Interest 7 l

Applicants are requested to discuss commercial and recreational river traffic (if any) from two points of interest:

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a) Practical methods of controlling same during of f-normal plant conditions, and b) The potential for hazardous cargo posing a threat to the CRBR.

52. Applicants will implement ef fective control measures over commercial and recreational traf fic in the Clinch River within both the 10 mile plume exposure pathway emergency planning zone (EPZ) and the portion of the Clinch River immediately adjacent to the CRBRP within the exclusion boundary. Traf fic within the EPZ will be controlled by the Tennessee Wildlife Resources Agency (TWRA), assisted if necessary by the U.S. Coast Guard. Upstream lockage through Melton Hill Dam will be controlled by the U.S. Army Corp of Engineers. A Exh 94 at 19, TR 7997. Control of the exclusion boundary will be accomplished through arrangements with TWRA and the Coast Guard and will include prompt warning and removal of persons in the area. A Exh 94 at 19-20, TR 7997-7998.

53. Barge traf fic past the CRBRP site would be either going to or coming from the Melten Hill Reservoir, since there are no origins or destinations between the site and Melton Hill Dam. In the past, no hazardous materials have been transported by barge past the site. In the f uture, coal and steel are the only materir.ls likely to be shipped by barge. A Exh 94 at 20-21, TR 7998-7999. In addition, there are only a few industrial sites in the vicinity of the CRBRP

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which could support an industry using or producing commo-dities in large enough quantities to take advantage of water transportation. A Exh 94 at 21, TR 7999. It is unlikely 4

that any hazardous cargo would be shipped by barge past the l CRB RP site. A Exh 94 at 21, TR 7999. In the unlikely event l that a new industry which would ship hazardous material by barge were located on the Melton Hill Reservoir, plans for a barge f acility would have to be submitted to the Corps of Engineers for review and approval. This application would have to identify any hazardous materials to be shipped, if known, and if none were identified, the permit wculd allow only non-hazardous materials. A Exh 94 at 22, TR 8000.

Board Area of Interest 8 Applicants are requested to discuss the design characteristics of the containment /

confinement structures and the steam generator, with respect to challenges to those structures arising from transient (or accident) induced overpressure and over-temperature conditions. This discussion should address any engineered safety systems or components that will be relied upon for protection ( e. g. , containment shell cooling),

and should reference supportive test or operational experience.

54. The steam generator modules in each IHTS loop are arranged as a superheater in series with two parallel evaporators.

IHTS sodium on the shell side of the heat exchanger transfers heat to water or steam on the tube side. The shell and tube bundles are shaped in a " hockey stick"

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configuration . A Exh 87 at 14-15, TR 7391-7 392. The structural design of the steam generator modules accounts for both steady state and transient loads. The most severe thermal transient imposed on the steam generator is that predicted to result from postulated water side isolation and dump of an evaporator with concurrent failure of the water inlet isolation valve. Because of its hockey stick configuration, the steam generator module will accommodate the differential expansion between the tube and shell resulting from this temperature transient without interference or excessive stresses. A Exh 87 at 188, TR 7565; A W Clare, TR 7737. The ability of the steam generators to withstand thermal transients will be confirmed by testing prior to f abrication of the steam generator modules for the plant. A Exh 87 at 188-189, TR 7 565-7 566; A W Clare, TR 7737.

55. The steam generator design accommodates the three nuclear safety considerations related to the steam generator modules

-- shutdown heat removal, control of releases of radioactive material, and mitigation of eff ects from sodium-water reactions. The modules remove shutdown heat.from the IHTS sodium as part of the SHRS. Because of the redundancy of the heat removal paths, the steam generator modules are not required to have a reliability greater than conventional SHRS components. A Exh 87 at 189-190, TR 7 566-7 567.

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Fur thermore , since neither the IHTS sodium nor the steam generator water / steam is significantly radioactive, accidents involving release of these materials are not a significant nuclear safety concern. A Exh 87 at 190, TR 7567.

56 . The primary nuclear safety consideration related to the steam generator modules is mitigation of the eff ects on the

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IHX from a sodium-water reaction. A Exh 87 at 190, TR 7 567.

) 57 . The steam generator modules have been designed with special features to both prevent tube leaks and to mitigate the ef f ects of any leaks which might occur. A Exh 87 at 190-19 7, TR 7 567-757 4. In the area of prevention, special attention has been given to the choice of materials, i

design, inspection procedures, and construction techniques, 27 to ensure the integrity of the steam generator tubes.- / A l

l Exh 87 at 191-192, TR 7568-7569.

58 . Three levels of protection are also provided to mitigate steam generator tube leaks and ensure the integrity of the IHX. First, CRBRP has been designed with a sensitive leak detection system capable of detecting leaks as small as 27/

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The CRBRP design has anticipated the primary problems which have been experienced with LWR steam generators. The design includes a full flow feedwater deaerator, a full flow demine ralize r, and a complete system for feedwater chemistry control. Fur ther, the steam generator water side geometry is free of discontinuities and the tube-to-tube sheet weld

.has been specifically designed to eliminate crevices. AW Clare, TR 7738-7740; A Exh 87 at 119, TR 7496.

I

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2xlO-5 lb. water / second and alerting plant operators, thus allowing prompt corrective action. A Exh 87 at 193, TR l 7570. Second, a rupture disc is provided on the cover gas space of the IHTS expansion tank to relieve the slow l pressure rise associated w"ith a postulated intermediate-size j tube leak. Bursting of this rupture disc would result in automatic plant shutdown and water side isolation of the steam generator modules in the af fected loop. A Exh 87 at I 193, TR 7 57 0. Third, CRB RP is designed with an engineered safety feature, the Sodium Water Reaction Pressure Relief System (SWRP RS ) , which will limit loadings in the IIIX to an acceptable level in the event of a postulated large size tube leak. A Exh 87 at 193-194, TR 7 57 0-7 571. The maximum pressure predicted at the IHX in the event of a large tube leak is 331 psia. This pressure is within the structural capability of the IHX and thus would not result in release of radioactive primary sodium, A Exh 87 at 19 5, TR 7 572.

59. A large body of experimental anJ analytical data exists which demonstrates the adequacy of the SWP. PRS . These data l verify that the design basis steam generator leak imposed on i

I the SWRP RS is conservative. The data also verify the l

I results of the TRANSWRAP computer code used to evaluate the ef fects of steam generator tube leaks in CRB RP . A Exh 87 at 19 5, TR 7572.

P  %

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60. The CRBRP Containment System (CS) contains a number of engineered safety f eatures to mitigate the consequences of postulated design basis accidents. The CS consists of a 186 foot diameter cylindrical carbon steel containment vessel surrounded by a reinforced concrete confinement building, with an annulus between the two. A Exh 87 at 197-198, TR 7574-7575. A pressure of 1/4 inch water gauge lower than atmospheric is maintained in the annulus between the containment vessel and the confinement building to ensure that any potential leakage from the containment vessel would be into the annulus. Exhaust from the annulus is filtered by a recirculating filtration system before being vented to the atmosphere. A Exh 87 at 19 8, TR 7 57 5. A containment isolation system provides a means for automatic and manual closure of valves in piping systems that penetrato the containment vessel wall. The containment isolation system will perf orm its f unction even in the event of a single active f ailure. A Exh 87 at 198, TR 7575. The containment is protected from high external pressure (or low internal pressure) by two independent vacuum relief lines, each of I

which is capable of relieving postulated external pressures in the annulus. A Exh 87 at 198, TR 7575.

61. The bounding DBA, with respect to containment vessel temperature and pressure retaining capability, is the postulated f ailure of the primary sodium storage tank (PSST) l

1. e

( .

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during maintenance and resulting oxidation of the sodium until all oxygen inside the containment is consumed. This l DBA assumes a series of highly unlikely events, including the postulated instantaneous f ailure of the PSST while it contains 35,000 gallons of sodium - even though the PSST will contain this volume of sodium only a few times during the plant lifetime. In addition, it was assumed that the cell in which the PSST is located would be air-filled at the -

time of tank f ailure and would be directly connected with the atmosphere of the reactor containment building (RCB).

In actuality, administrative controls will require that the cell be nitrogen filled and even if communication with the reactor containment building air environment did occur, it would be via circuitous paths of corridors and stairways, thus reducing the rate of exchange of reaction products and oxyge n. A Exh 87 at 200-20 2, TR 7577-7579.

62. The PSST failure is accommodated by 1) actuation of the containment isolation system and 2) passive accommodation of the pressure and temperature resulting from the sodium burning. The marimum containment pressure and temperature predicted for the PSST failure DBA are approximately 0.8 psig and 138' F. These predicted conditions are well within the containment vessel design pressure (10 psig) and design tempe rature (250' F). Because the oxygen initially present in the RCB is consumed in the fire, and no in-leakage is

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assumed, a subatmospheric pressure is created as the temperature approaches the initial temperature. The sub-l atmospheric internal RCB pressure is limited by the  !

operation of the vacuum relief lines. A Exh 87 at 203-204, TR 758 0-7581.

63. The ef fectiveness of the CS to deal with the radiological conseqpences of design basis accidents was analyzed using a site suitability source term (SSST) which is identical to that used for LWR's, except that the CRBRP source term contains 1 percent plutonium (none is assumed for LWR's). A Exh 87 at 205, TR 7582. The doses resulting from this source term were calculated using ICRP-30 and NUREG-0172 dose models, as appropriate, and making a number of conservative assumptions regarding fission product and plutonium inventories and the f allout and plateout of aerosols and contaminated leakage. A Exh 87 at 206-208, TR 7583-7585. The analysis considered the effects of certain engineered safety features, but did not consider features provided specifically for accidents beyond the design base. A Exh 87 at 208, TR 7 58 5.

l 64.

All SSST doses are well below the 10 C.F.R. Part 100 dose guideline values. A Exh 87 at 208, TR 7585; S W Hulman, TR 8524. The maximum dose resulting from the release to containment of any design basis accident is many times less than the corresponding SSST dose and well below the 10

l

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C.F.R. Part 100 guidelines. A Exh 87 at 208-209, TR 7 585-l 8586; S W Hulman, TR 8525.

! 65. The design features of the containment / confinement structures which mitigate the consequences of beyond design

basis accidents (by maintaining containment integrity and ~

controlling radiological releases ) include those for annulus cooling, reactor cavity venting, containment venting and pur ging, and containment cleanup. A Exh 89 at 16-17, TR ~

7778-7779; S Exh 41 at 15-16, TR 8286-8287. In addition, the containment instrumentation provides measurements of containment atmosphere pressure, atmosphere temperature, steel shell temperature, hydrogen concentration and containment radiation levels. The instrumentation, designed to remain functional following a Safe Shutdown Earthquake, provides the operator with the information necessary to minimize accident consequences. A Exh 89 at 17, TR 7779.

66. The Annulus Cooling System removes heat from the containment by blowing air through the annulus between the steel containment and the concrete confinement building. System operation is initiated by operator action from the control room. Six f ans of standard design are located outside the

' containment in the reactor service building. The annulus is partitioned to provide a spiral air flow path upward and around the containment shell to the top of the confinement l building where the heated air is discharged. All active l

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components are redundant, the equipment is located in a tornado-hardened Seismic Category I structure, and power is  !

provided by Class 1E power supplies. A Exh 89 at 23-24, TR 7785-7786; S Exh 41 at 95-96, TR 8373-8374; A Exh 91,

$ 2. 2.10.

67. The Reactor Cavity vent System is actuated by redundant rupture disks which would open only for beyond. design basis accidents. Vented materials would flow from the reactor cavity into the containment building. The flow path facilitates heat exchange between the vented gases and the cell structuress thereby reducing the thermal load in the containment atmosphere.

A Exh 89 at 18, TR 7780.

68. The Containment Vent System, initiated by operator action from the control room, allows depressurization of the containment at a rate up to 24,000 cubic feet per minute through two redundant pipes which penetrate the containment vessel wall. Each pipe has two normally closed containment isolation valves. The vent lines and containment penetrations are designed to ASME standards. A Exh 89 at 20, TR 7782.

69. The containment Cleanup System consists of an air washer, a jet venturi scrubber, and a high ef ficiency wetted fiber bed scrubber, in series. Gases and aerosols from the contain-ment building pass through these components sequentially j before the cleaned gas is released to the atmosphere. The l

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system is initiated by operator action from the control i

room. All active components are redundant, the equipment is located in a tornado-hardened Seismic Category I structure, and power is provided by Class lE power supplies. All major components are widely used in industry and prototype testing of the containment cleanup system has demonstrated high efficiencies. A Exh 89 at 20-22, TR 7782-7784; S Exh 41 at 96-99, TR 8374-8377; A Exh 91, 5 A-7.

70. The Containment Purge System allows outside air to enter the containment through the use of redundant blowers and l containment purge penetrations. System operation is initiated by operator action from the control room only after the containment has been depressurized by opc:ation of the Containment Vent System. A Exh 89 at 22-23, TR 7784-7785.

71. There is considerable margin in the pressure capability of the CRBRP containment in the case of an HCDA, In the base case, the maximum pressure is less than 24 psig with a corresponding pressure capability of about 40 psig. A Exh 89 at 46, TR 7808; S Exh 41 at 91-92, TR 8369-8370. The highest pressure calculated for any of the sensitivity studies was approximately 30 psig, as compared with pressure 28/

capability of about 40 psig. A Exh 89 at 4, TR 7766.

28/ The 40 psig value represents the pressure capability of the containment. The 10 psig design pressure of the containment (Continued) i

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The Staff's detailed safety review, which has significantly advanced from the LWA stage, ' yielded the conclusions that an early containment failure induced by HCDA energetics is physically unreasonable (highly unlikely) and that the 40 psig pressure capability had been independently verified.

S W Theofanous, TR 8520-8521; S W Butler, TR 8522-8523. The i consequences of HCDA's in CRBRP are within the 10 C.F.R.

Part 100 dose guidelines. S W Hulman, TR 8525.

Board Area of Interest 9 i l

The Staff's attention is directed to the discussion of protective action guidelines (PAGs) at pages 29-30 of the Partial Decision of February 28, 1983. The Staff is requested to address the question of whether a PAG revision for the CRBR should be made, and to '

explain its answer.

72. Protective Action Guides (PAG ' s ) are established by the U.S.

Environmental Protection Agency (EPA) for nuclear incident response planning. For the plume exposure pathway emergency for design basis accidents is based upon coincident loading of.the safe shutdown earthquake and the containment design basis accident. S W Long, TR 8486; A W Clare, TR 7731, 7733

- 7735; A W Strawbridge, TR 7731 - 7734, 7735 - 7736; see also Findings 61-62, supra. An SSE following an HCDA would be an independent event. Each is of sufficiently low proba-bility that the sequential occurrence of an HCDA followed by an SSE should not be used as the basis for evaluation of-containment pressure capability under beyond design basis conditions. S W Long, TR 8466.

, 29/ All Staff witnesses sponsoring the Staff's testimony con-cerning HCDA's agreed that their review had yielded a signi-ficant advance in Staff confidence as to the adequacy of the CRBRP containment design and that the design was adequate for the issuance of CP. TR 8528 - 8530, i

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planning zone (EPZ) the EPA has established the range of PAG's as 1 to 5 rem for whole body exposure and 5 to 25 ren for thyroid exposure. Appropriate protective actions range from sheltering at the lower end of the range to evacuation 4 - at the higher end. A Exh 94 at 5-6, TR 7983-7984; S Exh 43 ,

at 3-4, TR 8577 - 8578.

73. Both Staf f and Applicants considered the radiological dose l conseqpences for other organs (i.e., bone surf ace, red bone marrow, lung and liver) resulting f rom HCDA's at CRBRP. A Exh 94 at 6-11, TR 7984-7989; S Exh 43 at 6-11, TR 8580 -

8 58 5; A Exh 91, Table 4-3. 'ke predicted doses for whole

body and for each organ were compared to the 10 C.F.R. Part 100 guidelines or their equivalents (based upon ICRP-26 weighting f actors ) and the ratio was calculated for each

• organ. It was determined that the controlling doses for HCDA's are whole body and/or thyroid. A Exh 94 at 6-11, TR 7984-7989 ; S Exh 4 3 at 11-14, TR 8 58 5 - 8 588. While PAG's for other organs can be derived by scaling from the existing whole body or thyroid PAG's, because the whole body and thyroid doses are controlling, it is unlikely that any

specific PAG's for other organs would be either necessary or usef ul in CRBPP emergency planning. A Exh 94 at 8-11, TR 7986-7989: S Exh 43 at 6-14, TR 8 58 0 - 8 588 ; A W i

S trawbridge, TR 8023; S W Hulman, TR 8598. I f the EPA issues revised PAG's, their applicability to CRBPP will be l

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, reviewed at the OL stage of review. S Exh 43 at 15, TR 8589; S W Branagan, TR 8599; S W Perrotti, TR 8601.

Board Area of Interest 10 The Staff's testimony at Tr. 3694 anticipates the need for further research and development on measurement capabilities to achieve DOE's goals for material control and accountability at the DRP. The Staff is requested to explain whether this additional effort is currently underway or definitive'y planned for the future, and the extent to which it is critical to the effectiveness of CRBR fuel safeguards measures.

74. Research and development activities on measurement capabilities for material control and accounting are not necessary for the effectiveness of safeguards at the DRP.

Primary reliance for protection against theft of nuclear material at the DRP is placed on physical protection systems. The primary role of material control and accounting is to provide assurance that the protective systems are working effectively. Physical security and material control and accounting do not have to be considered independently. While rapid material accounting may augment safeguards measures to prevent unauthorized diversion of fuel at the DRP, the DOE commitments for DRP safeguards provide a level of protection equivalent to that embodied in current NRC regulations without the need for a rapid material accounting system. S Exh 36 at 3-4, TR 8177-8178.

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Board Area of Interest 11 In discussing the energetics of

accidents beyond design basis, the Staff l offers the statement that there will be an "isentropic expansion yield to one atmosphere" (NUREG-0968, Vol. 2, p. A. 2-l 5). The Staff is requested to discuss briefly what is the physical significance of this statement and the extent to which it contributes to any conservatism in the analyses of energy releases.

Phenomenologically, how has the Staff satisfied itself that "approximately 2550 MJ would be required to produce a slug impact kinetic energy close to the head design capability of 75 MJ" (Ibid).

75. An energetic HCDA is an accident in which sufficient thermal energy is generated that, if converted to work, it would have the potential to produce significant dynamic structural loads on the primary system boundary. A Exh 89 at 70, TR 7832. Extensive analyses by the Applicants and independent analyses and reviews by the Staff have concluded that the likely outcome of an HCDA would be non-energetic. Id. at 4, TR 7766. Nevertheless, the integrity of the reactor coolant boundary can be maintained for dynamic loads derived from an assumed energetic HCDA. For purposes of analyzing the structural capability of the CRBRP design, an HCDA was assumed to result in a mixture of molten and vaporized fuel which, when expanded isentropically to one atmosphere, would have a work potential corresponding to about 660 MJ (about 100 MJ if expanded to the free volume within the reactor vessel) . Id. at 5-6, TR 7767-7768. The pressure-volume l

1 l

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relationship for this isentropic expansion was used in the l l

Applicant s' conservative calculation of the dynamic loads imposed upon the reactor coolant boundary. Id.

The concept of an "isentropic expansion yield to one 76.

atmosphere" is used only as a ref erence point to indicate the relative potential severity resulting from disrupted core conditions. It has been widely used because it is an unambiguous and easily defined quantity. It has no physical application in that such yields cannot be realized in real systems. Since it is only a reference value, it is not

! directly used in analyzing the capability of the system to accommodate HCDA loads. Thus, it makes no real contribution to the conservatism in the analyses of energy releases.

S Exh 41 at 50, TR 8324. The Staff has determined, through a detailed analysis of a realistic expansion process, that approximately 2550 NJ would be required to produce a slug impact kinetic energy approaching the reactor vessel head

, structural capability. Id. at 30 -33, 51, TR 8302-8306, 8325.

Board Area of Interest 12 l

l NUREG-0968 contains many references to l items that are to be resolved at the OL review stage. In view of the apparently I

advanced stages of hardware design and pro-curement currently in being, the Board is concerned that said OL review (assuming a CP issues ) may require substantive changes of a castly and time consuming nature, or in the alternati ve, result in a compromise of

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performance safety. The Staf f is requested j to offer comments upon this situation and to provide whatever insights it can now offer i for avoiding such problems..

. 77. The items identified in the SER as requiring review at the OL stage, which have the potential for resulting in substan-tive changes of a costly or time consuming nature, fall into the following areas:

a) fuel design limits, methodologies, and bases; b) high temperature mechanical design limits and methodology; c) reactor vessel head structural capability; d) PRA/ reliability analysis; and e) natural circulation.

It is unlikely that any of these will result in a signi- ,

ficant impact on cost or schedule. S Exh 38 at 4-5, TR 8211-8212; S Exh 26, $$ 3.9.9, 4.2.1, 5. 2, 5.6.3; S Exh 27, App. D; A Exh 87 at 24-25, 97-99, 170-177, TR 7401-740 2, 7474-7476, 7547-7554; S W King, TR 8219. In any case, i

. confirmation or resolution of these items will not result in any compromise of saf ety. S Exh 38 at 5, TR 8212.

78. Staf f and Applicants have agreed on a course of completion for each item and Applicants have in place programs reasonably designed to address the issues. Staff and Applicants are developing a timetable to review and resolve each item which will minimize eff ects upon the final design ,

and construction schedules, and the record supports a

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conclusion that substantial changes will not be required in the CRBRP design. S Exh 38 at 5-6, TR 8212-8213; A W Strawbridge, TR 7943-7944; S W Stark, TR 8216-8217.

Board Area of Interest 13

With respect to the fuel system, the Staf f has identified certain operational f allback positions potentially available to mitigate unresolved problems (NUKE G-0968, Vol . 1, p. 4- 47, 48). The Staff is requested to discuss briefly the extent if any to which invoking such operational f allbacks might compromise the achievement of CRBR programmatic objectives.

79. The fuel system fallback positions consist of restrictions on CRBRP operation which can be imposed on the fuel system design basis, limits and methodology to resolve the Staf f's concerns, if f uture analytical and experimental data do not substantiate the Applicants' proposed design. Speci fically,

, the f allback positions are:

1) reduction of goal exposure (burnup);

2) reduction o' peak power;

3) lower operating temperature; and

4) adjustment of plant protection system trip points.

l S Exh 26 at 4-47--4-48; S Exh 39 at 3, TR 8225.

80. It is unlikely that any of the f allback positions would have l to be implemented. Applicants have committed to address i

Staf f's concerns with experimental and analytical pro gr ams . S W King, TR 8553, 8562, 8564 - 8565. Moreover, operations have been conducted at FFTF under temperature and I

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peak power conditions very close to CRBRP conditions without l

a single failure of the driver fuel rods, and results to date in programs at EBR,-II and FFTF to evaluate blanket performance confirm design performance predictions.

S Exh 39 at 6-7, TR 8228-8229; A Exh 87 at 212-213, TR 7 589-7590; S W King, TR 8255-8258.

81. Even if a f allback position were implemented, it is unlikely that CRBRP programmatic objectives would be compromised. S Exh 39 at 7, TR 8229; A Exh 07 at 212-214, TR 7 589-7 591.

The programmatic objectives of CRBRP are to demonstrate the technical performance, safety, reliability, maintainability, environmental acceptability and economic f easibility of an LMFBR operating in a utility environment in a manner that would allow extrapolation of results on various systems and components to larger plants of a similar design. One of the purposes of a demonstration plant is to identify problem areas which are not anticipated from existing experimental data or from existing operating plants. If the problem areas are identified prior to or during CRBRP operation, design modifications can be made in the plant and shown to be ef fective prior to entering into full-scale deployment of 30/ The Applicants' predictions of favorable fuel performance l with stainless steel cladding are based on extensive experience at EBR-II, FFTF, and foreign reactors, and conservative analytical methods which have been qualified l against the actual experience. A W Schwallie, TR 7609-7611, 7619-7621; A W S trawbridge, TR 7619; A Exh 64 at 4.2-316. ,

l 9 l l

- F ,l the reactor concept and plant design. In the area of fuel pe rfo rmance, the lessons learned in the initial operation of CRBRP can easily be f actored into reload designs and will create the technology data base to be used for the follow-on plant designs. Fur thermore , the success of the CRBRP demon-l stration will be based on the total technology developed for overall plant system performance and not just on fuel performance. A Exh 87 at 212-214, TR 7589-7591; S Exh 39 at 3-5, TR 8225-8227 ; A W Schwallie, TR 7607.

Board Area of Interest 14

Operation with leaking fuel pins could conceivably of fer the opportunity for these pins to " inhale" some amount of sodium whenever the reactor is shut down. Should this occur, subsequent return to operation at power might then result in a significant increase in pellet-to-cladding gap conductance with an attendant of f-normal performance of the fuel. The Staf f is requested to comment upon whether it sees this as a problem requiring resolution and the reasons for its answer.

82. Both the Staf f and Applicants have considered the eff ects of leaking f uel pins on of f-normal performance of the f uel. S Exh 26 at 4-20 and 4-21; A Exh 72, .I 15.4.1.1. S odium in contact with fuel inside a pin may result in increased pellet-to-clad gap conductance but would not adversely affect fuel performance during steady state or transient conditions. Id.; S Exh 40 at 2-3, TR 8249-8250; A Exh 87 at

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215-216, TR 7 592-7593 ; A W Schwallie, TR 7 612-7 614; S W King, TR 8258-8260.

83. The Staf f is concerned that continued operation with f ailed fuel rods might cause local swelling with the potential for ,

flow restriction. S Exh 26 at 4-20 and 4-21; S Exh 40 at 3, TR 8250. Staf f and Applicants have agreed to operational restrictions on CRBRP which would require removal of the fuel assembly containing the failed pins at the next reactor -

shutdown or upon exceeding a predetermined delayed neutron 31/

detector signal. S Exh 40 at 3, TR 8250; A Exh 87 at 217, TR 7594. Those restrictions will be reviewed based on the results of tests being conducted at EBR-II to investigate the limits of steady-state and transient operation with f ailed fuel rods. S Exh 40 at 3, TR 8250; A Exh 87 at 216, TR 7 593 ; S W King, TR 8261.

t Board Area of Interest 15 The Applicants have proposed a reliability assurance program that focuses primarily on plant protective systems. The Board requests Applicants to address the question of whether said program will (or ought to) take account of findings derived from the CRBR quality assurance program, and if so, describe the administrative mechanism envisaged to accomplish this.

31/

--- Applicants anticipate operation with less than 1% f ailed fuel. A W Clare, TR 7637.

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84. The reliability assurance program (RAP) is not a formal part of the QA program. However, certain QA program activities are applied to fulfill requirements of the RAP. For instance, the RAP failure evaluation activities are considered part of the QA program. Quality assurance personnel participate in f ailure evaluations and determination of what corrective action is appropriate.

Formal quality assurance program procedural and documentation requirements assure that proper failure evaluations are performed. The effectiveness of quality assurance programs in other nuclear power plants is inherently included in the quantitative and qualitative failure rate assessments. Where specific CRBRP quality assurance program inspection or control activities prevent specific f ailure modes identified in the RAP, these data are considered in the RAP assessments. These assessments are made as part of the Component Level Evaluations and System Level Evaluations. As part of the quality assurance program for the CRBRP as a whole, surveillance and auditing of RAP activities and documentation will be performed to assure that RAP procedures are being met and that overall RAP requirements are satis fied. A Exh 87 at 168-169, TR 7545-7546.

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Board Area of Interest 16 The SER discusses the impact of aerosol behavior on containment shell cooling. The Staf f is requested to comment on whether changing concrete aggregate f rom calcitic to dolomitic limestone could significantly alter the behavior of the aerosols, and explain the basis for the answer.

85. The results of sodium-concrete testing with both calcitic and dolomitic limestone aggregate demonstrated no detectable difference in sodium-concrete reactions and depth of penetration of concrete. A Exh 89 at 41-42, TR 7803-7804:

S Exh 41 at 86, TR 8364; A W Strawbridge, TR 7951; SW Swanson, TR 8543 ; S Exh 26 at A. 4. 4. - A. 4.6, A. 4. 25 -

A.4.26. The type of aggregate would not af fect the sodium aerosol behavior in the reactor containment building (RCB). No significant differences in the physical characteristics of the reaction products were observed in tests of the two types of limestone aggregate concrete.

Virtually all of the sodium aerosols in the RCB would be generated in the RCB by burning of the sodium vapor being vented from the reactor cavity, and would be unaffected by the reaction occurring in the reactor cavity. A Exh 89 at 40-42, TR 7802-7804; S Exh 41 at 85-87, TR 8363-8365; A S Strawbridge, TR 7952-7953. l l

Board Area of Interest 17 What is the status of the Staf f's review I of, and what is the Staf f's position with  ;

respect to, "The Eight Areas of Concern" '

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listed in Section I, Table II of NUREG/CR- i 32247 l

86. Each area of Staf f concern has been resolved to the Staf f's satisfaction. S Exh 41 at 51, TR 8325.

87. Area 1. The Staff was concerned that a TOP event could become prompt-critical in such a way that internal fuel 3 motion would be a key f actor in the energetics

determination. However, calculations demonstrated that a TOP HCDA would not lead to a prompt-critical excursion.

Even for the worst-case situation, a margin against the autocatalytic regime was found. S Exh 41 at 52-5 3, TR 83 26-8327; A Exh 89 at 145-146, TR 7907-7908.

88. Area 2. The LOF-d-TOP concern was similar to that for the TOP, supra, i.e., autocatalysis. Applicants' analysis ,

using experimentally-based modeling of lead channel fuel motion, demonstrated no potential for LOF-d-TOP. The Staf f 's independent assessment, even though conservatively structured, also indicated the unattainability of LOF-d-TOP unless unrealistic assumptions were made or the plenum fission gas driven fuel compaction mechanism were involved . S Exh 41 at 53-54, TR 83 27-83 28; A Exh 89 at 146-l 147, TR 7908-7909; see Area 3, Finding 89, infra.

89 . Area 3. The Staff was concerned that autocatalysis might result from the plenum fission gas acting on the fuel column j to force rapid f uel compaction from above. The Applicants' best-estimate results f rom analyses, including

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experimentally based modeling of fuel disruption and other phenomena, predicted no potential autocatalytic 1

conditions. Staf f's more conservative independent assessment also showed the lack of a direct autocatalytic tendency from pressure-driven compaction but indicated a general tendency for development of the LOF-d-TOP situation

( Area 2), particularly near the end of the burnup cycle. If f urther analysis confirms this concern, Applicants have committed to make a f easible design change to prevent the fission gas from rapidly acting on the fuel in a compactive manner. S Exh 41 at 39-40, TR 8313-8314; A Exh 89 at 107-1 08, 148-149, TR 7869-787 0, 7910-7911: S W Theofanous, TR 8457: A W Fauske, TR 79 68.

90. Area 4. The Staf f was concerned that steel blockages could form to prevent fuel removal through normal axial blanket flow channels during the early phase of the LOF event. Both Staf f's and Applicants' analyses indicate complete, core-wide steel plugging will not occur prior to substantial f uel disruption in the hottest subassemblies. Because of co-melting of fuel and cladding in a large part of the core, a significant number of fuel removal paths will be available through the axial blankets. S Exh 41 at 55-56, TR 8329-

8330; A Exh 89 at 149-150, TR 7911-7912.

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91. Area 5. The Staf f was concerned with the basis for maintaining continuous subcriticality in the early melt-out phase. Detailed analyses of the overall accident sequence, separate effects analyses, and fuel f reezing and plugging data demonstrate that, while mild recriticality events cannot be ruled out during the melt-out/ annular pool phase, suf ficient fuel removal to assure permanent subcriticality can occur prior to melt-out of the inner blanket and formation of a large-scale homogeneous pool. S Exh 41 at 56-57, TR 8330-8331; A Exh 89 at 150-152, TR 7912-7914.

92. Area 6. The Staf f was concerned with the degree of suberiticality required to prevent pool recriticality f rom thermal and fluid dynamic upset conditions. Permanent removal of about 40% of the original fuel inventory is suf ficient to completely eliminate the recriticality potential. Through analyses and special experiments, it was determined that the dominance of fuel removal in the accident sequence eff ectively eliminated the concern. S Exh 41 at 57-58, TR 8331-8332; A Exh 89 at 150-152, TR 7912-7914.

93. Area 7. The Staf f was concerned with the ability of the f uel to discharge upward f rom a boiling pool at decay power, with an associated concern of sodium re-entry. Analyses demonstrated that during the time when fuel loss is occurring, sodium re-entry is precluded as a source of

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pressure compaction of the annular pool material. Analyses of the disruption process indicate that the upward discharge from local openings in a sealed, boiled-up core, which leads to this area of concern, is irrelevant. S Exh 41 at 59, TR 8333 ; A Exh 89 at 150-152, TR 7912-7914.

94. Area 8. The Staff was concerned with the force required to produce a mechanically induced relief path via upper internal structure displacements. Analyses and scale model -

tests were used to estimate the forces required to cause significant displacement of the upper internal structure to be approximately 91 atmospheres. Such a high pressure would be of extremely low likelihood, even under HCDA conditions. A Exh 89 at 153-154, TR 7915-7916; S Exh 41 at 60, TR 8334; S W Butler, TR 8460; S W C. Bell, TR 8461-8462.

OTHER MATTERS

95. During the evidentiary hearings the Board inqpired and heard from Applicants' and Staf f's witnesses concerning several points in the July 8, 1983 Limited Appearance statement filed by the Natural Resources Defense Council and the Sierra Club  ! [hereinaf ter "NRDC's Statement"]. B Exh 1 25. That Statement reiterated the substance of an argument previously presented to the Board at closing argument during the LWA proceedings.

32/ Hereinafter "NRDC".

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96. The argument advanced by NRDC at the LNA proceedings was not i

the product of testimony sponsored by a well qualified witness, but surf aced for the first time af ter the evidentiary presentations, by way of closing argument by NRDC's technical representative, Dr. Cochran. TR 6626-33, 6639-43, 6562-63. NRDC argued that: 1) the Staff has i

adopted a reliability objective of 10-6 per reactor year that HCDA's would not result in consequences exceeding 10 C.F.R. Part 100 guideline values; 2) the Staff's analyses in Appendix J of the FES Supplement show that HCDA's have a probability of more than 10 4 per year; 3) the consequences of HCDA's will exceed the Part 100 guideline values; and accordingly, 4) HCDA's must be DBA's. Iji.

97. Neither Staf f nor Applicants have adopted or relied upon a single-valued quantitative reliability objective, nor have they used quantitative reliability analyses as the basis for a decision on whether an HCDA should be a DBA. PID 33/

~~~

It should be noted that NRDC's reference to the Standard Review Plan is incorrect. B Exh 125 at 7, TR 7659. The l quoted portion of the Standard Review Plan does not j

establish a generally applicable numerical reliability goal, much less a goal for CRBRP. The Standard Review Plan .

addresses offsite hazards ( e. g. , nearby industrial f acili ties ) . See S Exh 6 at 1; A W Clare, TR 7742. There is no inconsistency here with the Appeal Board's decision in Florida Power and Light Company (St. Lucie Nuclear Power Plant, Unit No. 2 ) , ALAB-60 3, 12 NRC 30 (1980). The record contains no evidence that there are important sequences which remain unidentified, and which would suggest a need for additional inquiry and scrutiny. Moreover, NRDC has i never suggested any important sequences of this nature, or '

(Continued)

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at 73; A W Clare, TR 7741-7750.

98. NRDC misapplied the Staf f's analyses in Appendix J of the FES Supplement. The referenced 10-4 probability of HCDA 34 initiation in Appendix J--/ was not proferred as a reliability analysis for comparison against a single-value reliability objective, and thus, to demonstrate whether an HCDA should be a DBA. S Exh 17 at 8, 14, TR 57 55, 5761; A Exh 46 at 12, 21, TR 5388, 5397; see Applicants' Proposed ^

PID at F-43. Rather, it was developed as a conservative estimate of accident initiation probability, for purposes of addressing the environmental risk of accidents under the Commission's applicable Policy Statement (45 Fed. Reg. 40101 (June 13, 19 80) ) . ' S Exh 24 at J-1. In short, the probability values relied upon by Intervenors at the LWA stage are simply not applicable.

99. NRDC's estimates of the consequences of HCDA's are without support in the record. NRDC's " first order approximations" or "adj ustments", for meteorological factors, to any which might fall within the ambit of the Commission's TMI-l decision. Metropolitan Edison Company (Three Mile Island Nuclear Station Unit No. 1), CL I-8 0-16, 11 NRC 674 (1980).

34/

-~~ The 10-4 value is the Staf f's conservative estimate of the probability of HCDA initiation, and it does not include the conditional probability that, given an HCDA, it will result in doses exceeding the Part 100 dose guidelines. S Eih 24 a t J-8 -- J-ll .

35/ The Staf f's use of realistic methodology for analysis of HCDA's and conservative ' methodology for DBA's is consistent (Con tinued) 1

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Applicants' and Staf f's dose calculations, are simply incorrect. See NRC Staf f Proposed Findings 104 and 105, at 7 2-73, January 24, 19 83. NRDC's additional " adjustments" for such f actors as plutonium isotopic composition, uncertainty in dose guideline values, etc., are similarly ,

incorrect or insignificant in their effect. See Applicants '

Propos ed PID at F-4 3 -- F-53. No significant new information in this respect has been presented in NRDC's -

Statement to warrant further inquiry or to af fect the Board's previous findings.

100. NRDC's Statement attempts to extend its LWA argument with the following supplemental points: 1) the Staf f's HCDA dose calculations for thyroid should be based upon inf ants rather than adults and judged against reduced dose guideline values;

2) Applicants' PRA results now show that the probability of CDA initiation is realistically 10-4 per reactor year or less; and 3) the site suitability source term dose calcula-tions for CRBRP must include releases through the beyond design basis vent / purge system, since both a Savannah River j with longstanding Staf f practice. S W Hulman, TR 8505 -

8506. NRDC's Statement provided no sound reason for the Staf f to deviate from this practice for CRBRP. S W Hulman, TR 8506. Moreover, the arguments presented on page 12 of NRDC's S tatement, TR 7764, concerning- meteorology are based on an incorrect premise. NRDC's Statement urges the use of one percent meteorology, when the Staf f's practice is to employ five percent meteorology, even for conservative DBA

a nalys'es . S W Hulman, TR 8509.

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production reactor and the Fort St. Vrain reactor included such releases.

Thyroid Doses 101. NRDC's Statement at 2-4, TR 7654-7656, urges that for purposes of judging the radiological consequences of the Staff's HCDA dose calculations: 1) the thyroid dose calculations should be based upon infants rather than ,

adults; and 2) the 300 rem thyroid dose guideline value of 10 C.F.R. 3 100. ll (a) , footnote 2, should be a) reduced to account for exposure to infants and recent data from the Marshall Islands; and b) further reduced by a factor of 2 at the CP stage to account for uncertainty.

102. As NRDC concedes, changing the 300 rem dose guideline value to account for infant exposure  !

or the Marshall Islands data would constitute a challenge to the Commission's 26/ While the NRC Staff agrees that the infant thyroid is more radiosensitive than the adult, it disagrees that the infant should be used as the basis for the dose calculations under consideration here. S W Hulman, TR 8527.

J7/ NRDC's Statement merely enclosed a single Table from a Brookhaven National Laboratory Report, and stated that the

" data speak for themselves." B Exh 125 at 4, TR 7656. The full report, however, indicates that the data do not, in fact, speak for themselves. The report explicitly states that: a) the thyroid dose estimates are subject to considerable uncertainty and may, at least in some cases, be considerably higher than estimated; b) none of the unexposed groups is a completely valid control group; c) because of

the small numbers of people involved and uncertainties of f the doses received, the data do not lend themselves to dose-response analysis; and d) the absorbed dose estimates to the (Continued)

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regulations. B Exh 125 at 4, TR 7656; 10 C.P.R. S 2.758; S W Hulman, TR 8505. As to basing dose calculations on infants, 10 C.F.R. Part 100 refers to TID-14844 for guidance concerning such calculations, and that guidance, which the Staff has consistently followed, is based upon adults. AW Strawbridge, TR 7715; S W Hulman, TR 8504. Further, NRDC's argument mistakes the purpose of the dose guideline values.

They are not intended as acceptable limits for emergency doses, but are reference values for use in the evaluation of reactor sites. S W Hulman, TR 8502 - 8504; 10 C.F.R. S 100.ll(a), footnote 2; A W Strawbridge, TR 7716. As for the use of reduced (by a factor of two) dose guideline values as the basis for judging HCDA doses, it should be emphasized that the 10 C.F.R. Part 100 guidelines were not developed or intended for accidents beyond the design basis, and they were applied to HCDA dose calculations by the NRC Staff with the express stipulation that they be used for realistic assessment of HCDA's, rather than as limits. S Exh 27, App A at A.1-5; S W Hulman, TR 8505 - 8506. In addition, the uncertainty in the CRBRP meteorological data and design are now sufficiently low that the Staff sees no need to apply a reduction factor to account for uncertainty. S W Hulman, TR exposed Marshallese are approximate, and the uncertainties in many of the parameters involved in obtaining them make it impossible to state their statistical reliability. AW Strawbridge, TR 7717-7719; A Exh 96.

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8501 - 8502. In light of this and the fact that the dose guideline values are not used for assessing site suitability under HCDA conditions, there is no basis for crediting NRDC's argument that the f actor of 2 reduction recommended by the Staf f for site suitability evaluation at the CP stage should be applied to HCDA doses.

HCDA Initiation Probability 103. NRDC's Statement urges that a probability of lO~4/yr. for HCDA initiation from loss of offsite power is realistic based upon Applicants' Phase I PRA study. This ignores the fact that the study is not complete, and the following 38 express statement in the abstract of the studyr- /

Significant conclusions and use of portions of this report should be avoided until completion of the CRBRP PRA. This report should only be used as a starting point for f urther refinement and investigation during Phase II .

2 The record shows that: a) a previous PRA study (CRBRP-1) was completed in 1976; b) the present PRA was begun in 1982 and the Phase I work was based upon the design as of February,1982; c) the magnitude and scope of the PRA increased significantly during the NRC Staff / Applicants interaction; d) as a result, the work on Phase I was stopped and rebid competitively, with award to another contractor; 38/ B Exh 125, Attachment 1, TR 7 674; A W Clare, TR 7743-7744.

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and e) the Phase I report only reflects the progress of the Phase I work to the point where it had stopped. A Exh 87 at 175-176, TR 7552-7553; A W Clare, TR 7743-7745.

104. The record also shows that the 10-4 HCDA initiation probability for loss of power was: a) based upon average nuclear plant offsite power failure data, rather than the actual experience on the grid and switchyards feeding CRBRP; b) based upon the CRBRP design before the emergency power systems were upgraded; c) based upon conservative failure criteria for safety systems; and d) did not include consideration of recovery. A W Clare, TR 7745-7748. The record thus shows that the probability value advanced by NRDC's Statement is not reliable information, and that it would have no effect on the Board's conclusions.

Savannah River and Fort St. Vrain Reactors 105. NRDC's Statement at 9-17, TR 7661 - 7669, argues again that releases from the beyond design basis vent / purge system must be included in the site suitability dose calculations. This argument mistakes the nature, function and purpose of the vent / purge system. The vent / purge system would play no meaningful role in the context of CRBRP design basis accidents and site suitability evaluation. A W Strawbridge, TR 7722-7723; see Applicants' Proposed PID at F-51 --

F-53. The only new argument advanced by NRDC is that a Savannah River production reactor does include, and the Fort

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~l St. Vrain Reactor may include, releases from a vent / purge system in their site suitability calculations. B Exh 125 at 9-13, TR 7661-7 66 5. Although the Savannah River reactor is not licensed by NRC, and its logical relationship to CRBRP is doubtf ul, the record does show that a) neither the Savannah River nor Fort St. Vrain reactor has installed the functional counterpart of the CRBRP beyond design bases vent / purge system; b) both reactors have installed, as part of their confinement systems, filtration systems which are the functional counterpart of the CRBRP design basis annulus filtration system; and c) the CRBPP site suitability analyses do include releases from the annulus filtration system. A W S trawbridge, TR7723-7725.

Emergency Planning at Nearby Facilities 106. The Board's February 28, 1983 Partial Initial Decision, Finding 52 at F-42, noted interest in emergency responses to CRBRP accidents at the three major DOE Oak Ridge f acilities

-- the Oak Ridge Gaseous Diff usion Plant (ORGDP ) , the Oak Ridge National Laboratory Plant (ORNL) , and the Y 12 Plant. A Exh 94 at 12, TR 7990.

107. ORGDP is located 2. 5 to 3. 5 miles from CRBRP and has a total plant population of about 4300, with about 90 workers on each of f-shif t. A Exh 94 at 12-13, TR 7990-7991. ORNL is located 4.5 to 5.5 miles from CRBRP and has a total plant population of about 4200, with from 50-160 workers on each l 1 l

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of the of f-shif ts. A Exh 94 at 13, TR 7991. The Y-12 Plant is located about 9-11 miles from CRBRP. The present total population is about 8000. Evening shif ts have about 1000 workers, midnight shifts about 450, and holiday and weekend

shifts about 125. A Exh 94 at 13, TR 7991.

100. Each of the three nearby f acilities has a long-standing emergency program in ef fect which includes three basic elements. First, each f acility has an existing

organizational structure, with designated emergency directors and emergency squads, of 10-15 shif t workers.

These emergency personnel are supported by specialized personnel in such areas as environmental monitoring, medical, transportation, health physics, etc. Emergency staf f are on duty 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, seven. days per week.

Support personnel are or. call during of f-shifts. Second,

emergency plans and procedures are in effect which identify credible emergencies, specify actions to be taken, identify

emergency procedures and instruct personnel in such areas as personnel evacuation, emergency notification, medical re quirement s, etc. A EMI 94 at 14-15, TR 7992-7993. Third, each of the f acilities has extensive emergency resources to assist each other and the public. These resources are supplemented by assistance from other DOE facilities and EPA's Eastern Environmental Radiation Facility in i

Montgomery, Alabama. Assistance to the public is coordinated

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by the DOE Regional Radiological Response Center, located approximately 12 miles from CRBRP. A Exh 94 at 14-15, TR 7992-7993.

109. There are no inherent constraints at any of the three facilities to prevent rapid sheltering or evacuation in the event of an accident at CRBFP. A Exh 94 at 16-17, TR 7994-

7995. At ORGDP it would be desirable to maintain a small operating and support staf f to avoid shutting down the diff usion cascade unless clearly necessary. Radiation protection for thess' few personnel would be implemented. A Exh 94 at 16-17, TR 7994-7995. Because the Y-12 Plant is located approximately 9-11 miles from CRBRP, the resulting lower doses at Y-12 make evacuation unlikely. If evacuation were necessary, a small security staf f would be maintained. The lower doses at Y-12 would allow for implementation of suitable protection measures for security personnel. A Exh 94 at 17, TR 7995.

110. Special measures will be undertaken at each of these three facilities to establish procedures for dealing with CRBRP accidents, including establishing prompt notification procedures, conducting drills and exercises simulating CRBRP accidents, and providing for the needs of essential personnel. A Exh 94 at 17-18, TR 7995-7996.

111. Because of the existing emergency programs in effect at the nearby facilities, these f acilities are compatible with

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CRBRP emergency plartning. There are no impediments to emergency planning which cannot be dealt with eff ectively.

A Exh 94 at 18, TR 7996.

PRA Results 112. At a Conference with Counsel on June 29, 1983, the Board noted its interest in the issue of whether the results of the presently ongoing PRA might change the conclusions reached at the CP proceeding. TR 73 40-7441.

113. The present CRBRP PRA was begun in early 1982 and is scheduled for completion in late 1984. A Exh 87 at 175-17 6, TR 7552-7 553. The Applicants have committed to factor the PRA results into the design and operation of CRBRP. There is reasonable assurance that this can be achieved and that the conclusions reached at the CP stage will not be adversely af fected by the results of the PRA, based upon the following factors: 1) comprehensive consideration of fundamental engineering principles has identified the four functions necessary to prevent accidents beyond the DBA spectrum; 2) application of proven mechanistic engineering approaches to the CRBRP design f eatures provides high confidence that the four functions will be accomplished with ,

high reliability; 3) giver; l) and 2) above, any design i modifications indicated by the PRA are expected to be l

l relatively minor, leaving unchanged the fundamental designs l of the design features which address the four functions; 4) l l

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the results of the PRA will be available in late 1984, when construction will have proceeded only to the point that building foundations have been placed, and 5) operational changes can be made up to, or even after initial plant operations begin. A Exh 87 at 176-177, TR 7553-7554; S Exh 32 at 43-50, TR 8078-8085.

Additional Matters 114. The Applicants' application contains the general information required by 10 C.F.R. S 50.33 (a)-(e) . A Exh 86.

115. Classified information, i.e., the CRBRP Physical Security Plan, Safeguards Contingency Plan and Security Personnel Training and Qualification Plan, will be submitted separately from unclassified information, and adequately protected from disclosure. A Exh 71 at 13.7-1--13.7-10.

The Preliminary Safety Analysis Report (PSAR) adequately describes the Applicants' plans to implement physical security measures for protection against sabotage of the facility and theft of special nuclear material. A Exh 71, 3 13.7; S Exh 26, S 20.

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39/ Very little equipment will have been installed by that time (much of the equipment comprising the design features will not be installed until years after the PRA is completed).

Design changes could be readily implemented prior to installation, and many could even be accomplished after installation. A Exh 87 at 176-77, TR 7553-7554.

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116. The PSAR adequately describes and assesses the saf ety of the site on which the CRBRP is to be located. See, A Exh 59, j $ 1. 2; S Exh 26, $ 1.2 (general plant description); A Exh 60, I 2 ; S Exh 26, {2 (site characteristics); A Exh 72,

$ 15A; S Exh 26, { 15 (radiological source term for assessment of site suitability).

117. The PSAR adeqpately describes the f acility, including design and operating characteristics, unusual or novel design features, and principal ,saf ety considerations. A Exh 59,

(( l. 2 and 1.5 ; S Exh 26, { l. 2.

118. The PSAR adequately describes the preliminary design of the CRBBP including: i) the principal design criteria ( A Exh 62, i 3.1; S Exh 26, I 3.1); ii) the design bases and relation of the design bases to the principal design criteria (A Exh 62, i 3: A Exh 72, $ 15; S Exh 26, (( 3, 15, 15A, and 15B); iii) Information relative to materials of constru ction, general arrangement, and approximate dimensions suf ficient to provide reasonable assurance that the final design will conform to the design bases with

adequate margin for safety. A Exhs 64-71, (( 4-12; S Exh 26, $$ 4-12.

119. The PSAR contains an adequate preliminary analysis and evaluation of structures, systems and components important to assessing public health and safety, including i) margins of safety during normal and transient conditions (A Exhs 71 l

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and 72, $$ 15 and 16; S Exh 26, $$ 15 and 16); and 11) adequacy of structures, systems and components provided for the prevention of accidents and mitigation of conseqpences. A Exhs 64-6 5, 5 4; A Exhs 66-67, 5 5; A Exhs 67, $ 6 ; A Exh 67-68, $ 7; A Exh 70, $ 11; A Exhs 71, i 12; A Exh 71-72, i 15; S Exh 26, { { 6, 7, 11, 12, 15,15A and 15B.

120. The PSAR adequately identifies and justifies the selection of variables, conditions or other items which are probable subjects of technical specifications. A Exh 72, $ 16; S Exh 26, $ 16.

121. The PSAR adeqpately describes the preliminary plan for the Applicants' organization, training of personnel, and conduct of operations. A Exh 71, i 13; S Exh 26, i 13.

122. The PSAR adequately describes the Applicants' quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the CRBRP. A Exh 73, $ 17; S Exh 26, $ 17.

123. The PSAR identifies those structures, systems or components requiring research and development (R&D) to confirm the adequacy of the design and describes the R&D program and the schedule for the R&D program, showing that safety questions will be resolved at or before completion of construction. A Exh 59, $ 1.5; A Exh 63, $$ 3.8A, 3.8B, and 3.9 ; A Exhs 64-65, $ 4. 2.1. 5 ; A Exh 67, $ 5. 5. 3.1. 5.1; A Exh 71,

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I 14.1. 4; A Exh 74, App. A . 3. 4; S Exh. 26, (( 3.8.A.1, 3.9.9, 14.1.4 ; S Exh 27, App. A.3 ; S Exh 29, $ $ 3.9.9. 2. 3, 4.2.1.

124. The Applicants are technically qualified to engage in the

. proposed activities. A Exh 59, i 1. 4 ; S Exh 26, { l.2, 13, 17.2, 23 ; A Exh 86.

125. The PSAR adequately describes Applicants' preliminary plans for coping with 62ergenciec and the preliminary estimates of evacuation times for the ten mile EPZ are adequate. A Exh 71, { 13. 3, App 13.3A; S Exh 26, i 13.3; A W Bowman, TR 8011-8018.

126. The earliest and latest construction completion dates for the CRBRP are January 1989 and January 1992, res pectively.

A Exh 86 at 11.

127. Applicants will comply with the applicable portions of NUREG-0718, " Licensing Requirements for Pending Applications for Construction Permits and Manuf acturing Licenses." A E xh 7 4, App. H: A Exh 82, O/A QCS 4 21.14 a nd 4 21. 46 ; S Exh 26, { { 4. 4, 5. 3, 5. 6 . 2. 2, 5. 6 . 3, 6. 2, 6. 3, 7.4.2.2, 7.5. 2.8, 7.5. 2.11, 8. 3.3. 5, 9.8, 11.5, 12.1, 12.2.4, 13.2, 13.5.2, 13.6.1, 13.6.2, 17, 18; S Exh 27, App. A , D, T abl e 1.1.

128. Applicants have adequately described the preliminary design j of equipment to maintain control over radioactive materials 40/ Cf . 10 C.F .R. I 50.34(f).

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. in gaseous or liquid ef fluents during normal operation and identified the design objectives for keeping levels of radioactive materials in ef fluents as low as reasonably achievable. A Eih 7 0, { { 11. 2 and 11. 3 ; S Eih 26, { 11.4.1 and 11. 4. 2.

129. ApplicahEs' have adequately described the preliminary design of equipment installed to control radioactive effluents.

Id.

130. Applicants have adequately estimated the expected annual release to unrestricted areas of the principal radionuclides in liquid ef fluents ( A Exh 70, Table 11. 2-6 ; S Exh 26,

{ 11.4) and in gaseous effluents ( A Eih 7 0, Tabl e 11. 3-10 ; S Exh 26, $ 11.4) produced during normal operations.

131. Applicants have provided an adequate general description of provisions for disposal of solid wastes containing radioactive material. A Eih 7 0, { 11. 5 ; S Eth 26, { 11. 4.

132. The Staf f has identified those unresolved generic safety issues which have applicability to the CRBRP. S Exh 27, App. B. For each of the applicable issues, the Staf f has described how the problem has been or will be resolved for the CRB PP , including appropriate investigative programs and their duration, the need for interim measures, and alternative courses of action if the planned resolution does not accomplish the desired result. Id; 5 W Stark, TR 8559 -

8560.

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in gaseous or liquid ef fluents during normal operation and identified the design objectives for keeping levels of radioactive materials in ef fluents -as low as reasonably achievable. A Exh 7 0, { { 11. 2 a nd 11. 3 ; S Exh 26, (( 11. 4.1 and 11. 4. 2.

129. Applicants have adequately described the preliminary design of equipment installed to control radioactive effluents.

JB1-130. Applicants have adequately estimated the expected annual release to unrestricted areas of the principal radionuclides in liquid ef fluents (A Exh 70, Table 11. 2-6 ; S Exh 26,

{ 11.4) and in gaseous effluents ( A Exh 7 0, Table 11.3-10; S Exh 26, { 11.4) produced during normal operations.

131. Applicants have provided an adequate general description of provisions for disposal of solid wastes containing radioactive material. A Exh 7 0, $ 11. 5 ; S Exh 26, { 11. 4.

132. The Staf f has identified those unresolved generic saf ety issues which have applicability to the CRBRP. S Exh 27, App. B. For each of the applicable issues, the Staf f has described how the problem has been or will be resolved for the CRBRP, including appropriate investigative programs and their duration, the need for. interim measures, and alternative courses of action if the planned resolution does not accomplish the desired result. Id; S W Stark, TR 8559 -

8560.

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133. The ACRS report was introduced into the record. S Exh 31.

134. The CRBRP will permit the conduct of widespread and diverse research and development. A Exh 86; S Exhs 23 and 24.

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VI. CONCLUSIONS OF LAW Based upon our review of the entire record in this pro-ceeding and the foregoing Findings of Fact, and pursuant to 10 C.F.R. { 2.104(b) and $ 50. 35(a), the Board concludes as follows:

A. The application and the record of the proceeding contain suf ficient information, and the review of the application by the Staff has been adequate to support the foregoing findings and the following -

conclusions.

B. In accordance with the provisions of 10 C.F.R.

$ 50. 35(a ):

(1) The Applicants have described the proposed design of the f acility, including, but not limited to, the principal architectural and engineering criteria for the design, and have identified the major features or components incorporated therein for the protection of the health and safety of the public.

( 2) Such f urther technical or design information as may be required to complete the saf ety ,

1 analysis, and which can reasonably be left for later consideration, will be supplied in the final safety analysis report.

(3) Saf ety f eatures or components,. if any, which 1

require research and development have been

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described by the Applicants, and the Applicants have identi fied, and there will be condu ct ed, a research and development program reasonably designed to resolve any saf ety questions associated with such f eatures .or components .

(4) on the basis of the foregoing, there is

. reasonable assurance that a) such safety questions will be satisf actorily resolved at or before the latest date stated in the application for completion of the proposed f acility; and b) taking into consideration the site criteria contained in 10 C.F.R. Part 100, the proposed f acility can be constructed and operated at the proposed location without undue risk to the health and safety of the p ublic.

C. The Applicants are technically qualified to design and construct the proposed f acility. 10 C.F . R.

$ 50.35(a)(7). The Applicants constitute an electric utility and a financial review is not required. 10 C.F.R. $$ 2.104(b)(1)(iv) and (b)( 2) ;

9 10 C.F.R. $ 5 0. 2 ( x) and 5 0. 3 3 (f ) ; 47 F ed. Re g.

1 13750 (March 31, 1982).

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D. The issuance of a permit for construction of the f acility would not be inimical to the common def ense and security or to the health and safety of the public.

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APPENDIX A BOARD AREAS OF INTEREST 1/

1. In its safety Goal Development. Program announce-ment (48 Fed. Je R . 10772, March 14, 1983) the Commission stated

that during the 90-day period (ending June 8, 1983) for public comment on the proposed evaluation plan "it 1

s expected that preliminary information on new radiological source terms will become available. . . " (Id., at 10778). The Staff is requested to advise whether that information will be evaluated for any impact on this proceeding, and the reason for its answer. .

2.

As regards fuel performance, to date the use of the term " failed fuel" has not consistently permitted delinea-tion of the various failure modes that might have been alluded to (e.g., clad perforation, fission product leakage, clad i

bulging or rupture, melting of fuel pellets, etc.).. The Appli-cants are requested to summarize the anticipated performance of the~CRBR fuel associated with normal operation and accidental' -

transients, describe various failure modes that must be dealt with, identify any operational limits (e.g., maximum linear

, heat generation rates, maximum cladding hot spot temperatures, etc.) to be imposed, and to review the basis for confidence (e.g., supportive evidence) that the proposed fuel behavior characteristics will be realized.

3. Avoidance of primary coolant pipe rupture seems to depend in part upon the fact that coolant temperature is well below its boiling temperature and that coolant pressure is near atmospheric pressure (5 10 atmos.). Applicants are requested to present a technical summary of how these coolant characteristics will result in a reduced likelihood of pipe rupture in piping designed for CRBR use.

4. Applicants are requested to explain.how the CRBR will be configured to assure that convective circulation of the sodium coolant will be available to prevent fuel damage, if needed. This explanation should reference any supportive experimental or operational evidence. The Staff is requested to advise the Board whether it accepts convective circulation as a viable' mechanism for fuel protection, and the reason

, for its answer.

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~ Evidentiary' Hearing, dated May 24, 1983.. q l

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5. In the area of quality, the Applicants are

requested to explain whether (and/or how). differing functional levels of effort will be applied, depending upon whether a component or system is necessary for safety, important to safety, or not-safety related. The divisions of authority and i

functional responsibilities for quality assurance and quality -

control amongst the various contractors and the Applicants should be discussed with emphasis on how the management of

the various CRBR contractor fabrication and construction efforts l will be coordinated to assure the minimizing of QA and QC oversights, especially where interfacing is involved. Appli-

cants are also requested to describe what efforts will be undertaken to insure that accurate as-built plans and speci-4 fications will be available when needed, if the CRBR is constructed.

6. The SER discussion of quality seems to emphasize

' quality assurance and the various separate contractor organiza-tions that will implement it. Does the Staff consider that QC

. responsibilities and activities are separate from QA cr an

! integral part thereof? The Staff is requested to discuss its answer to~this question and to explain briefly how it will i

monitor QA and QC efforts for adequacy.

7. Applicants are requested to discuss commercial and recreational river traffic (if any) from two points of interest

(a) Practical methods of controlling same during off-normal plant conditions, and (b) 1&te potential for hazardous cargo posing

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a threat to the CRBR.

i 8. Applicants are_ requested to discuss the design L characteristics of the containment / confinement structures and the steam generator, with respect to challenges to those structures arising from transient (or accident) induced over-pressure and overtemperature conditions.- This-discussion

, should-address any engineered safety systems or-components l- that will be relied uaon for protection.(e.g., containment i

shell cooling), and should reference supportive test or opera-tional experience.

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9. The Staff's attention is directed to the i

discussion of protective action guidelines (PAGs) at pages 29-30 of the Partial Decision of February. 28, 1983. The Staff-is requested to address the question of whether a PAG

', revision for the CRBR should be made, and to explain its answer.

10. The Staff's testimony at Tr. 3694 anticipates

!' the need for further research and development on measurement capabilities to achieve DOE's goals for material control and accountability _at the DRP. The Staff is requested to explain

whether this additional effort is currently underway or definitively planned for the future, and the extent to which it is critical to the effectiveness of CRBR fuel safeguards measures. -

i 11. In discussing the energetics of accidents beyond design basis, the Staff offers the statement that there will be an "isentropic expansion yield to one atmosphere" (NUREG-0968,

Vol. 2, p. A. 2-5). The Staff is requested to discuss briefly what is the physical significance of this statemenc and the extent to which it contributes to any conservatism.in the l analyses of energy releases. Phenomenologically, how has the

, Staff satisfied itself that "approximately 2550 MJ would be required to produce a slug impact kinetic energy close to the

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l' head design capability of 75 MJ" (Ibid).

12. NUREG-0968 contains many. references to items

that are to be resolved at the OL review stage. In view of the apparently advanced stages of hardware design and procure-1 ment currently in being, the Board is concerned that said OL'

review (assuming a CP issues) may require substantive changes

of a costly and time consuming' nature, or in the' alternative, result in a compromise of performance safety. The Staff is requested to offer comments upon this situation.and to provide whatever insights.it can now offer.for avoiding such' problems.

13. With respect to the fuel system, the Staff has identified certain operational fallback positions potentially available to mitigate unresolved problems (NUREG-0968, Vol. 1,

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p. 4-47, 48). The Staff is requested to discuss briefly the extent'if any to which invoking such-operational fallbacks l

might' compromise the achievement.of CRBR programmatic objectives.

I 14. Operation with leaking fuel pins could conceivably offer the-opportunity for these pins to " inhale"~some amount of sodium whenever the reactor is-shut down. Should this occur,

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. l subsequent return to operation at power might then result in a significant increase in pellet-to-cladding gap conductance with an attendant off-normal performance of the fuel. The Staff is requested to comment upon whether it sees this as a problem requiring resolution and the reasons for its answer.

15. The Applicants have proposed a reliability assurance program that focuses primarily on plant protective.. ,,.4 systems. The Board requests Applicants to address the questids' of whether said program will (or ought to) take account of findings derived from the CRBR quality assurance program, and if so, describe the administrative mechanism envisaged to accomplish this. .

16. The SER discusses the impact of aerosol behavior on containment shell cooling. The Staff is requested to comment on whether changing concrete aggregate from calcitic to dolomitic limestone could significantly alter the behavior of the aerosols, and explain the basis for the answer.

17. What is the status of the Staff's review of, and what is the Staff's position with respect to, "The Eight Areas of Concern" listed in Section I, Table II of NUREG/CR-32247 l

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APPENDIX B .

4 EXHIBIT LIST 4

I. Applicants' Exhibits 1

Exhibit Marked No. Description For I.D. Offered Admitted i 59 Preliminary Safety Analysis 7368 8025 8025 Report (PSAR) Vol.1, Chapter 1 60 PSAR, Vol. 2, Sec. 2.1-2.4 7368 8025 8025 -

61 PSAR, Vol. 3, Sec. 2.5-2.A 7368 8025 8025

62 PSAR, Vol. 4, Sec. 2.B-3.1 7368 8025 8025 63 PSAR, Vol. 5, Sec. 3.2-3.A 7368 8025 8025 64 PSAR, Vol. 6, Sec. 4.1 through 7368 8025 8025 pg. 4.2-475 65 PSAR, Vol. 7, (pb. 4.2-476 7368 8025 8025 through pg. 4.4-194 -

66 PSAR, Vol. 8, Sec. 4.1-5.4 7368 -8025 8025 67 PSAR, Vol. 9, Sec. 5.5-7.4 7368 8025 8025 68 PSAR, Vol.-10, Sec. 7.5-9.2 7368 8025 8025 69 PSAR, Vol. 11, Sec. 9.3-9.9 7368 8025 .8025 70 PSAR, Vol. 12, Sec. 9.10-11.6 7368 8025 8025 71 PSAR, Vol. 13, Sec. 12.0-15.2 7368 8025 8025 4

72 PSAR, Vol. 14, Sec. 15.3-16.6 7368 8025 8025 73 PSAR, Vol. 15, Sec. 17.0-17.J 7368. 8025' 8025 74 PSAR, Vol. 16, Appendix A - 7368 8025 '8025 Appendix J 75 PSAR, Vol. 17, Questions. 7368 -8025 18025:

001.1-001.75 I

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Exhibit Marked No. Description For I.D. Offered Admitted 76 PSAR, Vol. 18, Questions 7369 8025 8025 001.76-001.324 77 PSAR, Vol. 19, Questions 7369 8025 8025 001.325-001.499 78 PSAR, Vol. 20, Questions 7369 8025 8025 001.500-001.704 79 PSAR, Vol. 21, Questions 7369 8025 8025 011-120 80 PSAR, Vol. 22, Questions 7369 8025 8025 130-222 81 PSAR, Vol. 23, Questions 7369 8025 8025 241.1-310.67 82 PSAR, Vol. 24, Questions 7369 8025 8025 321.1-430.7 83 PSAR, Vol. 25, Questions 7369 8025 8025 CS 210.1-CS 410.19 84 PSAR, Vol. 26, Questions 7369 8025 8025 CS 421.1-CS 721.1 85 PSAR, Vol. 27, Questions 7369 8025 8025 CS 760.1-CS 810.16 86 Applicants' Statement of 7369 8025 8025 General Information Pursuant to 10 C.F.R. S 50.33 87 Applicants' Testimony Concerning 7369 8025 8025 Whether HCDA's should be DBA's 88 WARD-D-0185, " Clinch River Breeder 7369 8025 8025 Reactor Plant Integrity of Primary add Intermediate Heat Transport System Piping in Containment,"

September 1977 r i 1

Exhibit Marked No. Description For I.D. Offered Admitted 89 Applicants' Testimony Concerning 7370 8025 8025 HCDA Analysis 90 CRBRP-3, Volume 1, " Hypothetical 7370 8025 8025 Core Disruptive Accident Considera-tions in CRBRP; Energetics and Structural Margin Beyond the Design Basis," Rev. 4, March 1982 91 CRBRP-3, Volume 2, " Hypothetical 7370 8025 8025 Core Disruptive Accident Considera-tions in CRBRP, Assessment of Thermal Margin Beyond the Design Base," Rev. 6, June 1983 92 ANL/ RAS 83-11; An Assessment of the 7370 8025 8025 Unprotected LOF Accident at EOC-4 in the CRBRP Heterogeneous Core Design 93 GEFR-00523; An Assessment of HCDA 7370 8025 8025 Energetics in the CRBRP Heterogeneous Reactor Core 94 Applicants' Testimony Concerning 7370 8025 8025 Board Areas of Interest Related to Emergency Planning 95 Applicants' Response to Board Areas 7370 8622 8623 of Interest 5 and 6 96 Conard, Robert A. , Review of Medica 17754 7758 7759 Findings in a Marshallese Population l

Twenty-Six Years After Accidental Exposure to Radioactive. Fallout, Brookhaven National Laboratory, January 1980.

97 Statement of Qualifications of 8019 8019 8020 Robert J. Bowman l

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II. Staff Exhibits r

i Exhibit Marked No. Description For I.D. Offered Admitted 22 NUREG-0786, " Site Suitability 7366 8616 8618 Report in the Matter of Clinch River Breeder Reactor Plant,"

Revision to March 4, 1977 Report (June 1982) [NRC Staff LWA Exhibit No. 1) 23 NUREG-0139, " Final Environmental 7366 8616 8618 Statement Related to Construction j and Operation of Clinch River Breeder Reactor Plant" (February 1977) (NRC Staff LWA-1 Exhibit No. 71 24 NUREG-0139, Supplement No. 1, 7366 8616 8618

" Supplement to Final Environmental Statement Related to Construction and Operation of Clinch River Breeder Reactor Plant," Volumes 1 and 2 (October 1982):(NRC Staff LWA-1 Exhibit No. 8]

25 Errata- Corrections to NUREG-0139, 7366 8616 861.8 Supplement No. 1 (December 10, 1982) (NRC Staff LWA-1 Exhibit No. 19]

26 NUREG-0968, " Safety Evaluation 7366 8616 8618 Report Related to the Construction-

of the Clinch River Breeder Reactor.

Plant," Volume 1, Main Report (March 1983) 27 NUREG-0968, " Safety Evaluation 7366- 8616 8618 Report Related to the Construction of the Clinch River Breeder Reactor Plant," Volume 2, Appendices (March 1983) 28 Errata Sheet for NUREG-0968, Vol.1. 7366 8616 ~ 8618 Main Report (July 11, 1983) t S . , , - . , n e - - , , - . .

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2 II. Staff Exhibits Exhibit Marked No. Description For I.D. Offered Admitted 29 NUREG-0968, Supplement No. 1, 7366 8616 8618

" Safety Evaluation Report Related to the Construction of the Clinch River Breeder -

Reactor Plant" (May 1983) 30 NUREG-0968, Supplement No. 2, 7366 8616 8618

" Safety Evaluation Report Related to the Construction -

of the Clinch River Breeder Reactor Plant" (May 1983) 31 Letter from Jesse C. Ebersole, 7366 8616 8618 Acting Chairman (ACRS), to Honorable Nunzio J. Palladino, Chairman (NRC),

Subject:

"ACRS Report ~on the Clinch River Breeder Reactor Plant" (April 19, 1983) 32 NRC Staff Testimony Concerning 7366 8616 8618 Adequacy of the Design Basis Accident Spectrum (With Attachment) 33 " Comparison of Clinch River 7367 8616 8618 Breeder Reactor Design Basis Accidents With Those For Light Water Reactors and Liq' uid Metal Cooled Fast Reactors, EGG-NTAP-6152 (J. Hanson, Idaho National Engineering Laboratory) (January 1983) 34 NUREG/CR-3240, " Comparison of CRBR 7367 8616 8618 Design Basis Events With Those of Foreign LMFBR Plants" (A. Agrawal,.

Brookhaven National Laboratory)

(March 1983)

3 II. Staff Exhibits Exhibit Marked No. Description For I.D. Offered Admitted 35 Memorandum from Robert J. 7367 8616 8168 Bosnak (Chief, Mechanical Engineering Branch, DE), to J. Nelson Grace (Program Director, Clinch River Breeder Reactor Program Office),

Subject:

" Seismic Margins in Small HTS Piping" (July 12, 1983) (With Attachments) 36 NRC Staff Testimony of Robert 7367 8616 8618 J. Dube On Board Question 10, Concerning Material Control and Accountability 37 NRC Staff Testimony of Thomas 7367 8616 8618 L. King On Board Question 4, Concerning Natural Convective Circulation 38 NRC Staff Testimony of Thomas L. 7367 8616 8618 King and Richard M. Stark On Board Question 12, Concerning Items Identified For Resolution at the Operating Licensing Stage 39 NRC Staff Testimony of Thomas L. 7367 8616 8618 King On Board Question 13, Concern-ing Fuel System Fallback Positions (With Attachment) 40 NRC Staff Testimony of Thomas L. King 7367 8616 8618 On Board Question 14, Concerning ,

Operation With Leaking Fuel Pins {

41 NRC Staff Testimony Regarding Analyses 7367 8616 8618 )

of Core Disruptive Accidents (With i Attachment) 42 NUREG/CR-3224, "An Assessment of CRBR 7367 8616 8618 Core Disruptive Accident Energetics" (T. G. Theofanous and C. R. Bell)

(March 11, 1983)

4 II. Staff Exhibits Exhibit Marked No. Description For I.D. Offered Admitted 43 NRC Staff Testimony On Board 7367 - 8616 8618 Question 9 Regarding Protec-tive Action Guides 44 NRC Staff Testimony of John G. 7367 8759 8759 Spraul and Algis J. Ignatonis On Board Question 6, Concerning Quality Assurance 45 Letter from J. Nelson Grace, 8120 8120 8121 NRC, to Gordon L. Chipman, Jr.,

DOE, dated July 25, 1983 re:

CRBR Required Actions Resulting from Salem ATWS Events 46 Portions of Compendium to 8532 8533 8534 NUREG CR-3224 47 Qualifications Statement of 8796 8796 8797 Virgil Brownlee III. Board Exhibits 125 Limited Appearance Statement of 7652 --

7652 Dr. Thomas B. Cochran Regard-ing Iscues Raised in the Con-struction Permit Proceeding 4

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APPENDIX C WITNESS LIST I. Applicants' Witnesses Transcript Name . Position Pages George H. Clare Manager, System Integration, TR 7735-7758, Westinghouse, Oak Ridge (CRBRP) 8619-8755 Robert E. Hottel Manager of Systems Engineering, TR 7375-7758 Westinghouse-Oak Ridge (CRBRP) ,

Ambrose L. Schwallie Manager, Fuel and Removable Assembly TR 7375-7758 Design, Westinghouse Advanced Energy Systems Division H. P. Plancheon Manager, Plant Control and Opera- TR 7375-7758 tions, Westinghouse-Oak Ridge (CRBRP)

Lee E. Strawbridge Manager, Nuclear Safety and TR 7375-7758, Licensing, Westinghouse Advanced 7760-7974, Energy Systems Division 7976-8025 Truman W. Ball Principal Engineer, Westinghouse TR 7760-7974 Advanced Energy Systems Division Robert Bowman Principal Civil Engineer, TR 8010-8025 Tennessee Valley Authority Hans K. Fauske President, Fauske & Associates, Inc. TR 7760-7974 H. Wayne Hibitts Chief, Safety and Environmental TR 7976-8025 Branch, CRBRP Project Office Eric K. Sliger Supervisor of Radiological TR 7976-8025 Preparedness Section, Tennessee Valley Authority Vernon Dale' Hedges Assistant Director for Quality TR 8619-8755 Assurance, CRBRP Project Office Joe W. Anderson Manager of Quality Assurance, TR 8619-8755 Tennessee Valley Authority Joel E. Karr Acting CRBRP Project Quality TR 8619-8755 Assurance Manager, Stone & Webster

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II. Staff Witnesses Transcript Name Position Pages Richard A. Becker Reactor Engineer, NRC CRBR TR 8028-8169 Program Office 8551-8566 Hukam C. Garg Electrical Engineer, NRC Equipment TR 8028-8169 Qualification Branch Shou-nien Hou Senior Mechanical Engineer, NRC TR 8028-8169 Mechanical Engineering Branch Thomas L. King Chief of Technical Review Branch, TR 8028-8169 NRC CRBR Program Office 8189-8262 8538-8566 Bill Morris Chief of NRC Electrical Engineer- TR 8028-8169 ing Branch Charles E. Rossi Section Leader, NRC Instrumenta- TR 8028-8169 tion and Control Systems Branch Robert Schemel Senior Human Factors Engineer, TR 8028-8169 NRC Human Factors Engineering Branch Jerry J. Swift Reactor Engineer, NRC CRBR Program TR 8028-8169 Office Ashok K. Agrawal Nuclear Engineer, Brookhaven TR 8028-8169 National Laboratory John E. Hanson Program Manager, Los Alamos TR 8028-8169 National Laboratory Edmund T. Rumble III Corporate Vice President, Science TR 8028-8169 Applications, Inc.

l Robert J. Dube Section Leader, NRC Fuel Facility TR 8172-8188 Licensing Branch, Office of Nuclear Material Safety and Safeguards Richard M. Stark Project Manager, NRC CRBR Program TR 8205-8219 Office 8559-8566 l Cardis L. Allen Senior Reactor Engineer, NRC CRBR TR 8265-8534 l Program' Office i

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II. Staff Witnesses. .  ;

Transcript Name Position Pages Larry W. Bell Nuclear Engineer, NRC Accident TR 8265-8534 Evaluation Branch Howard B. Holz Reactor Engineer, NRC Technical TR 8265-8534 Review Branch, CRBR Program Office Lewis G. Hulman Chief of NRC Accident Evaluation TR 8265-8534 Branch 8571-8616 ,

John K. Long Reactor Engineer, Technical Review TR 8265-8549 Branch, NRC CRBR Program Office Charles R. Bell Associate Group Leader, Los Alamos TR 8265-8534 National Laboratory Thomas A. Butler Staff Member, Los Alamos National TR 8265-8534-Laboratory 8549-8550 David Swanson Consultant, Applied Science Assoc. TR 8265-8549 Theofanis G. Theofanous Professor of Nuclear Engineering, TR 8265-8534 Purdue University Edard F. Brmugma, Jr. Health Physicist, NRC Radiological ~ TR 8571-8616 Assessment Branch Donald J. Perrotti Emergency Preparedness Analyst, 'NRC TR 8571-8616 4

Emergency Preparedness Branch John G. Spraul Quality Assurance Engineer, NRC. TR 8756-8800 Quality Assurance Branch,.0ffice of Inspection and Enforcement Algis J. Ignatonis Project Engineer, NRC Region II - TR 8756-8800 Office Virgil Brownlee Section Chief, NRC Region II Office TR P774-8800 1

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD t

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In the Matter of )

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UNITED STATES DEPARTMENT OF ENERGY )

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PROJECT MANAGEMENT CORPORATION ) Docket No. 50-537

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TENNESSEE VALLEY AUTHORITY )

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j (Clinch River Breeder Reactor Plant) )

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CERTIFICATE OF SERVICE Service has been effected on this date by personal delivery or first-class mail to the following:

Marshall E. Miller, Esquire Chairman Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission East-West Towers 4350 East-West Highway

, , Bethesda, Maryland 20814 (2 copies by hand)

Dr. Cadet H. Hand, Jr.

Director Bodega Marine Laboratory University of California West Side Road Bodega Bay, California 94923 (Air Express)

Mr. Gustave A. Linenberger Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission East-West Towers 4350 East-West Highway l Bethesda, Maryland 20814. (by hand)

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i' Stuart Treby, Esq.

Sherwin E. Turk, Esq. ~

Elaine I. Chan, Esq.

Geary S. Mizuno, Esq.

Office of Executive Legal Director U. S. Nuclear Regulatory Commission Maryland National Bank Building 7735 Old Georgetown Road Bethesda, Maryland 20014 (2 copies by hand)

• Atomic Safety & Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555

• Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555

• Docketing & Service Section Office of the Secretary U. S. Nuclear Regulatory Commission Washington, D. C. 20555 (original, 3 copies, and

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return copy)

William M. Leech, Jr., Attorney General William B. Hubbard, Chief Deputy Attorney General Michael D. Pearigen, Assistant Attorney General State of Tennessee Office of the Attorney General 450 James Robertson Parkway

Nashville, Tennessee 37219 Oak Ridge Public Library s Civic Center Oak Ridge, Tennessee 37830 ,

Herbert S. Sanger, Jr.,, Esquire -

Lewis E. Wallace, Esquire W. Walter LaRoche, Esquire James F. Burger,-Esquire Edward J. Vigluicci, Esquire Office of the General Counsel ,

Tennessee Valley Authority 400 West Summit Hill Drive e .

l Knoxville, Tennessee 37902" (2. copies)

( 's

! Barbara A. Finamorec Esquire S. Jacob Scherr, Esquire Natural Resources Defense Council, Inc.

1725 Eye Street, N.W.

Suite 600 Washington, D.C. 20006;

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Lawson McGhee Public Library 500 West Church Street Knoxville, Tennessee 37902 William E. Lantrip, Esquire Attorney for the City of Oak Ridge Municipal Building Post Office Box 1 Oak Ridge, Tennessee 37830 Leon Silverstrom, Esquire William D. Luck, Esquire U. S. Department of Energy 1000 Independence Avenue, S.W.

Room 6B-256--Forrestal Building Washington, D. C. 20585 (2 copies by hand)

Commissioner James Cotham Tennessee Department of Economic and Community Development Andrew Jackson Building, Suite 10007 Nashville, Tennessee 37219

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George Edgar Attorn for Project Management Corporation DATED: August 15, 1983

• / Denotes hand delivery to 1717 "H" Street, N.W., Washington, D.C.

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