ML20054H826

SITE-SUITABILITY Report in the Matter of the Clinch River Breeder Reactor Plant.Docket No. 50-537

ML20054H826
Person / Time
Site:	Clinch River
Issue date:	06/30/1982
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0786, NUREG-0786-R01, NUREG-786, NUREG-786-R1, NUDOCS 8206250025
Download: ML20054H826 (70)

Text

__ _

NUREG-0786 Site Suitability Report in the Matter of Clinch River Breeder Reactor Plant Docket No. 50-537 U.S. Department of Energy Tennessee Valley Authority Project Management Corporation Revision to March 4,1977 Report U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation i

f r a ano o

i hDR DO K O O 37 A PDR

NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The N RC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the NRC/GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations,and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free upon written request to the Division of Tech-nical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.

GPO Printed copy pnce: _$4.75

NUREG-0786

= :: . - - _ -- - _.

Site Suitability Report in the Matter of Clinch River Breeder Reactor Plant Docket No. 50-537 U.S. Department of Energy Tennessee Valley Authority Project Management Corporation Revision to March 4,1977 Report Manuscript Completed: May 1982 Date Published: June 1982 Clinch River Breeder Reactor Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission W:shington, D.C. 20555 p ~%,,

p

e...,*'

TABLE OF CONTENTS Page PREFACE TO 1982 EDITI0N.......................................... ix I. INTRODUCTION AND

SUMMARY

........................................ I-1 1.A Introduction ............................................... I-1 1.B Summary Conclusions ........................................ I-3 II. DESIGN CHARACTERISTICS .......................................... II-1 A. Facility Design and Principal Design Criteria .............. II-1

8. Fast Reactor Experience .................................... II-3 C. Key Aspects of the System Design ........................... II-6

1. Reactor Shutdown System ............................... II-6

2. Piping Integrity ...................................... II-8

3. Fuel Failure Propagation .............................. II-9

4. Residual Heat Removal ................................. II-11 D. Containment Design Considerations .......................... II-13

1. Sodium Hazards ........................................ II 2. Dose Mitigation Features of the Containment / Confinement System ...................... II-15

3. Containment Design Basis Accidents .................... II-16

4. Accommodation of Core Melt and Disruptive Accidents ... II-18 III. GE0 GRAPHY AND DEMOGRAPHY OF SITE ENVIRONS ....................... III-1 A. Site Description and Exclusion Area Control ................ III-1 B. Population Distribution .................................... III-1 C. Nearby Industrial, Transportation, and Military Facilities . III-6 D. Site Suitability Source Term Dose Consequences ............. III-8 E. Emergency Planning ......................................... III-10 IV. PHYSICAL SITE CHARACTERISTICS ....... ........................... IV-1 A. Meteorology .................... ........................... IV-1 l B. Hydrology ...................... ........................... IV-2 C. Geology and Seismology ..................................... IV-3 D. Foundation Engineering ..................................... IV .6 APPENDIX A - Clinch River Breeder Reactor Plant Design Criteria....... A-1 iii

LIST OF TABLES Page Table I - Subsystems of the Containment / Confinement System ......... 11-20 Table II - Deleted Table III - 1970 Census and Projected Cumulative Populations ......... III-3 Table IV - Site Suitability Source Term Assumptions and Dose Results .................................................. III-11 V

LIST OF FIGURES P_ age Figure 1 - Location of Clinch River Site in Relation to Counties and State ..................................................... I-5 Figure 2 - Clinch River Site Local Area .............................. 1-6 Figure 3 - Clinch River Site Exclusion Area and Property Boundaries .. I-7 Figure 4 - Plant Building Arrangement ................................ I-8 Figure 5 - Reactor Cavity, Vessel, and Head .......................... I-9 Figure 6 - The CRBRP Cycle ........................................... I-10 Figure 7 - Cumulative Population Distribution (1990).................. III-4 Figure 8 - Cumulative Population Distribution (2030).................. III-5 i

i l

t i

Vii l

l l

PREFACE TO 1982 EDITION In March 1977, the Office of Nuclear Reactor Regulation issued its Site Suitability Report (SSR) for the proposed Clinch River Breeder Plant (CRBRP).

That SSR documents the result of the staff's evaluation of the suitability of the proposed CRBRP site for a facility of the general size and type as the CRBRP from the standpoint of radiological health and safety considerations.

The staff concuded in that SSR that the proposed CRBRP site is suitable for such a facility.

Since the SSR was issued, several modifications have been made to the CRBRP design, additional data related to the site and its environs have been collected, and the Fast Flux Test Facility, a technological precursor to the CRBRP, has been completed and has commenced operation. In addition, new emergency planning requirements have been promulgated by the staff. This

, report is an update of the March 1977 SSR that reflects these matters and discusses them in tenas of the previous staff conclusion regarding the suitability of the proposed CRBRP site.

This report supersedes the March 1977 SSR. Substantive changes from that SSR are indicated by vertical bars in the right hand margin of the pages on which the changes occur.

Because the staff has not completed its safety review of CRBRP, a process which may lead to changes in design or design criteria, descriptions of specific CRBRP design features in this report are presented only as rapresentative of a facility of the general size and type as the CRBRP.

Similarly, the CRBRP design criteria in Appendix A are included only as representative design criteria for such a facility. The results of the staff safety review, when completed, will be documented in the Office of Nuclear Reactor Regulation's Safety Evaluation Report for the CRBRP.

That report is presently scheduled to be issued in March 1983.

Finally, although some changes have taken place since the March 1977 SSR was issued, the staff's conclusion in this report is unchanged from that in the 1977 SSR; that is, the proposed CRBRP site is suitable for a facility of the general size and type as the CRBRP from the standpoint of radiological health and safety considerations.

The principal staff contributors to this revision are:

Richard Becker, CRBR Program Office Larry Bell, Accident Evaluation Branch Richard Codell, Hydrology Geology Engineering Branch Charles Ferrell, Siting Analysis Branch John Long, Reactor Systems Branch i Richard McMullen, Geosciences Branch Bill Morris, CRBR Program Office Don Perrotti, Emergency Planning Licensing Branch Irwin Spickler, Accident Evaluation Branch Jerry Swift, CRBR Program Office i

ix J

, - - e

i l

l l This report was prepared by Mr. Richard M._ Stark, the NRC Project ,

Manager.- Mr. Stark may be contacted at the U.S. Nuclear Regulatory I i

Commission, Washington, D.C. 20555 or (301) 492-9732.

4 I

i 4

I I

i

(

l l

l l

X

I INTRODUCTION AND

SUMMARY

I.A Introduction Project Management Corporation (PMC) and the Tennessee Valley Authority (TVA) filed a 104(b) application with the United States Nuclear Regulatory Commission (NRC) for a license to construct and operate the proposed Clinch River Breeder Reactor Plant (CRBRP), a demonstration liquid metal fast breeder reactor (LMFBR) plant. The Application and Environmental Report (ER) were docketed on April 11, 1975. The Preliminary Safety Analysis Report (PSAR) was docketed on June 13, 1975. Legislation enacted by Congress in January 1976, authorized realignment of responsibilities of the participants in the project. The license application was amended in May 1976, to recognize that the U.S. Energy Research and Develop-ment Administration (ERDA) had the overall responsibility for managing the design, construction, and operation of the CRBRP.

In April 1977, at the applicants' request, licensing activity was suspended.

On October 1, 1977, the participant with the overall responsibility, ERDA, became the Department of Energy (DOE) as a result of PL 95-91, Department of Energy Organization Act and Executive Order 12009. The licensing activity was initiated in September 1981.

The proposed location of the plant is a 1364-acre site in the eastern part of Roane County, Tennessee and within the southwestern section of the City of Oak Ridge, Tennessee. The site is on a peninsula formed by the Clinch River and is approximately 3 miles south southeast of the Oak Ridge Gaseous Diffusion Plant, about 4 miles southwest of the Oak Ridge National Laboratory, and about 5 miles west of the TVA Melton Hill Dam (Figures 1-4).

Prior to a decision on a construction permit, NRC's regulations provide that the Director of Nuclear Reactor Regulation may authorize a limited amount of site work to be carried out. This authorization is known as a Limited Work Authorization (LWA). The regulations provide for the authorization of two types of work. Under one type, LWA-1, site preparation work, installation of temporary construction support facilities, excavation, construction of service facilities, and certain other construction not subject to the quality assurance requirements of 10 CFR 50, Appendix B, may be authorized. Under the second type, LWA-2, the installation of structural foundations may also be authorized.

An LWA-1 may be granted only after the hearing board has made all of the National Environmental Policy Act (NEPA) findings required by the NRC's regula-tions in 10 CFR Part 51 for the issuance of a construction permit, and has also determined that there is reasonable assurance that the proposed site is a suitable location for a nuclear power reactor of the general size and type proposed from a radiological health and safety standpoint. An LWA-2 may be granted if, in addition to the findings described above, the hearing board determines that there are no unresolved safety issues relating to the work to be authorized.

1-1

I The applicants have requested an LWA-1. A schedule for and description of site preparation activities to be conducted pursuant to 10 CFR 50.10(e)(1) (i.e.,

LWA-1) is provided in the current CRBRP license application and principally covers site preparation work, installation of temporary construction support facilities, excavation, and construction of service facilities; the installa-tion of structural foundations is not included.

l The CRBRP is a single-unit electric power plant with a liquid sodium-cooled loop-type breeder reactor utilizing a ceramic fuel of mixed uranium plutonium dioxides (UO2 -Pu02 ). With the initial reactor core, the design power is 975 megawatts thermal (MWt) and 380 megawatts electrical (MWe). Inplant uses of power would result in a net plant output of approximately 350 MWe. The anticipated gross thermal efficiency is 39% and the net plant efficiency is estimated to be 36%. The plant systems are being designed for a maximum power of 1120 MWt, and the potential radiological consequences are being evaluated based on this power level.

A single enrichment of 32.8% low-240 plutonium is used in the fuel assemblies.

The 14-inch long axial blanket sections above and below the 36-inch fuel section of each rod contains depleted UO2 pellets with 99.8% U238 and 0.2% U 23s. Each of the 156 fuel assemblies in the reactor core has 217 of these fuel rods.

Dispersed heterogeneous 1y throughout the inner core region are 82 inner blanket assemblies. Surrounding the inner core are 126 radial blanket assemblies which also have 14-inch long upper and lower blanket sections. Both inner and outer blanket assemblies contain 61 rods loaded with depleted UO 2 pellets. Refueling will be accomplished annually on a 2 year equilibrium cycle. Under equilibrium conditions, the fresh fuel batch is burned for one refueling cycle, at which time six inner blanket assemblies are removed and replaced by six fresh fuel j assemblies for reactivity to complete the second refueling cycle. The entire i

fuel and inner blanket batch is replaced at the end of the second refueling cycle. The radial blanket assemblies will remain in place for four refueling cycles. At the end of the second full core replacement, the inner row of blanket assemblies is replaced. On the subsequent refueling cycle (5th refuel-ing cycle), the outer row of radial blanket assemblies is replaced. The first 2 pre-equilibrium years will have the same fuel and blanket assembly rotation pattern, (except that only three inner blanket assemblies will be replaced after the first year) but the capacity factor is anticipated to be 35% the first year, 55% the second, and attaining the equilibrium value of 75% on the third year of operation. The reactor vessel and core are shown in Figure 5.

l Heat will be removed from the reactor core and blankets by the primary sodium l coolant. The primary system will operate with an inlet temperature of 730 F l and a mixed mean reactor outlet temperature of 995 F. Heated sodium will flow l in each of the three primary loops from the reactor vessel outlet through a 36-inch diameter pipe to a pump, and then through a 24-inch diameter pipe to the shell side of an intermediate heat exchanger (IHX), from which it will return through a 24-inch diameter pipe to the reactor core inlet. Each primary pump, rated at 33,500 gpm, will be driven normally by a 5,000-hp variable speed motor to provide load-following capability. A schematic of the heat removal cycle is shown in Figure 6. )

l I

I-2 )

i

1 l

l l

The entire primary heat transport system, consisting of the reactor vessel and the three connected loops, with associated pumps, valves, and heat exchangers, is contained within reinforced concrete vaults. These vaults are lined with steel membranes and the atmosphere (nitrogen, 2% oxygen) is sealed and control-led within each vault. The applicants propose to design the plant such that primary system piping systems and components are located in vaults at elevations beneath the operating floor. Access to these vaults will be through sealed hatches. The entire primary system is housed within a cylindrical steel con-teinment vessel approximately 186 feet in diameter and 240 feet high. The steel vessel is surrounded by a concrete confinement building, which is designed to protect the steel containment vessel from external missile hazards. The annular space between the two structures will be maintained at a negative pressure so that any leakage from within the steel containment could be collected and pro-cessed. The containment system is described in Section II.D.2. The containment /

confinement structures extend about 190 feet above the operating floor within containment, which is at site grade level.

In the intermediate heat exchangers, the heat is transferred from the radio-active primary sodium to the nonradioactive sodium in three secondary systems.

The 29,500 gpm pumps which provide the driving force for the sodium flow are in the cold legs of the intermediate loops. The intermediate heat exchangers are located inside containment, and the secondary pumps are located outside the reactor containment in the steam generator building. The operating pressure in the intermediate loops will be higher than the pressure in the primary loops, to minimize the potential for leakage of radioactive sodium into the inter-mediate loops.

The intermediate sodium will circulate through evaporators and superheaters in the steam generation system, which will also be located outside the containment building. Heat from the sodium will convert the feedwater passing through the evaporators into a mixture of water and steam (50% quality) at 621 F and 1750 psig, which will be directed to the steam drum where the water will be mechanically separated from the steam. The dry steam will flow to the super-heaters where additional heat from the intermediate sodium system will superheat the steam to 900 F at 1450 psig. The 436.8-MWe turbine generator driven by this steam will generate electricity at 22 to 24 kV. The voltage will be stepped up by transformers in the switchyard to 161 kV for delivery to the TVA system.

Figure 4 shows the arrangement of buildings on the site.

Waste heat released by condensation of exhaust steam from the turbine will be rejected to the atmosphere through the cooling towers, as described in Section 2.4 of the Staff's Final Environmental Statement (FES) for the CRBRP (NUREG-0139).

I.B Summary Conclusions In determining the acceptability of the proposed CRBRP site, the staff has considered the.following factors: population density and use characteristics of the site environs, including the exclusion area, low population zone, and population center distance; and physical characteristics of the site, including seismology, meteorology, geology, and hydrology. The staff concludes that the above characteristics of the Clinch River site are acceptable.

I-3

Although the staff's radiological safety review is continuing, and the staff is unable at this time to state a final position on whether the CRBR design pro-perly implements all the staff's design criteria, the staff believes sufficient information is available to identify: (1) a facility of the general size and type proposed; and (2) those design parameters that impact upon the question of site suitability. The identification of such a facility and design parameters is based on information submitted by the applicants and independently generated by the staff. In some cases, the applicants present design may not meet staff design criteria. Where this has occurred, in order to determine site suitability the staff has determined whether the state of technology would allow the staff's design criteria to be met. The staff finds that it is able to identify an adequate range of reasonable plant design parameters to conclude that the proposed CRBRP site is suitable for a facility of the general size and type proposed from the standpoint of radiological health and safety considerations.

The staff has concluded that the design can be made to meet the requirements set forth in 10 CFR Parts 50 and 100 and also must include specific capabilities to assure that the consequences associated with accidents beyond the design basis accident spectrum will be acceptably low, and comparable to those of a light water reactor. Requirements in this latter regard were specified by the staff and communicated to the applicant in a letter dated May 6, 1976. This letter is incorporated in the FES (NUREG-0139) as Appendix I.

By assuring conformance with the NRC's rules and regulations and by imposing those design requirements needed to assure that there is a low probability that large accidental releases would occur, the staff finds that sufficient review has been conducted regarding the principal features of the proposed plant, and that the design exhibits no unusual features or characteristics which cannot be subsequently modified within the present state of technology l to conform to the staff's requirements.

l t

I-4

i SCOTT CAMPBELL FENTRESS UNION GRAINGER MORGAN '4N DE RSON pi JEFFERSON 0*=

a aam* KttoX

?GC n uo r viat

• * ;" G 510" CUMBERLAND ROANE SEVIER Ltw S cit, LO DON BLOut T '

BLEDSOE RHEA ' * -#

M d' McMINN MONROE ,

E

,- N POLK $

70 0 to

_j IHHf

~

J

$CALLC'M'Ltb y . . - . . . . -- . . r. . A,n - 3 maawysLLe e s no rvn.Lt ,p

,;;*/ T (N N r , .,,

i V. G.A.=&

. . ===

• r MISS l. ALA ,

I t Figure 1 - Location of Clinch Riser Site in Relation to Counties and State 1-5

l l

l i

l I

l N

e 0 w RSA% ,

gppg;g, gg v.M Ana Nw CouhTv , (R;a

']% B0c%0AP' g

• qsyk f'WYh,',' =0A 00hTY ' A I"

• CR(it RIDGE f60 '

OR'2, p nA? tW. . 1 ENE WNe ,

, . , LABOEAT08Y s MAse w AN ' s .g \ ., ,

, ,, r' us:n COUNTY

% ..c=*

g".', $ .. % \ - E W - V se s  %' /

I PLAR h vssstow CAntv

, CREEL To ROANE COUNTY gSg Lou 00w county 9% %,, /*

~

\*

75 LEN0lR ' gg gw ** CITY 1

SSE , 3 SSW ..

S witts I

Figure 2 - Clinch River Site Local Area I-6 I i i

1 1

I

/ k

'/

n/j y4

\. (( \ i s I -

.. /' ___ .....,

4s

..- ./* \

s x,

/f,.y,' ~,ts........' f Exclusion Boundary

/ -

,. / ,. < ,

,,/ . /, \

ki .9: ,'

l , p' / .' S."

,;', s - \

s

..:, ,sg ,s "2;;," N  ?. 4

- ~ " ~ i.,

/ , m' a . -. m n - -

-_ ' ...,. + s s , -===

u,a m e.

x u ss e x p;t 9 6... ,"n ..

w omic. ( .,-.. .....

\

m , .sx

..............x c.),s

,,m .1, .,,,., e, s, < ,.

• \,\, 'f Z

• e, *

/l..

e ,  ;- .. .~

5 6 = J 1, -

.i........ ,g <

/,f'.

gi,.,.. .c..... ,..seu j,/ims

..:a.......... /f

--a=

-- 'a j

Exclusion Boundar ,. =t.

I.

/., ......-

• [!T,'.'f; . .' ,

in.

. T ' . , '"' . . -' Exclusion Boundary

.. .. ./ '. .

um =c uuss m-i ai. <, - .4 1 . . . . . . . . . . , . . .

n ..a s.

, . , . . . . . ,N'h. " .. .... e.;,, "Ja

....... I ,

)

..u.....

- . .i .s ,.....)..- .

n t

s. .,a,..< .........

e ........

^S ,

, , , , ..d., j A

Figure 3 - Clinch River Site Exclusion Area and Property Boundaries I-7

N

1. REACTOR CONTAlWMENT ButLDING

2. REACTOR SERVICE BUILD 6hG

3. RADWASTE AREA l 8 ILDING l 15 l / 5 CONTROL BUILDING 8, b* 6. DIESEL GENERATOR 8UILDING 97 i D T. INTERMEDIATE BAY yp 8. STEAM GENERATOR BUILDING 15 m) 13

~

g 9. MAINTENANCE BAY F

\ 1 f 10 AUXILIARY 8AY j

3- l 1. TURBINE GENERATOR BUILDING 4 5 6 10 12. SHOP AND WAREHOUSE 7  ; 13. GATE HOUSE

% '- 'l 8 14. OVE RFLOW HEAT REMOVAL SERVICE AREA 1 '2'

g 16 y ,~ - '

~ ~ ~ 'L- 19 15. PARKING LOT s - 34 16. EMERGENCY COOLING TOWERS 1[ 12 / 37. COOLING TOWER I7 18. C. W. PUMP HOUSE Q ' '

9 g 20 ,

89 GENERATING YARD u 20. START-UP YARD l

@17 i

e l 1 )D

._ /

j _ kj NOTE: HEAVY LINES IN01CATE SAF ETY-RELATED i (CATEGORY I) STRUCTURES SECURITY 8ARRIER O SOO 1000 I

SCALE OF FEET Figure 4 - Plant Building Arrangement

4 5 - .

- - . W----- T5 ![ ^

h a_------ __ . ,

p T

6 - M ._. dr ;l T ?k-FJ 4_ , m 7 9x g

iq '.

%[% 'W V. / l I

hyN/

E.

R EE:e Y 'i P fM0s Q' A ee g ==u 7

L 4 w5tM f- h*y ,,t, n=

. w

- :p. , .

'd#

, =

gp=,&,. 'y,L*e.,f Concrete gg Cu, .u..a Assembly

'. h"l 'kt / I r

- i y, * * .e s ]j }* '4 fI i' m.

. .h Reactor [} I // / /If' N  ;/ Outlet Piping ZRT ., -

Lb V

i

,f

/

k i I L h,

' @$ 4 #f'

.' Si,'.

?

j i

,;&s >

j f s ._ t

[

/~ $,;

q3

-1 J] '

'j .3

, . -)

% l.

e Ii , s ,

T

)

K Core' Y f  % eactor R I .

.~ oo., v ..

.- .N L+

-l -

, 60  % - h hedellon

' ? ' '

I i, \\

) f GA.Q l

k j b /

Reactor Cavity -

% ' '[

Q3 ) g jj

[ '  %

'x -

[

y i.h

'[

@ d gs d

'h Eha**E+%%.m .htw Figure 5 - Reactor Cavity, Vessel, and Head I-9

I r

o e t

a e %v aA n g r

n 9 it o e 3 t 4 la u er ct r a i /

CV I

e r L I u

s s

e e r r '

P in e s

wbur n '

e pm LT r

1 d

C n

o P

m rv u

P l N 1

Ii

[ r e

e r

• t as s

u wre e s p mRNdt l

e e ea c r ee y P in P u FH C hb r ig u P HT R I

I

\ B R

C g - N is mm e p au te g - h T

>05 r SD r

j 4 e t r F -

1 a 1 "0 e

6 hr 5

- r F e po 4 e D p a art r

0 0

9 S u

2' v

Eo v i u

g F

l 4T

)

r 3 e (

g F F e n s t

a p

- a ri t h 1 eo U 6 3 ed ac t

eex 5

6 N ao t

iL 9 InmHE d emu m

ri ed L

J I t

n So F ) )

3 5 yr m (s 9 aup imid o 94 t r

PS L oo r

o t W c M a 1 e 2 R 1 1

75 lll)l1l111l ll1,( l f J!

l t

l l

II DESIGN CHARACTERISTICS II.A Facility Design and Principal Design Criteria In accordance with the NRC's regulations (10 CFR 50.34(a)), an application for a nuclear power plant construction permit must include as part of the supporting technical information the principal design criteria for the proposed facility.

The principal design criteria establish the necessary design, fabrication,  !

construction, testing, and performance requirements for structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the public health and safety.

The principal design criteria which establish the minimum requirements for water-cooled nuclear power plants are identified in Appendix A to 10 CFR Part 50. Although these criteria provide guidance specifically criented toward water cooled reactors, the regulations state that these criteria "are intended to provide guidance in establishing the principal design criteria for such other (types) units."

In the field of liquid metal cooled fast breeder reactors (LMFBRs), three bodies have been extensively involved in the development of design criteria for LMFBRs. The American Nuclear Society (ANS) Standards Committee ANS-54 was established in 1970 to develop general design criteria for LMFBRs. In 1975 the committee issued criteria for trial use and comment. The Project Management Corporation (PMC) provided the AEC Regulatory staff with a draft copy of General Design Safety Criteria in March 1974. In July 1974 the AEC Regulatory staff (Directorate of Licensing) issued Interim General Design Criteria for the Clinch River Breeder Reactor Plant. PMC suggested changes to these interim criteria. Based on its review of the interim criteria, including consideration of the proposed PMC changes and related CRBRP PSAR information, the NRC staff (Office of Nuclear Reactor Regulation) developed and issued CRBRP Design Criteria in January 1976 specifically for the CRBRP licensing review. These are included as Appendix A.

These criteria are being reviewed as a part of the CP safety review but in their current form provide an example of the kinds of requirements acceptable to the staff for the principal design criteria of the CRBRP. They are not intended to be generally applicable to other LMFBR plants or concepts. The staff considered the guidance in 10 CFR Part 50 Appendix A as follows:

(1) where there is no substantial difference between CRBRP and light water ,

reactors (LWRs), the staff considered the LWR criteria applicable and adopted l the appropriate criteria; (2) for those LWR criteria considered to be generally applicable to CRBRP, the staff adopted, to the maximum extent practicable the LWR criteria with minor modifications; and (3) on the basis of its review, the staf7 identified and developed additional criteria for CRBRP where there are significa'nt differences between LWRs and the CRBRP.

s e

e d

9 II-1

During the course of its construction permit review, the staff will evaluate the applicants' specific engineering criteria and will require that any neces-sary modifications be made to these specific criteria to achieve satisfactory conformance with each of the principal criteria. Based on its review of the CRBRP PSAR to date, the staff finds that no considerations exist which would preclude proper implementation of the principal design criteria.

The Commission's regulations require that an applicant design, manufacture, and operate the plant to minimize the likelihood of accidents. To this end, a staf f-approved quality assurance program will be implemented to assure the necessary high integrity and reliability of the reactor system and safety features that would prevent or control accidents. In its review of the CRBRP PSAR, the staff is evaluating the adequacy of the Quality Assurance program for conformance with 10 CFR 50 Appendix B. Based on the review to date, the staff finds that there are no significant concerns which would prevent achieving an acceptable program.

The basic safety approach used by the staff is that the CRBRP should achieve a level of safety comparable to current generation light water reactor plants, according to all current criteria for evaluation, and that the design approaches to accomplish the required level of safety be similar or analogous to current practice. The staff position regarding this matter was also stated to the applicant in the letter dated May 6, 1976 (NUREG-0139) Appendix I.

In conducting its review of the CRBRP the staff is assuring itself that all structures, systems, and components important to safety will be designed and constructed to withstand the effects of various postulated accidents or natural phenomena. As discussed in Section 4 of this report, the staff has considered the environmental factors associated with the Clinch River site and has not identified any factors which would preclude proper design and operation of systems and structures which are required to shut down the reactor and maintain it in a safe shutdown condition. For example, the staff has reviewed the geological and seismological investigations conducted by the applicants and has determined the acceptable level of the vibratory ground motion to be utilized in the seismic design of the plant. However, there are additional engineering factors which constitute the remainder of the seismic design bases. Such additional factors include, the damping characteristics of systems and structures, development of a conservative design spectrum, and consideration of foundation conditions and their interaction with structures. Although the staff has reached the stated conclusions regarding the selection of the Safe Shutdown Earthquake (SSE), the staff's review of the seismic design criteria applicable to the CRBRP is not complete. As previously stated, it is the staff's intent that CRBRP achieve a level of safety comparable to other light water reactors.

Consequently, the staff is reviewing the proposed seismic design criteria to assure that they provide sufficient margins and utilize conservative limits such that their implementation will provide a level of safety comparable to that of light water reactors in resisting earthquake motions. The staff believes it feasible that requirements similar to those for the seismic design of light water reactors can be established and implemented for the CRBRP. This is based on the staff's knowledge of the state of technology. This will be addressed in the staff's Safety Evaluation Report.

II-2

The procedures employed by the staff in the review of other aspects of the CRBRP are also comparable to those employed for light water reactors (LWRs).

For example, the rigorous design codes and standards applied to light water reactors (LWRs) are also applied to CRBRP and in those instances where operating conditions exist, which are unique to an LMFBR, supplementary codes and standards are employed. An example is the application of codes which supplement the design requirements for compr ,ents which experience service temperatures higher than LWRs.

The containment system has many similarities to an LWR system and, in addition, will be designed to accommodate the effects of large sodium releases. This subject is addressed later in this report (II.D.3). In accordance with the regulations (10 CFR 100) the staff is also evaluating the containment system design and its performance by assuming a release of fuel and fission products to the reactor containment building (III.D.2). In analyzing the capability of the proposed containment system to acceptably reduce the leakage to the environ-ment, the staff is considering the effects of the negative differential pressure in the annulus as well as the potential for leakage that could bypass filtered pathways, as is done with LWRs. These calculations are subsequently used to judge the adequacy of the site in accordance with 10 CFR Part 100.11.

The staff has concluded that major emphasis must be placed on the prevention of accidents which could lead to core melt and disruption and subsequent loss of containment system integrity. To accomplish this objective, certain features and characteristics will be incorporated in the design to assure that the probability of an identified core disruptive accident initiator is acceptably low or to assure that identified accident scenarios are terminated before they progress to the core melt and disruptive stage. Those features which will be provided to satisfy these requirements and consequently have a significant bearing on the probability or consequences of accidental release of radioactive materials are discussed in Section II.C.

To ensure that the consequences of events beyond the design bases are comparable to similar events in LWRs, the staff has concluded that provisions should be made in the design so that there is a low likelihood of early containment system failure from the consequences of core melting and disruptive accidents (Section II.D.4).

II.B Fast Reactor Experience The United States experience with various LMFBR concepts has included the construction and operation of seven reactors, the Clementine, EBR-1, LAMPRE, Fermi, SEFOR, EBR-II, and FFTF. Significant safety events occurred in the operation of EBR-I and Fermi. In the EBR-I incident, the central core melted and was displaced generally outward during an experiment related to a positive component of the power coefficient. The radioactive gases released were dissipated harmlessly. Subsequent studies indicated that the incident did not represent a generic defect in LMFBRs, and was related to the particular design method of clamping the fuel elements in place. The core was later rebuilt and operated successfully until the reactor was decommissioned. In the Fermi accident, a metallic plate in the lower plenum became separated from its mountings and was swept upward to block the sodium flow into several subassemblies. In this coolant flow-starved condition, several subassemblies II-3

s

~

me ted as power was increased. There was no release of radioactivity outside the containment. In this case also, the reactor was repaired and operated successfully until its decommissioning. Flow entrance designs have been developed which cannot be blocked by a single sheet of metal in this fashion.

Other fast reactor experience in this country has been unmarred by serious safety problems. Notably, the SEFOR reactor operated without incident during its lifetime, and the EBR-II has performed with significant stability and dependability. EBR-II is a 20 MWe, 62.5 MWt liquid metal pool type fast breeder reactor using both uranium alloy fuel and mixed oxide fuel. The facility has been operational since August 1964. Although it has operated as a base-load experimental LMFBR power station, emphasis within the last 9-10 years has been primarily placed on the experimental irradiation of LMFBR fuels and structural materials. Extensive experience has been provided by EBR-II in other areas as well, including sodium purification, instrumentation and control system, inert gas systems, steam generator systems, fuel tag gas identification procedures, and fuel handling.

FFTF, a technological precursor to the CRBRP has completed construction and startup testing. Its first full length operation cycle (100 days at full power) was initiated in April 1982. The startup testing and initial operating experience were completed with a minimum of problems and have shown the FFTF performs as designed. The prior experience with LMFBRs, both in the US and abroad, provides the staff with further assurance that it is reasonable to expect that the implementation of its design criteria for the CRBRP can be met and that the state of technology is available to assure that the CRBRP can be built and operated successfully as planned.

Foreign experience with the operation of LMFBRs has been successful. In the United Kingdom, the Dounreay Fast Reactor (DFR) has operated since 1959 and is now decommissioned. More recently the Prototype Fast Reactor (PFR) has been completed and operated at power. Difficulties in the nonnuclear balance-of plant systems have delayed full power operation of PFR. In France, the Rapsodie reactor has operated since 1967 and the Phenix reactor has been operating successfully at its rated power of 250 MWe. The USSR has operated a 5-MWt experimental reactor (BR-5), a 12-MWt test reactor (BOR-60), and a dual purpose reactor (BN-350) for production of electricity and desalination of water. Some difficulties have been experienced in the performance of plant equipment, notably the steam generators. As far as the staff is able to determine, these operational difficulties have not resulted in any significant releases of radioactivity.

The operating experience of the French fast reactor Phenix has been especially noteworthy. This LMFBR demonstration plant was taken critical in August 1973.

After successfully completing the necessary prepower operational testing, it was connected to the Electricite de France grid in December 1973. The beginning of full industrial operation took place in July 1974. During its first year of operation, it achieved an availability of about 84%, while in the second year it reached about 74%. During this time it demonstrated excellent stability, both with respect to the nuclear plant and to the balance of the plant. The many refueling operations gave no indication of any problems concerning the introduction or removal of the fuel assemblies. Invaluable experience was gained in integrated system operation, pumps, monitoring for failed fuel and sodium chemistry.

II-4

In mid-1976, a leak developed in one of the Intermediate Heat Exchangers (IHX) in Phenix. There was no release of radioactivity. The problem was diagnosed and corrected in all six IHXs by removing the IHX to be repaired while operating at reduced power with the two unaffected loops. The problems that gave rise to these leaks have been found to be design dependent and are amenable to engineering solutions.

Phenix has had other minor operational problems which have been corrected with only minimum delay and difficulty. Recently, Phenix experienced a sodium le.ak l in one of its steam generators. This fault is currently being diagnosed but '

appears readily correctable. Phenix has also demonstrated a breeding ratio of 1.145 (predicted 1.128); has demonstrated the total fuel cycle; has an overall thermal efficiency of 45.2%; and by the end of 1980 had an average capacity factor of 61.9%. Phenix has shown that fast reactors have achieved sufficent technical maturity to give further confidence that there are no serious engineering problems associated with them that cannot be solved in a satis-factory manner.

II.B.1 The Licensing Background of LMFBRs Some of the early, small US LMFBRs were experimental government owned and operated facilities and were not reviewed in accordance with procedures for licensed reactors. However, licensing reviews were carried out for EBR-II, Fermi, and SEFOR. Preliminary assessments of the demonstration phase concepts were also prepared by the Regulatory staff. In the case of EBR-II and FFTF, full staff reviews were carried out even though these were not licensed facilities. There is, therefore, considerable staff experience in the safety evaluation of LMFBRs. Since the CRBRP design has evolved rather directly from that of FFTF, there is previous experience in the revie< of generic issues, such as the fuel, piping integrity, and core disruptive accidents (CDAs).

Thus, the staff has developed a considerable background in the review of the issues that are likely to arise in the case of CRBRP. In addition, the NRC has critically reviewed the ERDA base research program for LMIBRs and has provided its comments on the state of research programs in this area to meet licensing needs (L. Gossick to R. Fri, June 11, 1976).

The Advisory Committee on Reactor Safeguards (ACRS) has from time to time provided advice on LMFBRs. In its letter of August 20, 1976, " Report on Hypothetical Core Disruptive Accident for Liquid Metal Fast Breeder Reactors,"

the Committee concluded that the safety evaluation of LMFBRs should include consideration of CDAs at this time and that protective measures against their consequences should take appropriate account of the relative likelihood of l large excursions as compared to small ones. The ACRS also concluded that the  !

provisions for the containment of molten fuel should receive at least as much, if not more, emphasis than the possible occurrence of an (energetic) CDA and considered it prudent that plant design include provisions for dealing with a molten core mass in a way that public health and safety are not compromised.

The staff believes that the ACRS generic conclusions in this matter are consistent with the staff position of May 6, 1976 on the CRBRP. The ACRS involvement in previous LMFBR plant safety matters, including the FFTF safety review, provide a sound basis for dealing with the specific CRBRP considerations.

II-5

I The question of whether core melt accidents should or should not be included in the design basis of the plant has been given considerable attention. Such accidents are not included in the design bases of LWRs. The May 6, 1976 letter l already cited states the staff's position that for the CRBRP the probability of core melt and disruptive accidents can and must be reduced to a sufficiently low level to justify their exclusion from the design basis accident spectrum.

Four major design features are emphasized to assure this low level of probability: the redundancy and diversity of the scram systems, the redundancy and diversity of the heat removal systems, means to detect and cope with subassembly faults, and the assurance of continuing high integrity of the heat transport system.

II.C Key Aspects of the System Design The staff has concluded that the CRBRP design should assure the capability to minimize the risks associated with core meltdown events to an extent comparable to LWR designs. To ensure that the probability of core melt and disruptive accidents is low, emphasis is being placed on the prevention of conditions which could lead to such accidents. To help ensure that this is accomplished, the staff is emphasizing and requiring the achievement of an adequate degree of diversity, redundancy, and reliability in key safety features and aspects of the design. Proper implementation of the staff requirements in these areas will assure that the probability of core melt and disruptive accidents are sufficiently unlikely to justify their exclusion from the design basis accident spectrum.

The following sections include discussions of four examples of these features and design aspects. They are the reactor shutdown system, piping integrity, fuel failure propagation, and residual heat removal. The CRBRP design treats these areas somewhat differently from LWRs, as is appropriate for fast reactors.

For example, the CRBRP requires a continuous and reliable source of coolant flow to remove heat from the core in order to preclude sodium boiling to avoid fuel overheating and reactivity insertions. Due to the subcooled nature of the sodium coolant and the absence of a pressurized system, there is no need for emergency core coolant makeup systems similar to those employed in the emergency core cooling systems (ECCS) of LWRs. The CRBRP core cooling systems, therefore, reflect the need for heat removal rather than rapid coolant makeup. Transients which result in a large power / flow imbalance wherein more heat is generated than can be removed place emphasis on the reliability of the reactor shutdown system to limit the heat generation rate by nuclear shutdown. The higher power per unit of core volume of a fast reactor such as CRBRP also places higher demands on heat removal. The use of sodium coolant is well adapted to these problems because of its good heat transfer properties and subcooled operation.

Nevertheless, special attention has been given to these four areas described in the following sections. As indicated in these sections, the staff has concluded that feasible engineering solutions are available and will be adopted in each of these areas.

II.C.1 Reactor Shutdown System The staff is reviewing the instrumentation, control, and electrical portion of the proposed CRBRP reactor trip system, referred to as the reactor shutdown system, to determine the feasibility and soundness of the technical approach.

II-6

The CR8RP has proposed two reactor shutdown systems (the primary shutdown system and the secondary shutdown system) to reduce the reactor power level when the plant conditions require such an action.

The applicants have stated that the design of each of these two systems will conform to all applicable criteria in 10 CFR Part 50, Appendix A; CRBRP Design Criteria; NRC Regulatory Guides; and IEEE Standards. In addition, the appli-cants have listed other design criteria and bases which are unique to the CRBRP l reactor shutdown system. Some of these include specific requirements regarding i independence, diversity, and redundancy between the two systems. Other require- I ments include considerations of the special nature of the primary and inter-mediate coolant (liquid sodium) interfaces with the reactor shutdown system (RSS) sensors involved.

Although there are particular design aspects (such as those noted above) which are not the same as previously reviewed LWR designs, major portions of the two shutdown systems are very similar to designs reviewed in the past. The proposed systems will utilize solid state technology and include concepts of redundancy and testability which are similar to those of present LWR designs.

Therefore, based on the review of the proposed design information, operating data on solid state protection systems and development programs provided by the applicants to date, the staff concludes that the proposed design for the reactor shutdown systems for CRBRP is within the state of the art and is, therefore, a feasible design approach to satisfy NRC requirements. However, the staff recognizes that certain aspects of the design still require develop-ment by the applicant and review by the staff as more information becomes available.

The instrumentation, control, and electrical portions of the two shutdown systems have two distinct control rod and drive mechanisms associated with each. The primary system consists of 9 rods, each of which is connected to a lead screw which in turn is connected to a collapsible roller nut drive. Such devices have been previously used for nuclear plant applications. In response to a protective signal for rapid insertion of the control rods, the roller nuts disengage from the lead screw when the drive mechanism is deenergized and the control rods are inserted by the force of gravity. For a faster response, the CRBRP primary shutdown system rods are spring assisted. In order to reduce the probability of common-mode failures, the secondary system is designed to provide diversity in the latching mechanism, coupling, number of absorber pins, enrichment of absorber material, geometry, and other features.

The staff requires that the two systems be redundant in that either system acting alone will be designed to be capable of shutting down the reactor during l extreme conditions. No electrical or other external power is required for a scram of any control rod.

On the basis of the information presented in the PSAR and relying on its experience with such systems, the staff considers it feasible, by the use of the dual system, to reduce the probability of scram failure to a level con-sistent with the requirements of excluding CDAs from the design basis. The currently proposed dual system, which is under review, appears to have the II-7

i potential to comply with this requirement. The ongoing staff review is concen-trating on the additional information with regard to implementation of all the design criteria and design bases established for the CRBRP reactor shutdown system.

II.C.2 Piping Integrity Primary sodium flows through the core with a velocity of about 20 ft/second.

The hot sodium (995 F) is pumped to an intermediate heat exchanger (sodium-sodium); the stainless steel pipe from the reactor vessel outlet to the pump is 36-inch outside diameter (OD) with a wall thickness of 0.5 inches. The l remainder of the primary system piping is 24-inch OD and 0.5 inches wall thickness. The seam welded hot leg pipe is fabricated from Type 316 stainless steel plate; the cold leg piping is Type 304 stainless steel. Cold leg primary sodium returns to the reactor at 730 F.

, The sodium in the primary system is at a low operating pressure since the operating temperature is well below the temperature at which sodium will boil at near atmospheric pressure. The flow through the primary system is main-tained by the pressure head generated by the pumps and is about 160 psig at the pump exit. The sodium pressure in the cold leg is about 133 psig. There is a check valve in the cold leg piping between the IHX and reactor vessel inlet nozzle. The check valve will prevent backflow of large quantities of sodium through an idle loop.

As a result of the need for large mass flow-rates of sodium through the core for normal heat removal purposes, large rapid reductions in core flow can result in an undercooling situation wherein the fuel continues to generate heat, unless timely shutdown takes place. Sufficient heat removal must be continuously provided to prevent sodium boiling in the core. It is for this reason, that the cold leg piping of the primary loops between the check valve and the reactor vessel inlet must be conservatively designed and of high integrity. Cold leg pipe breaks between the reactor and check valve have the potential for initiating boiling in the core.

In pressurized water reactor (PWR) designs, which employ the loop concept similar to the CRBRP, the primary coolant (water) is maintained at high pres-sures (1600 to 2000 psi) to prevent boiling. Since the operating temperature of the sodium in an LMFBR is below its saturation temperature (about 1600 F at atmospheric pressure), the sodium coolant is not pressurized and there is no stored energy for flashing to vapor in the coolant as compared to that in a LWR in the event of a pipe break. These conditions for the CRBRP, plus the specific plant layout, result in the core not being uncovered and uncooled even in the event of a leak in the system. The inerted environment maintained in the cells where the primary piping is located is also less hostile with respect to degradation processes than the LWR service.

It is the staff's opinion, based on the following considerations, that the heat transport system can be designed for a high level of integrity and for continued assurance of this integrity throughout the operating history of the plant. The specifications include stringent nondestructive examination requirements. The material is characterized by high fracture toughness and corresponding large critical flaw size, a negligible growth rate of postulated defects and the II-8

probability of throughwall growth rather than elongation of defects. The system has low stored energy and is monitored by sensitive leak detection instruments. The staf f preliminary conclusion is that double ended rupture of l

the CRBRP primary cold leg piping (an event that could potentially lead to a CDA unless otherwise mitigated) need not be considered a design basis event.

This conclusion is conditioned on an acceptable preservice and inservice inspection program, a material surveillance program, continued research and development verifying material degradation processes, and verification of leak l detection system performance. The staff considers it feasible to implement l programs to satisfy these requirements. The staff intends to continue its review of the sodium cold leg piping to insure that the issues are resolved properly.

Because of its higher operating temperature, the same conclusions have not yet been reached concerning the hot leg piping (995 vs 730 F). The staff has studies underway to evalute the potential for and consequences of hot leg piping ruptures. Preliminary results obtained so far indicate that this event has more benign consequences with respect to core thermal conditions than the cold leg rupture. For example, a hot leg pipe rupture followed by a scram and a pump trip and normal flow coastdown does not appear to lead to boiling in the core. Analyses of this event are continuing and the results will be factored into any future requirements to assure that hot leg pipe ruptures, like the cold leg case, need not be considered as events that would lead to a CDA.

The applicants have submitted the results of their analysis regarding the integrity of the hot leg which is part of the current evaluation for the SER.

Pending satisfactory resolution of the staff review, the staff is proceeding on the basis that hot leg pipe ruptures should be considered as design basis events.

The staff has concluded that the combination of sufficient integrity for the primary coolant system boundary coupled with containment design features to cope with large sodium releases can be implemented to avoid unacceptable consequences from this source.

II.C.3 Fuel Failure Propagation The staff has specified that means to detect subassembly faults, to cope with these faults, and to protect against progressive subassembly fault propagation should be provided. These provisions are intended to help ensure that the probability of damage to a significant portion of the core due to subassembly' scale initiating events is very remote. The main concern is for rapid pro- l pagation; however, slow propagation which progresses to significant core l melting must also be prevented.

The events which could lead to significant subassembly damage include sub-assembly coolant inlet blockage, flow obstructions within the subassembly pin array, and individual or few pin ranaom failures. In the latter case of single or few pin random failures, propagation would likely occur on a pin-to pin basis within a subassembly before propagation to an adjacent subassembly occurred.

II-9

l Extensive analytical and experimental work has been performed at EBR-II, and is continuing, to verify that conditions which might arise during pi' ant operations l are unlikely to cause pin-to pin failure propagation. Aspects that have been covered include the effects of dimensional changes, wire wrap failure, fission gas release from pin failures, and other similar conditions which could lead to local flow disturbances or mechanical loadings. The results of this work thus far indicate that there should not be a significant potential for failure propagation beyond a few fuel pins under the anticipated operation conditions and limitations. Experimental and analytical work has been conducted on the effects of blockages within a pin bundle. The results, thus far, indicate that substantial blockages at the non-fuel inlet or outlet regions do not cause overheating; that inert planar blockages covering a few coolant subchannels in the fuel region do not cause any significant overheating; and that small heat producing (fuel material) blockages do not cause significant overheating of adjacent areas. There is, therefore, a substantial basis to anticipate that local faults affecting single or a few pins within a subassembly will not rapidly propagate to adjacent pins The design approach being utilized in the coolant inlet region of the CRBRP core should prevent large sudden flow blockage, such as that which led to extensive damage to two subassemblies in the Fermi reactor. Multiple inlet ports at different planes, with interposed strainers should prevent large pieces of debris from significantly reducing coolant flow to a subassembly module. Although sources of particulate debris in sufficient quantity to produce significant flow blockage have not been mechanistically identified, it may be postulated that this might occur. Such debris would be expected to be distributed rather generally throughout a large region of the core, so that it would be detectable by the core outlet thermocouples if it significantly reduced core flow. A condition which would not be detected by the outlet thermocouples before significant flow reduction had occurred is the preferential blockage of a single subassembly. Consistent with its practice for LWRs, the staff anticipates the requirement for a loose parts monitoring system which would detect loose parts prior to their causing possible flow blockage and detect structural degradation or wear of a primary loop component.

The applicants have proposed to utilize detection of fission gases in the cover gas and monitoring of the coolant fcr delayed neutrons as the principal means to determine the integrity of the fi l pins during operation. The applicants plan to continue to operate the reactor with some degree of cladding failure having taken place. Thermocouples will be located near the outlet of each fuel subassembly; however, these are not currently expected to be capable of detect-ing coolant temperature changes on an individual subassembly basis, which might be indicative of a condition such as developing subassembly flow blockage. As stated previously, these thermocouples would detect general flow reduction.

These means of core monitoring are regarded by the applicants to be diagnostic in nature. It is the staff view, that this instrumentation will be a necessary adjunct to assuring that fuel performance stays within acceptable limits from the standpoint of safety requirements. The proposed monitoring systems are technically feasible and the design and development relatively straightforward.

Whether or not more sensitive and faster response monitoring systems will be required depends on the outcome of ongoing research and development work and completion of the staff safety review. The current staff position is that of II-10

I not being yet convinced that the staff requirements regarding subassembly propagation have been satisfied.

As a minimum, the staff will require that a fuel surveillance program be established and carried out during initial operation to verify fuel performance as a function of burnup. In addition, it is likely that operation of the plant with the degree of failed cladding proposed by the applicants will be excluded until such time that it is better established that the proposed limits are acceptable from the standpoint of limiting failure propagation potential. Such t a limitation may also be needed to ensure adequate sensitivity of the cover gas

! and delayed neutron failed fuel detection systems. This restriction could be later removed. While the staff considers it unlikely that ongoing research and development work will disclose a condition under which a serious degree of fuel failure propagation could occur within a subassembly, there is need to consider what would have to be done if this were not borne out. Under this circumstance, one option would be to operate the plant under a set of conditions more restrictive than those currently proposed. Another option would be to develop and provide for installation of instrumentation which could reliably detect subassembly flow disturbances before heat transfer conditions were seriously disrupted.

From the review of the CRBRP design to date, the staff has concluded that fuel pin failures which might occur under various plant operation conditions, including design transients, are unlikely to create conditions under which significant fuel failure propagation within a subassembly would occur. The general world experience with LMFBRs supports this view. The applicants' research and development completed and planned appears reasonably constituted to assure acquisition of data on fuel performance under normal and off-normal conditions from which confirmatory analyses can be done. It is likely, however, that initial fuel surveillance requirements and limitation of continued opera-tion with failed fuel more stringent than those presently proposed by the applicants will be required.

Based on the review of the proposed fuel element and inlet plenum design, the preliminary analysis and planned research and development programs, and subject to fuel surveillance requirements and exposure limitations, the staff has concluded that it is possible to limit the potential for fuel failure propaga-tion beyond a single subassembly to such a level that it need not be considered as an initiator of a whole core accident.

II.C.4 Residual Heat Removal System Design Criterion 35 specifies the guidance and criteria for the reactor residual heat removal system. In implementing the staff requirements for residual heat removal, the applicants must provide a design which can safely withstand a wide spectrum of anticipated and unlikely failures and accidents.

In so doing, the design is made such that the probability of accidents leading to severe core damage and substantial releases of radioactivity is very remote.

To illustrate, it is expected that once or twice during the plant lifetime all offsite power will be lost. When this occurs, power to main heat transport system pumps is lost, resulting in a loss of normal flow. The reactor is shut down, but decay heat is generated and must be removed if damage to the fuel is 11-11

l to be prevented. Three onsite diesel generators will be available to provide emergency alternating current (ac) power. l l

Because of the importance of effective decay heat removal, the CRBRP will provide redundant and diverse decay heat removal systems. Removal of reactor decay heat following reactor shutdown is through the Primary Heat Transport System (PHTS) and the Intermediate Heat Transport System (IHTS) loops. With pony motor flow, each loop by itself is adequate to removal all short term and long term decay heat from the reactor and transport it to the Steam Generator -

System (SGS). Each SGS loop is capable of removing all short term and long term decay heat from the IHTS provided that the Main Condenser /Feedwater train is available for that loop. The Main Condenser /Feedwater train is the normal path for heat removal from the SGS; however, it is comprised of non-safety class equipment and its operation is not required for safe shutdown of the plant.

In the event that the Main Condenser /Feedwater train is unavailable, the Residual Heat Removal Systems (RHRSS) provide reactor decay heat removal capa-bility for all plant shutdown conditions, including those associated with postulated accident events. The proposed RHRSS consist of redundant Steam Generator Auxiliary Heat Removal Systems (SGAHRS) and a diverse Direct Heat Removal System (DHRS). The main function of SGAHRS is to provide redundant decay heat removal paths when either the main condenser heat sink or the main feedwater supply is unavailable. The SGAHRS performs its functions using two subsystems - short and long term heat removal subsystems. The short term subsystem removes heat received from the Heat Transport Systems (PHTS, IHTS, SGS) by venting of steam from the steam drum to the atmosphere through power relief valves. The expanded water volume is replaced by the Auxiliary Feedwater Subsystem (AFWS). This subsystem draws water from a seismic Category I Protected Water Storage Tank (PWST) and pumps it to the steam drum. Three pumps are provided to supply the total auxiliary feedwater flow rate required for all three loops. Two of these pumps are driven by electrical motors powered by normal or emergency plant A.C. power; each of these pumps has the capacity of delivering 50% of the total auxiliary feedwater flow rate. Diverse motive power is used for the third pump by means of a steam turbine which uses steam bled from the steam drum (s); this pump has the capacity to deliver 100% of the total auxiliary feedwater flow rate. The long term heat removal subsystem is a Protected Air Cooled Condenser (PACC) located at a higher elevation than the steam drums. This PACC rejects heat to the atmosphere. Saturated steam is supplied to the condenser from the steam drum, and saturated water is returned by gravity flow. The PACC utilizes a fan to force air across the condenser tubes for peak performance; however, it can operate in a natural convection mode under reduced decay heat loads.

In the event that none of the steam generator decay heat removal paths are available, the applicants claim that the DHRS provides an independent, diverse system for use in decay heat removal on an emergency basis. The NRC staff has not concluded that the DHRS provides the necessary diversity because it does

~

not retain its functional capability in the event of a reduced sodium inventory.

The reactor sodium level is not sufficient in this case to provide the overflow for DHRS operation. DHRS operation requires a full sodium inventory in the PHTS to provide the necessary overflow to the overflow tank which in turn is connected to a pump and heat exchanger. Furthermore, operation of DHRS requires II-12

forced circulation through a PHTS loop by means of a PHTS pony motor to provide sufficient core flow and coolant distribution such that fuel and reactor vessel temperatures are maintained at acceptable levels. The staff is continuing to review the safety adequacy of this dependence of the DHRS on the PHTS loop and pony motor.

Assuming that the staff concerns regarding the degree of diversity for DHRS are not resolved, there are alternative diverse cooling concepts that can be implemented to avoid complete dependence on the steam generation system for an alternative path for decay heat removal.

Although the CRBRP will be designed to have adequate natural convection capability to dissipate plant sensible and decay heat following reactor shutdown, the NRC staff will not allow credit for this until sufficient experimental bases are provided from the applicants. In the early testing of FFTF, sufficient natural circulation cooling was demonstrated for that facility. This experimental data set does reinforce the applicants' contention that this mode is a workable diverse mode to remove plant sensible and decay heat. Whether or not sufficient experimental bases exist to accept natural circulation cooling for diversity is currently under evaluation for the SER. Furthermore, in the event that natural circulation cannot be relied upon to provide a diverse means of assuring flow in the normal heat rejection paths, the applicants will be required to commit to providing: (1) motive and control power to assure that for onsite electric power system operation (assuming offsite power is not available) forced convection flow is maintained throughout the entire decay heat removal train (PHTS, IHTS, SGS, SGAHRS) assuming a single failure; and (2) motive and control power diversity to assure forced convection flow throughout the entire heat removal train (PHTS, IHTS, SGS, SGAHRS) for a nominal time period of two hours following reactor shutdown and assuming the loss of normal onsite AC and offsite power supplies.

The staff has concluded that the state of the art is such that it is technically feasible to provide an adequate residual heat removal system, and will require, in the course of its safety review, that a system meeting the above criteria is provided.

II.D Containment Design Considerations II.D.1 Sodium Hazards At the operating temperatures of CRBRP, sodium will ignite and burn readily if sprayed into air or reduced-oxygen atmospheres (the burning rate will be slower in the reduced-oxygen atmospheres). As a pool, sodium burns readily in air but slowly in reduced-oxygen atmospheres. The heat of a sodium fire, if transferred to concrete, can cause the concrete to give up water and carbon dioxide.

Furthermore, if hot sodium is allowed to contact concrete directly, the water released from the concrete reacts exothermally with the sodium generating sodium hydroxide and hydrogen. Some organic compounds have also been reported, by Sandia, from this reaction. The net effect of these reactions is an increase in containment cell temperatures and pressures, with structural l degradation of the concrete and the production of potentially explosive hydrogen. Primary sodium is made radioactive during reactor operation and for periods up to two weeks following shutdown, and in addition may be contaminated 11-13

I with small amounts of activated corrosion products, tritium, fuel and fission products from failed or contaminated fuel. Secondary sodium is essentially non-radioactive except for small amounts of tritium.

l In the event of a primary or intermediate coolant system release, sodium l presents a variety of challenges to the containment, summarized as follows:

(1) Mechanical - The deterioration of concrete by sodium can weaken structures, cause cracking, and enlarge leak paths. Therefore, cell liners are used to reduce the likelihood of direct contact between sodium and concrete.

(2) Thermal - The chemical heat of sodium reactions with oxygen or concrete can build up pressures within inerted cells or the containment building which must be included as part of the containment design basis. Accordingly, the direct contact between sodium and concrete is avoided, and the PHTS cells are designed for 30 psig.

(3) Explosive - The generation of hydrogen from reactions with water (or I concrete) can lead to explosive mixtures in the air atmospheres of the RCB. Water is therefore kept to a minimum in buildings containing large amounts of sodium. Hydrogen recombiners are provided in LWRs to control hydrogen that may evolve as a result of a Loss of Coolant Accient (LOCA).

The CRBRP applicant claim that no hydrogen recombiner is required in the presence of sodium oxide, as it has a catalytic effect in promoting recombination and keeping the hydrogen concentration below the explosive limit.

(4) Non-radiological toxicity - If released from containment or the steam generator building, large quantities of nonradioactive sodium could be an inhalation and environmental hazard. Prompt methods to suppress or extinguish sodium fires, as well as isolation, can prevent the release of the hazardous smoke.

(5) Radiological toxicity - If released from containment, radioactive sodium could be a biological and environmental hazard. This problem is also controlled by the methods for extinguishing sodium fires as well as by

imposing technical specifications to limit operation with contaminated sodium.

(6) Filters - The dense smoke from sodium fires can rapidly plug ventilation ,

filters. Scrubbers or prefilters are generally required to eliminate this l problem. I l

l Although the adequacy of the applicants' measures is under review, the staff believes that there is sufficient experience with handling sodium at experi-mental and testing facilities to conclude that features can be incorporated in the design to alleviate the above sodium hazards.

II.D.2 Dose Mitigation Features of the Containment / Confinement System The CRBRP containment systems consist of a set of barriers, filters, vents, pressure relief systems and associated equipment designed to mitigate the release of radioactivity to the environment under all anticipated circumstances II-14 i l

and under many highly unlikely accident conditions. The design incorporates a confinement / containment system based on a low leakage, free-standing 1-1/2 inch steel structure surrounded by a 4-ft thick concrete confinement and missile barrier shell, with a 5-ft annulus between them. This annular space is main-tained at a negative pressure, and during accident conditions the exhaust is filtered. This type of system is sometimes referred to as a dual containment.

l The inside of the steel containment vessel is compartmentalized, consisting of reinforced concrete vaults which contain the reactor vessel, Primary Heat Transport System (PHTS), and auxiliary support systems. The proposed design internal pressure for the containment system is 10 psig, and the maximum ,

allowable leakage rate at the design pressure is 0.1% per day. Penetrations will be designed to ensure that leakage from the steel shell which bypasses the annulus filters is limited to less than 0.001% per day. Appropriate technical specifications and leakage testing requirements will verify these values. The containment isolation system provide the means to close valves in lines penetrating containment to assure that a barrier to the release of radioactive gas and particulates is provided. During normal operation, air for ventilation purposes is brought into the containment building in a once-through ventilation system. The staff has not agreed with the applicant regarding the justification for this ventilation arrangement which results in maintaining large purge valves open during normal operation. Similar proposals for continuous contain-ment purging has been suggested for LWRs but the staff has a position of limiting containment purging during normal LWR plant operations to reduce the possibility of a valve failing to close during accident conditions. Although the applicant proposes to utilize the automatic containment isolation for the ventilation system, the staff has not concluded that this alone will be acceptable.

There are other alternatives, such as coolers, acceptable to the staff which can be utilized to control the environment within containment during normal operations. The staff is continuing its review of this subject, as well as the containment isolation system, and will report its results in the Safety Evaluation Report.

The applicants' design provides that the annulus region between the steel and concrete structures will be maintained at a slight negative pressure (0.25 inch water gauge, W. G.) by exhausting up to 4000 cfm from the annulus through blowers. The applicants have committed to provide a redundant filter system with provisions for partial recirculation of the annulus exhaust, to be used in case radioactivity is detected in the containment building. This filter system will be required to conform to the appropriate air cleanup system requirements of Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorp-tion Units of Light-Water-Cooled Nuclear Power Plants." In the event of the release of radioactivity within the containment building, the following steps will be taken automatically:

(1) Penetrations into the containment building atmosphere will be closed.

(2) The annulus exhaust system, which normally withdraws ( 4000 cfm to main-tain the annulus negative pressure of 1/4-inch W. G., will be directed I through the annulus filter system.

In order to evaluate the suitability of the CRBRP site to meet the radiological exposure guidelines of 10 CFR Part 100 (see Section III.D), the staff is II-15

l l

l assuming that the above-described features are effective in mitigating the effects of all credible events. Based on the available experience and informa-tion, the staff concludes that it will be technically feasible to properly incorporate these features into the CRBRP design.

However, as discussed in Section II.D.4, in the event of the occurrence of an event beyond the design bases involving a core melt accident, pressure in excess of the 10 psig design pressure could build up in the containment building as a result of releases of chemical energy and core decay heat. If this pressure reaches a level that jeopardizes containment, the applicants have proposed to manually activate a containment cleanup system which discharges through a filtration system to the environment. Providing measures to cool the containment environment and/or mitigate the heat transferred to the containment atmosphere can alleviate the buildup of such pressures. The applicants have proposed an annulus air cooling system to cool the containment shell under such ,

conditions. The means of dissipating heat and pressure generated in the reactor cavity in the event of a core meltdown or in the cells as a result of l sodium spill effects beyond the proposed design basis are important aspects in the review of the adequacy of the containment system to mitigate the consequence of events beyond the design bases. As noted in Section II.A, special provisions in accordance with the staff's requirements of May 5, 1976 are being provided for these types of events. For convenience and completeness they are included in this discussion. Sections II.D.3 and II.D.4 discuss these subjects in more detail.

The principal subsystems of the reactor containment system proposed by the applicants are summarized in Table I.

II.D.3 Containment Design Basis Accidents The CRBRP containment design has evolved to a-concept similar to that used in modern US LWRs. The staff believes that such systems are practical to build, and can incorporate sufficient dose mitigation features to assure that site boundary doses will remain within the 10 CFR 100 guidelines. The al'ility of this containment system to mitigate the site suitability source term is described in section III.D.

The staff issued general safety design criteria for the CRBRP, including Criterion 41 " Containment Design Basis," which states in part "...the reactor containment structure, including access openings and penetrations, and if necessary, in conjunction with additional post accident heat removal systems including ex-vessel systems, shall be designed so that the containment struc-tule and its internal compartments can accommodate, without exceeding the design leakage rate, and with sufficient margin, the calculated pressure and temperature conditions resulting from normal operation, anticipated operational occurrences and any of the postulated accidents."

In an LMFBR, the accidents which represent the principal challenges to con-tainment are sodium fires coupled with potential sodium-concrete reactions which result from failure and subsequent release of sodium from the Primary Heat Transport System (PHTS) equipment. Following sodium release, combustion with the oxygen in the cell atmosphere occurs resulting in increasing cell pressures and temperatures. In order to satisfy the above criterion, the II-16

containment design basis, including the inner cell system, must envelope the pressures and temperatures resulting from consideration of a spectrum of sodium spray and pool fires. The staff's present view is that these effects are not coupled with any sodium concrete reactions because the applicants have proposed that the ccll liners will be engineered safety features. The sodium releases under consideration cover a spectrum of posulated component and piping failures of different sizes and locations, and other properties sufficient to provide assurance all possible fire accidents are covered.

The staff has analyzed a wide spectrum of primary sodium pipe breaks up to and including double ended ruptures. The staff finds that the extent of inner cell pressurization depends heavily on the existence of a steel liner constructed in accordance with staff requirements for engineered safety features (ESF) which would prevent the released sodium from coming in contact with the cell concrete.

For example, if an ESF steel liner exists, the staff analyses indicate that a maximum cell pressure of approximately 28 psig would result from a sodium spray fire release in a nitrogen-2% oxygen environment. This environment eixsts in the primary sodium equipment cells to mitigate the extent of sodium combustion with oxygen. Without the mitigating effects of the steel liners, the staff analyses indicate that maximum cell pressures of approximately 80 psig could result from a sodium spray fire coupled with subsequent sodium-concrete interactions. The staff considers the above 28 psig cell pressure to be within the capability of reinforced concrete cell structures used in present day light water reactor plants.

Accordingly, the staff has also conducted analyses in which the inner cell is assumed to fail due to overpressurization allowing free communication with the air environment in the RCB. In this instance, maximum calculated pressures in the RCB were found to be about 50 psig. These calculations reinforce the need for the Primary Heat Transport System (PHTS) cell liners to be ESF, although an alternative, technically feasible, solution of increasing the RCB design pressure to 50 psig is available.

Additional preliminary analyses were conducted by the staff to examine the design option of incorporating a vent system to allow communication between the inner cell system and the outer RCB. The extent of inner cell pressurization can be mitigated by venting in a controlled manner from the cell to the RCB.

Depending on the size of the vet passage, it appears that the cell and RCB peak pressures can be maintained below 30 psig. Furthermore, with the incor-poration of a high-integrity ESF liner system in the cell design, the staff's preliminary results indicate that a cell overpressure protection vent system may not be required, since the inner cell and RCB design pressures of approxi-mately 30 psig and 10 psig, respectively, would envelope the calculated accident pressures.

Based on the above analyses and present day nuclear power plant construction capabilities, the staff considers it technically feasible to implement design provisions to satisfy the aforementioned general safety design criterion.

Examples of design features which can be incorporated to satisfy the safety requirements include increasing the cell and containment shell structural capability, and providing controlled venting of the cell. The applicants, in a letter dated October 15, 1976, committed to increasing the PHTS cell design pressure from 10 psig to 30 psig, and the staff is currently evaluating the II-17

safety adequacy of the applicants' proposal, plus pursuing with the applicants other design features that will be committed to in order to satisfy the staff's safety criterion for the containment design basis accident.

II.D.4 Accommodation of Core Melt and Disruptive Accidents As previously indicated, the staff has concluded that provisions should be made in the CRBRP design such that there is an extremely low likelihood that potential core melt and disruptive accidents could result in early containment system failure. This requirement arises from the basic position that the CRBRP should achieve a level of safety comparable to current generation LWR plants. The requirements referred to are those identified in the staff letter of May 6, 1976 (NUREG-0139, Appendix I). From among the spectrum of events beyond the design bases, the staff has examined various sequences that lead to core melting and disruption. These evaluations have resulted in a delineation of the accident conditions, including the nuclear, thermal, structural, and radiological behaviors, associated with such events in the CRBRP design. The staff concluded that some accidents could lead to energetic disassembly of the core and the production of vaporized fuel, as well as large scale fuel melting in the core. These characteristics must be considered in the design to ensure that the consequences of these accidents in the CRBRP are made comparable to core melt accidents in LWRs. The Reactor Safety Study (WASH-1400) indicates that most LWR core melt accidents do not result in early (less than one hour) containment failure, but many involve such failure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The uncontrolled nature of these failures leads to the presumption that the con-sequences (in LWRs) could be serious. The staff determined that the CRBRP i containment system should be protected from a broad range of conditions i involving energetic disassembly and production of vaporized fuel, and other I manifestations of core melt accidents. In the letter of May 6, 1976, the staff '

l specified a time of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to maintain containment system integrity for these conditions. Recently the staff has been considering alternative criteria for evaluating core melt accidents, in lieu of the 24-hour criterion.

The staff's study has led to the following grouping of potential core disruptive accident consequences in order of increasing severity:

(1) Primary system remains intact; no major releases of radioactive materials.

(2) Primary system initially intact but ultimately fails due to ineffective long term decay heat removal (of the order of hours or more); the steel liners in the reactor cavity could fail either by penetration of core debris or by excessive steam pressure on the back side created by water released from heated concrete structures. The reactor cavity atmosphere would be pressurized from chemical reaction products and/or sodium vapor i and would eventually vent into the outer containment. Ultimately, the sodium would boil off. Outer containment could fail due to hydrogen explosion or overpressurization and/or structural thermal degradation.

Core debris would penetrate into the concrete basement and may ultimately reach the foundation soil. Fission products in the upper containment building would be volatilized and released to the environment. Consequences may exceed 10 CFR 100 guideline values.

i l l

l II-18

(3) Primary system seals fail due to initial mechanical and/or thermal loads.

Some sodium fuel vapor and fission products are expelled into the head access area. The long term behavior is about the same as in II above.

)

(4) Primary system fails due to the excessive mechanical loads. Outlet piping (three loops) fails and sodium is expelled into the reactor guard vessel.

Substantial quantities of fuel, sodium or sodium vapor and fission products are released to the outer containment. Initial failure of the containment I

due to these effects is possible. The long term behavior is about the same as III above, but the consequences may be more severe.

The applicants have proposed to incorporate features to mitigate the above behavior and to reduce the probability of failure of the containment. These include a filtered vent system to relieve containment pressure, a containment purge system to reduce the potential for hydrogen explosions, fans to cool the annulus between the steel containment shell and the confinement structure, and vents to relieve pressure from gases generated behind the reactor cavity cell liners. These provisions are currently under review by the staff.

If the provisions proposed by the applicants are determined to be inadequate these may be upgraded, but the staff is also aware of other feasible design features which separately or in combination could reduce the probability of containment failure to an acceptable level. These include a cooling system to transfer decay heat and chemical reactant heat from the core debris and sodium deposited in the reactor cavity to outside containment and installation of protective materials in the reactor cavity to reduce the production of reactants and heat from interaction of sodium with concrete.

With respect to the effects of mechanical loads noted for Items 3 and 4 above, the applicants have chosen to contain the effects of energetics by means of a strong head design. Although the acceptability of the present head design has not been established yet, the staff has examined potential workable designs which can be used to implement the staff's containment protection requirements.

If the applicants continue to pursue the same approach (i.e., accommodation via a strong head design), the present head shear ring concept could be redesigned to significantly increase its structural capability. Other options, such as head hold down and/or missile barrier devices can be used to provide physical protection of the containment from potential missiles due to reactor head component failures. The staff believes that the technology exists to design and build such devices; similar devices and/or measures were utilized in the design of the Fermi reactor, as well as in Atomics International's design studies of the 500 MWe LMFBR demonstration plant.

The staff concludes that feasible courses of action are available that can be implemented to suitably mitigate the consequences of a spectrum of accidents beyond the design basis and provide the required containment system protection.

II-19

I TABLE I SUBSYSTEMS OF THE CONTAINMENT / CONFINEMENT SYSTEM ENGINEERED SAFETY FEATURES FOR DESIGN BASIS EVENTS (II.D.3)

DESCRIPTION FUNCTION Containment 1-1/2 inch steel, Primary Shell 3.7 x 106 cu. ft. containment barrier 10 psig Vacuum Relief Open at 3.5-inch Provide in-leakage Close at 1.75 inch for vacuum conditions Containment Automatically operated Close penetrations upon Isolation valves sensing radioactivity in containment Annulus Filter Two exhaust fans with Reduce radioactivity System heaters, prefilters, high release from annulus to efficiency particulate air environment while main-filters, adsorber bed and taining reduced pressure j after-filter in annulus PHTS Cells Inerted, lined, withstand Enclose & isolate 30 psig primary sodium system components l

t l

1 II-20

1 l

l t

III. GEOGRAPHY AND DEMOGRAPHY OF SITE ENVIRONS III.A. Site Description and Exclusion Area Control The proposed CRBRP site is located in Roane County in east-central Tennessee )

approximately 25 miles west of Knoxville and within the city limits of Oak l Ridge, Tennessee. The CRBRP site consists of approximately 1,364 land acres on a peninsula formed by a meander in the Clinch River. The site is bounded on the east, south, and west by the Clinch River and on the north by DOE's Oak Ridge reservation. The site is shown on a general map of the region in Figure 1 and on a local area map in Figure 2.

The topography of the site area is characterized by a series of parallel ridges generally oriented in a northeast to southwest direction. Chestnut Ridge, running across the northern part of the site, is the dominant topographic feature, reaching an elevation of 1100 feet above mean sea level (MSL). The grade elevation in the southern part of the site where the proposed plant structures will be located is 815 feet above MSL.

The proposed exclusion area will include the site property and the river adjacent to the site, less 112 acres along the northern boundary which have been set aside for an industrial park. A map of the site showing the exclusion area boundary lines, the site property limits, and the principal plant structures l is shown in Figure 3. The minimum exclusion area boundary distance is approxi- {

mately 670 meters (2,200 feet) measured from the center of the containment building southwest to the nearest point on the exclusion area boundary.

The site property is owned by the United States of America and is presently in the custody of the TVA. TVA will transfer to DOE the custody of those portions of the site which are required for the purpose of designing, constructing, and operating the CRBRP.

The proposed exclusion area will not be traversed by any public highways or railroads; however, the Clinch River along the eastern, southern, and western boundary is included within the exclusion area. Movement on the Clinch River will be controlled in the event of an emergency by the applicants in coordination with other appropriate agencies as specified in the radiological emergency plan.

The river bank on the plant site will be posted to inform river users of the nearby nuclear plant. A small family cemetery is located in the southern part of the site. Access to this cemetery will be controlled by the applicants.

The staff concludes that the applicants have the proper authority to determine all activities within the exclusion area, as required by 10 CFR Part 100, based on the applicant's custody of the site property and the commitment to make arrangements to control traffic on the Clinch River in the event of an emergency.

III.B. Population Distribution Approximately 4,440 people resided within 5 miles of the Clinch River site in 1980. This r epresents an increase of 1,700 persons in this area since the 1970 III-1

I census. Kingston, Tennessee, located 7 miles away in the west direction, is the largest nearby town and had a 1980 population of 4,367. Other major nearby communities are Oak Ridge, Tennessee (1980 population 27,522) located 9 miles northwest and Knoxville, Tennessee (1980 population 182,249) located 22 miles east-northeast of the reactor site. Table III shows the current and projected residential populations within 30 miles of the site. Approximately one-third of the area within 5 miles of the site is comprised of land owned by the U.S.

l' Government and in custody of either TVA or DOE. Although the Clinch River site is within the city limits of Oak Ridge, the residential area is located between seven and fourteen miles northeast of the site.

Transient population in the site vicinity, other than travelers on local roads and highways, consist primarily of workers at three large industrial activities on the Oak Ridge reservation. There are approxirntitely 5,600 employees at the Oak Ridge Gaseous Diffusion Plant, located about 3 miles north-northwest of the site, 5,000 employees at the Oak Ridge National i.aboratory located about 4 miles east-northeast of the site, and 6,300 employees at the Y-12 plant located about 9 miles northeast of the site. Recreational facilities within 10 miles of the site consist primarily of numerous small camping and water access areas. The nearest recreational activities of significance are a 100-unit commercial camping site about one mile southeast of the site and a stock car track about 2 miles east. The applicants estimate that the peak hour use of the recreational facilities within 10 miles of the site totals about 10,000 persons, based on 1980 information, and is projected to increase to about 11,000 by 1990. Over 50% of these recrea-tional visitors are attributed to spectators at the stock car track.

l The staff has compared the projected population in the CRBR site vicinity with

the acceptance criteria given in Regulatory Guide 4.7, " General Site Suitability Criteria for Nuclear Power Stations," and Standard. Review Plan Section 2.1.3.

The resident plus weighted transient population density within 30 miles of the site at projected time of plant startup (taken to be year 1990) was compared ,

with a density of 500 persons per square mile. Figure 7 shows the 1990 popula- j tion as a function of distance compared with a site having a uniform density of 500 persons per square mile.. Similarly, the resident plus weighted transient j

population density within 30 miles of the site at projected end-of plant-life 1 (taken to be year 2030) was compared with 1000 persons'per square mile. Figure 8 shows the projected 2030 population as a function of distance compared with a site having a uniform density of 1000 persons per square mile. From the figures it can be seen that the population density in the vicinity of the CRBR site is well within 500 persons per square mile at time of' plant startup and well within 1000 persons per square mile at end-of-life. The weighted transient population within 10 miles was estimated by the staff to be about 7,000 persons. This was added to the 52,000 residents in generating Figures 7 and 8. The majority of the transients are due to employees and local school population. Since a significant number are probably also residents within 10 miles, the staff estimates that simply adding the weighted transients to the residents introduces some double I

counting and that is, therefore, conservative. The weighting factors used were

based on annual average occupancy factors for the various groups.

In order to verify the applicants' population data, the staff obtained an independent estimate of the 1980 population within 50 miles of the site from  !

U.S. Bureau of the Census data and compared this to the applicants' 50-mile i population value for 1980. The staff found that the U.S. Bureau of Census j III-2

TABLE III

~ss 1980 CENSUS AND PROJECTED RESIDENT CUMULATIVE POPULATIch5 Radius, Miles 1980 1990 203GN _

0-5 4,440 4,680 5,380 )  %

0-10 52,040 57,980 67,580 'N. .

0-20 205,340 202,580 214,280 -

0-30 516,540 550,18(i 608,280 m s value of 837,300 was in good agreement with the applicants' valut of 830,'800.

The staff also compared the applicants' projected population ' growth rate for the year 2030 for the area within 50 miles of the ,

projections of the U.S. Bureau of Economic Analysis (site9mic' for Econ to tos' Areapopulation 50,' an ,

area comprising east-central Tennessee snd sottheastirn Kentucky. This cam-parison showed that the applicants' population grchth projdction o'r 2.5% per-l decade for the area within 50 miles of the site is less th3n'the the regional growth projection of 5.6% per decade for E:onomic Area 50 made by the U.S.

Bureau of Economic Analysis. In additioa, the Bureau of Economic Analysis projects the populat'on to be below the trip levels of Regulatory Cuide'4.7.

The applicants have specified a low population tone with an outer b6undary (

distance of 4,025 meters (2.5 miles) measured from the centeridf,the propored reactor location. Approximately one-third of this area consists ofiland within ,

the Oak Ridge reservation and the remainder is chardcterized br 1oy density rural development with no large concentrations of populatio'n. ~Baied-on data presented by the applicants, the staff estimates that less,than 1500 people yesided within 3 miles in 1980 and the applicants project virtually no change in,the LPZ popula-tion over the lifetime of the plant. Based on its' review to det4 4f the proposed LPZ with respect to such considerations as access' routes.\ populatu n distribution, and land use, the staff concludes that the protection of' the publk! health and safety can be reasonably assured by the applicatio' r' of advarice plan'ning measures, early warning capabilities, protection equipmen 6 and approprite protective features for those facilities which cannot be evacuated due to process and/or security requirements. The staff's conclusion is predicated oh the ass wption that a plant of the general size and type as the propesed CRBRP plaat can be designed to mitigate the most serious design basis accident consequences within the dose limitation prescribed in 10 CFR Part 100.  !

The nearest population center, as defined in 10 CFR Part 100, is Oak Ridge, Tennessee which had a population of 27,552 in 1980. The population center distance, based on the actual population distribution, is 7 miles in'the. north- I northeast direction. This distance is greater than the minimum population center  !

distance of one and one-third times the distance from the. proposed reactor i location to the outer boundary of the low populatidn zone, as required by 10 CFR Part 100. The applicants state that future residential development of Oak Ridge will not result in population growth closer than 5 miles from the site due to present zoning restrictions.

III-3

/,

10 J8III88i' CLINCH RIVER BREEDER REACTOR PLANT

~

_ s'

/

/

/

/

/

10 6 /

/ #00 #

/ f 0pg #

/ 4,fge

/

5 /

5 /

8 /

I 105 /

$ 5 / /

I

~

o _- /

/ 7. --YEAR 1990 POPULATION

_ /

I 4 I 10 9 -l 8 -l 7

6 5

-l 4 -

3 -f 2 d I

r 3 ' '

10 10 20 30 40 50 (MILES)

DISTANCE FROM PLANT, MILES Figure 7 CUMULATIVE POPULATION DISTRIBUTION (1990)

III-4

jI' CLINCH RIVER BREEDER REACTOR PLANT ,

- /

- I, '

/

/

/

/

/

_ /

/s

/' C 1000 PEOPLE / MILE2 10 6 /

~

/

/

- /

f

_ /

5 /

p -

/

5 e /

2 s 's /

/

/

~

g - f g

o

/

j

/  %

j IO30 Popq I

10 4 I a

7 N

6 N 5

1 1

3 -4 2 -

10 3 IIIIIIII1 10 20 30 40 50 (MILES)

DISTANCE FROM PLANT, MILES Figure 8 CUMULATIVE POPULATION DISTRIBUTION '(2030)

III-5

i i

The staff concludes that the nearby industrial, transportation, and military activities in the vicinity have been properly identified and the plant is adequately protected and can be operated with an acceptable degree of safety.

I III.C. Nearby Industrial, Transportation, and Military Facilities q Nearby industrial facilities to the proposed Clinch River site consist primarily '

of the nuclear-related facilities on the Oak Ridge reservation. The Oak Ridge Gaseous Diffusion Plant (0RGDP), located about three miles north-northwest of the site, produces enriched uranium. Anhydrous hydrofluoric acid (AHF) has been identified as a hazardous material stored at ORGDP whose accidental release could possibly impact on the safe operation of a nuclear plant at the Clinch River site by affecting plant operators in the control room. To evaluate this accident, it was postulated that an AHF storage tank failed and 2,000 pounds of the AHF evolved as a gas over a 15-minute period. Using conservative assumptions regarding meteorology and not taking credit for the buffer effect of the intervening ridges between the ORGDP release point and the Clinch River site, the concentration of hydrogen fluoride (HF) gas at the site was determined. The analysis indicated that based on the assumptions noted above, the concentration of HF gas at the Clinch River site could exceed the recommended exposure limits.

The applicants have committed to evaluate the habitability of the control room in accordance with the guidelines contained in Regulatory Guide 1.78, "Assump-tions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release" and, further, to install HF detectors in the control room air intakes which will alarm and automatically isolate the control room upon detection of the HF gas. The applicants also state that the transit time of the gas between the release point and the site would be of sufficient length to allow for communication between ORGDP and CRBRP and subsequent isolation of the control room. Communications procedures between ORGDP and CRBRP are to be included in the site emergency plan. The staff has reviewed the applicants' analysis and conclude tnat the potential accidental release of a large quantity of HF gas at ORGDP will not preclude the accept-ability of the Clinch River site on the basis that the installation of HF detectors in the control room air intakes and adequate communication procedures will assure the timely isolation of the control room. The staff will review the control room design during the course of the radiological safety review and will report its conclusion regarding the adequacy of the habitability systems in the Safety Evaluation Report.

The Oak Ridge National Laboratory (ORNL) is located about four miles east-northeast of the site. The approximately 5,000 employees at ORNL are engaged in basic and applied research in activities in nuclear and other technologies.

The Y-12 Plant, which employs about 6,300 persons, is located nine miles northeast of the site. Production and research and development facilities are provided at Y-12 for D0E. No activities have been identified at either ORNL or Y-12 which constitute a hazard to the safe operation of a nuclear plant at the Clinch River site.

One small industrial facility is located on a 33-acre tract in the 112-acre Clinch River Consolidated Industrial Park (CRCIP) along the northern boundary of the site approximately 1.5 miles from the proposed location of the plant structures. This industry, employing 30 people, fabricates neutron absorbers III-6

for power reactors and fuel elements for test reactors. This activity is considered to be compatible with the development of the Clinch River site for a j nuclear plant.

The major transportation artery in the vicinity of the site is Interstate 40 which passes approximately 1.25 miles to the south. State Route 58 is about 1.5 miles to the northwest and State Route 95 about three miles east at their closest points of approach. Hazardous materials for the nearby ORNL and ORGDP facilities are transported over these highways. An accident involving a tank truck carrying AHF has been postulated for evaluation purposes. The instal-lation of HF detectors in the control room air intakes, and the distances of these routes from the site, ensure that highway accidents involving AHF will not preclude the suitability of the site. During the course of the radiological safety review, the staff will evaluate what other hazardous materials, if any, are transported over these highways and require consideration in the design of the CRBRP control room. Based on the staff's previous review experience, it concludes that appropriate detectors can be provided, as required, in the control room design to safely mitigate accidental releases, and that such accidents, therefore, will not preclude the suitability of the site.

The closest major rail line is approximately 10 miles northwest of the site.

This distance is sufficient to eliminate potential railroad accidents as a factor in determining the suitability of the site.

There is some commercial barge traffic on the Clinch River past the site. Lock records at the Melton Hill Dam, approximately 5 miles upstream from the site, indicate that over a ten year period (1966-1975) an average of 4 barges per year went through the locks. The barges carried primarily steel products. The applicants state that no explosive, toxic, or hazardous materials have been transported by barge past the site. There is a potential for increased barge traffic due to the proposed construction of coal barge loading facilities above the site. However, no hazardous materials are expected to be included in this increased barge traffic. The staff concludes that postulated barge hazardous material accidents need not be considered in the design of the plant. In addition, barge collision accidents with the cooling water intake structure will r e affect the ability of the plant to operate safely since water intake from the Clinch River is not essential for a safe shutdown.

The closest airports to the site are two light plane facilities located at a distance of about 10 miles from the site. McGhee-Tyson (Knoxville), located 28 miles east-southeast of the site, is the closest major airport with scheduled commercial flights. The nearest flight path is V16 between Knoxville and Clinch Mountain which passes about 10 miles south of the site. We conclude that the distances of these aviation facilities are adequate to ensure that they will not adversely affect the suitability of the site.

The nearest fuel supply pipeline is a 6-inch natural gas pipeline which runs in a north-south direction and passes about one and one-third miles east of the proposed location of the plant structures. Based on the relatively small size the pipeline and its distance from the site, the staff concludes that this pipe-line will not preclude the acceptability of the site even if in the future a more hazardous gas such as propane were added to the natural gas in the pipeline.

III-7

The applicants state that there are no oil refineries or storage facilities, quarries, or mineral extraction operations in the vicinity of the site. There are no military bases or facilities within 10 miles of the site.

In order to evaluate the potential impact on the Clinch River site of the possible future expansion of existing facilities and the development of new DDE programs, the applicants requested their Oak Ridge Operations Office to survey the present activities and to establish the basis for a lone ange land-use plan for the Oak Ridge reservation. The results of this " :.ey, and DOE site selection and impact evaluation requirements, provide reasonable assurance that potential new activities on DOE controlled land will not impose an undue risk on the safe operation or the CRBRP.

In addition to the projected DOE programs, the Exxon Nuclear Company had requested a 2,500-acre site on the Oak Ridge reservation for storing and reprocessing spent fuel and had submitted an application to the Commission to construct this facility. The Exxon site was to be located approximately 2.5 miles north northeast of the Clinch River site. Since our original SSR was issued, plans for this facility have been terminated and the application was withdrawn.

The nature and extent of potential hazards resulting from man-related activities which are conducted at nearby industrial, military, and transporation facilities have been evaluated to determine if such activities have the potential for adversely affecting the suitability of the site for a nuclear plant. Based on its evaluation of information contained in the PSAR, as well as information independently obtained by the staff, the staff concludes that the activities in the vicinity of the Clinch River site are not likely to preclude the suitability of the site.

III.D. Site Suitability Source Term Dose Consequences The applicants have analyzed a spectrum of accidents for the CRBRP consistent with the Commission's defense-in-depth approach to safety. This concept is based on three levels of safety: cesigning a plant to conservative standards so that it will be safe in all phases of normal operations, protection against the consequences of anticipated malfunctions over the lifetime of the plant, and design features to protect against a series of postulated highly unlikely events. Included among the accidents analyzed are inadvertent reactivity insertions, a steam or feed line pipe brake, sodium-water reactions as the result of steam generator tube failures, PHTS and IHTS pipe leaks, fuel handling accidents, sodium fires, rupture of the RAPS surge vessel, and a failure in the liquid radwaste system.

In order to evaluate the effectiveness of the engineered safety features proposed for the CRBRP and the suitability of the Clinch River site to meet the exposure guidelines of 10 CFR Part 100, the staff has used a radiological source term directly analogous to the source term used to evaluate sites for LWRs. The radiological source term for the CRBRP consists of the usual LWR source term assumed to be released from the core, plus 1% of the plutonium in the core. The site suitability source term (SSST) for the CRBRP release from the core thus consists of 100% of the noble gases, 50% of the halogens (half of the iodines are assumed to plate out within a short time period), 1% of the solid fission products, and 1% of the plutonium. This activity is assumed to III-8 r

/

be uniformly mixed throughout the primary containment, and instantaneously available for release to the environment through engineered safety features such as containment structures and filter systems at the initiaticn of the accident.

As with LWRs, the SSST for the CRBRP is a nonmechanistic term and its use is intended to represent an assumed release from the core whose consequences would result in potential hazards not exceeded by those from any accident considered credible, in a manner comparable to the determination of the design basis loss-of-coolant accident doses for a LWR. A primary objective of the radiological safety review is to assure that no other accident sequences within the design envelope result in the release of fission products to the environment greater than those postulated for the site suitability source term.

Based on our eview and independent assessment of the information submitted by the applicants to date, the staff's preliminary finding is that the postulated design basis accidents can be mitigated through the installation of additional engineering safety features or by setting appropriate limits in the technical specifications so that the core involvement will be minimal and, hence, the amount of fission products available for release will not exceed the site suit-ability source term. However, the staff has identified several postulated design basis accident scenarios, such as a secondary system pipe break and a primary system cold trap fire, where it believes that it may be more appropriate to use more conservative assumptions than those used by the applicants. I The dose guidelines specified by the staff to evaluate the consequences of the postulated site suitability source term release to an assumed individual at the exclusion area boundary and outer boundary of the low population zone are those specified in 10 CFR Part 100 (300 rem thyroid and 25 rem whole body) with the following additional guidelines for potentially critical organs; 75 rem for the lung, and 300 rem for bone surfaces; coupled with the additional guideline that the mortality risk equivalent whole body dose from any postulated design basis accident (on a calculated dose basis) for the CRBR should be no greater than the mortality risk equivalent whole body dose value of 10 CFR 100 for an LWR (i.e., 34 rem whole body risk equivalent at the operating license stage, and 24.5 rem whole body risk equivalent at the construction permit stage). The additional dose guidelines, which are not specifically addressed in 10 CFR Part 100, are the equivalent mortality risk organ doses corresponding to a thyroid dose of 300 rem. The equivalency of the additional organ dose guide-lines to the thyroid guideline value and the mortality risk equivalent whole body dose guideline value were determined using the stochastic weighting factors in International Committee on Radiation Protection Publication 26. As in 10 CFR Part 100, these dose guidelines as set forth in these siting criteria are not intended to imply that these numbers constitute acceptable limits for emergency doses to the public under ace.ident conditions. Rather, these guidelines have been set forth as reference values which can be used in evaluating reactor sites with respect to potential reactor accidents of exceeding low probability of occurrence and low risk of public exposure. The dose guidelines specified by the staff for use during the construction permit review are 150 rem thyroid, 20 rem whole body, 35 rem to the lung, and 150 rem to bone surfaces. The use of lower values at the CP stage are considered appropriate at this stage of the review to account for the uncertainties noted in Section B of Regulatory Guides 1.3, " Assumptions Used for Evaluating the Potential Radiological Con-sequences of a Loss-of-Coolant Accident for Boiling Water Reactors," and 1.4, III-9

" Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," and is consistent with standard staff practice in LWR licensing reviews.

To mitigate the radiological consequences of the SSST, the applicants have committed to provide a containment system, which includes an Annulus Filtration System (AFS), as an engineered safety feature. Accidents beyond those con-sidered within the design bases have also been considered. These events are referred to as CDA, and may be characterized as resulting in releases of sub-stantial amounts of radioactivity from the reactor core due to %1 melting i and failure of engineered safety features. Because of the large consequences of such events, an objective of the staff's safety review is to require that the probability of occurrence of such events is so small that they can reasonably be excluded from consideration in the design of mitigation features.

The discussion of such "beyond the design bases events" is contained in the staff's FES.

The staff has incorporated these engineered safety features into its SSST dose model and, using conservative assumptions regarding their operation and effect-iveness in conjunction with meteorological dispersion values developed from onsite data, the staff has computed the doses from the postulated source term.

The assumption used in the analysis, and the resulting doses are shown in Table IV. The calculated doses indicate that the exposure guidelines, appropriate at .

the construction permit review stage, can be met for the Clinch River site. l During the course of the radiological safety review, the staff will require that i the ESFs conform to the NRC's rules and regulations, and that the radiological  !

consequences of the SSST not exceed the appropriate exposure guidelines. The I staff will also require that the CRBRP design be such that no design basis accident has consequences which exceed those of the SSST. The staff concludes that there is reasonable assurance that suitable engineered safety features will be provided for CRBRP such that the radiological consequences of postu-lated design basis accidents will be within the values of 10 CFR Part 100.

III.E Emergency Planning i The applicants have provided a description of the preliminary plans for coping l with emergencies. The staff has completed its initial review of the plans i against the requirements of 10 CFR 50, Appendix E, Part II. The staff I recently requested additional clarification of the applicants' plans.

The Federal Emergency Management Agency (FEtM), in its review of state and local plans for the nearby Sequoyah Nuclear Plant, found that the State of Tennessee Radiological Emergency Plans are adequate and capable of being fully implemented. FEl% will review the state and local plans for the emergency planning zones for the Clinch River site during the CRBRP operating license review for compliance to the criteria specified in NUREG-0654/ FEMA-REP-1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear ?ower Plants." The staff concludes that an effectively coordinated site state and local radiological emergency response plan can be achieved for 9e Clinch River site. l 1

l III-10

e i

=

Table IV Site suitability source term assumptions and dose results Power Level 1121 MWt Core Fraction Released to Containment:

Noble Gases 100%

Iodines 50%

Solid Fission Products 1%

Plutonium 1%

Primary Containment Free Volume 3.7 x 106 fta Primary Containment Leak Rate 0.1%/ day Bypass Fraction 0.001%/ day Annulus Filtration System Filter Efficiencies:

Particulate Iodine, Solids and Plutonium 99%

Elemental and Organic Iodine 95%

Annulus Filtration System Flow Rates, cfm:

Exhaust 3,000 Recirculation 11,000 Aerosol Fallout Coefficients in Containment, hr 2: l 0-2 hours 0.0853 2-8 hours 0.0659 8-24 hours 0.0571 Minimum Exclusion Area Boundary Distance 670 meters low Population Zone 4023 meters Atmospheric Dispersion Parameters (5% meteorology), sec/m3 :

0-2 hours at exlusion area boundary 1.22 x 10 3 0-8 hours at LPZ l.2 x 1G 4 8-24 hours at LPZ 8.4 x 10 5 24-96 hours at LPZ 3.9 x 10 5 96-720 hours at LPZ 1.4 x 10 5 Dose Consequences, rem Exclusion Low Popula-Area tion Zone Thyroid 12 7 Whole Body 0.6 0.3 Lung 0.4 0.4 Bone 10 9 III-11

l l

l IV. PHYSICAL SITE CHARACTERISTICS IV.A. Meteorology The Clinch River site is located in a broad valley of the southern Appalachian Mountains which is characterized by a series of parallel ridges, separated by long, narrow valleys extending in a northeast-southwest direction. The orienta-tion of the valleys and ridges, and differences in elevation have a measurable effect on the climate of a given site. In general, this region of the contiguous United States encounters atmospheric dispersion conditions which are less favorable than average for all areas of the country.

A description of the meterological conditions of the site, including the climatology of the region, local meterological conditions, and expected severe weather is presented in Section 2.6 of the staff's FES. Section 6.3.1 of that document describes the onsite meteorological program. The onsite meteorological measurement system originally was not comparable to the recommendations of Regulatory Guide 1.23, "0nsite Meteorological Programs," with respect to the location of wind and vertical temperature gradient measuring instrumentation. l The system has been modified and the staff finds that it conforms to its recommendations.

The applicants have provided joint frequency distributions of wind speed and direction by atmospheric stability class (based on vertical temperature difference) collected on the Clinch River onsite meteorological tower during the one year period February 17, 1977 through February 16, 1978. From these data the staff calculated estimates of the relative concentration (X/Q) values for short-term releases from plant buildings and vents using the wind speed and direction measured at the 33 foot level and the vertical temperature difference l measured between the 33 and 200-foot levels on the tower.

In accordance with the methodology described in Regulatory Guide 1.145, short-term (up to 30 days) X/Q values were calculated. A direction dependent atmos-pheric dispersion model with enhanced lateral dispersion during neutral and stable atmospheric conditions accompanied by low wind speeds was used. These enhanced lateral dispersion factors were based upon diffusion studies performed at several locations including the Clinch River site.

Two probablistic analyses were performed. The first analysis requires the development of the X/Q values for each of the 16 cardinal point sectors that is not exceeded 0.5% of the total time. The highest of each of these 16 sector X/Q values is defined as the maximum section X/Q value is compared with the overall site X/Q that is exceeded no more than 5% of the total time. Whichever value was higher was used to determine the consequences of accidental releases at the exclusion zone boundary of 670 meters and outer boundary of the low population zone of 4023 meters. For the Clinch River site the more conservative X/Q values were those based upon the 0.5% sector values and was thus utilized by the staff to evaluate the consequences of design basis accidental releases.

IV-1

j N Althcugh the atmospheric diffusion conditions at the Clinch River site are less favorable than the conditions throughout most of the United States, its X/Q values are still comparable to those which the staff has calculated for several other nuclear power sites in the region.

All structures and equipment exposed to tornado forces and needed for the safe shutdown of the plant will be designed to be consistent with the design basis tornado characteristics for Region I as recommended by Regulatory Guide 1.76,

" Design Basis Tornado for Nuclear Power Plants."

The staff finds that the applicants have provided data which is reasonably representative of conditions at the proposed CRBRP site and is sufficient to conservatively estimate atmospheric dispersion characterisitics. On this basis, the staff concludes that the meteorology at the proposed site is sufficiently characterized and that there are no meteorological characteristics that would preclude the determination of site suitability in accordance with 10 CFR Part 100.11 (refer to Section III.D of this report).

IV.B Hydrology The proposed site is on the north shore of the Clinch River and about 25 miles west of Knoxville, Tennessee. Proposed plant grade will be about 815 ft above mean sea level (MSL), which is about 74 ft above the normal river level of 741 ft MSL. The Clinch River drainage area is about 16,200 square miles, and the average flow is about 4800 CFS at the site; the river is regulated by a series of dams, both upstream and downstream from the site. Th'e site is most directly under the influence of the Melton Hill dam which is about 5 miles upstream.

Cooling tower makeup will be withdrawn from the Clinch River. The staff has concluded that an adequate normal cooling water supply can be provided.

Emergency cooling for safe shutdown and residual heat removal will be supplied by a mechanical draft cooling tower, which will have a sufficient supply of water available in its self-contained storage basin, consistent with the criteria suggested in Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants."

The potential for flooding of the site from several sources has been considered by the applicants, and independently verified by the staff, consistent with the criteria of Regulatory Guide 1.59, " Design Basis Floods for Nuclear Power Plants."

The maximum flood on record, judging from gage records and newspaper accounts, occurred in March 1886, with a reported water level of about 764 ft MSL at the proposed site. This flood occurred before construction of the present, extensive TVA dam system. Since completion of the system of dams in March 1973, the maximum water level at the site has been about 750 ft MSL which is about 65 ft below plant grade. A repetition of the worst flood of record, but with the present TVA dam systam, would yield a water level of about 751 ft MSL, 64 ft below plant grade.

The app.licants have evaluated and the staff has independently verified the l precipitation induced Probable Maximum Flood on the Clinch River. The estimated maximum stillwater level is about 778 ft MSL, 37 ft below plant grade. Wind I

l I

IV-2

l l

wave runup would add a maximum of about 4 ft against vertical surfaces. These i flood levels were found not to be as severe as the Design Basis Flood. )

The Design Basis Flood for the proposed site has been determined by the applicants to be caused by the assumed partial seismic failure of Norris Dam, I about 62 miles upstream from the site, coincident with the Standard Project Flood with the attendant failures of the Melton Hill Dam and Watts Bar Dam.

The Standard Project Flood is about half that of the Probable Maximum Flood and is generally representative of the maximum historical flood in the region. The maximum stillwater level at the site has been estimated by the applicants to be about 804 ft MSL, about 9 ft below plant grade. Maximum wave runup would add an estimated 5 feet at vertical surfaces. The staff concurs with this evaluation.

The applicants have proposed that the site drainage facilities, including roofs, will be designed such that an occurrence of the local Probable Maximum Percipi-tation will not constitute a threat to safety-related facilities. These proposed design bases meet the criteria suggested in Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Report for Nuclear Power Plants," Revision 2.

Groundwatar occurs at the site primarily in weathered joints and fractures in the surface rock, under water table conditions. All groundwater at the site flows toward the river, which is the groundwater sink. There are no ground-water users which could be affected by the unplanned release of liquid radwaste.

Groundwater travel time to the Clinch River has been estimated by the applicants to be a minimum of 28 years. Due to sorption, most radionuclides would travel i more slowly than the groundwater.

1 Based on independent engineering evaluation, and comparisons of hydrologic '

parameters at the site with those at other nuclear power reactor sites, the staff concludes that there are no unique hydrological phenomena related to site flooding, that an adequate water supply can be provided for normal and emergency i cooling, and that the hydrosphere offers no greater potential for surface water and groundwater contamination from unplanned releases of liquid radwaste than at other nuclear power reactor sites.

IV.C. Geology and Seismology IV.C.1 Geology The proposed site is located in the southeast section of the Valley and Ridge Physiographic Province of eastern Tennessee. Surface rocks at the site consist of two major geologic units, the Knox Group and Chickamauga Group. The former is predominantly a dolomite of Cambro-Ordovician age. The Chickamauga is the foundation rock for the site and consists of alternating layers and laminations of silstone, limestone, and shale with some chert. The bedrock is included in the zone of extensive thrust faulting in east Tennessee. The bedrock contains minor structures such as small faults (a few feet in length) and small folds.

The strike is approximately N45 E and dips on the average about 40 southeast.

The bedrock is overlain in some areas by terrace deposits of up to 40 feet thick, weathered rock, and extensive zones of clayey residual ceil. The overburden thickness ranges from 8 to 56 feet deep over the site area. Most of the plant island is founded on the Chickamauga Unit A limestone and Unit A upper siltstone IV-3

i which do not have significant weathering except near the ground surface.

Weathering and solutioning of the Unit B limestone in the site area appears to extend a maximum depth of about 100 feet primarily along jointing.

The foundation level of the plant island is about 15 to 80 feet below the top i of continuous rock which is defined in the PSAR as rock which does not contain any significant weathered or solutioned discontinuities.

Nearby Tectonic Structures ,

The closest major fault is the Copper Creek fault and its trace is located 3,000 feet from the site. At this location, the fault strikes N52 E and dips 4 away from the site to the southeast at an approximate dip of 25 degrees.

Displacement of this fault is about 7,200 feet with the Rome Formation thrust over Chickmauga Group rocks. This fault has a mapped length of 100 miles, but becomes complex and merges to the north with other faults. The Copper Creek

. fault is one of many late Palezoic thrusts that developed during the Allegheny Orogency (Pennsylvanian-Permian, 330-240 million years before present, (MYBP).

These structures are not considered active and are not used in determination of

, the Safe Shutdown Earthquake. Radiometric dates of 290 10 MYBP and 280 t MYBP were obtained for mylonite fault gage material taken from the fault zone

of the Copper Creek thrust. This finding, coupled with lack of evidence of recent offset and an understanding of the tectonic development of the Paleozoic thrust faulting in east Tennessee, indicates that this major fault and other small

. faults in the site area associated with it.are tectonically old. Therefore, the staff does not consider them hazardous to the. safe operation of a nuclear plant at this location. These faults are not capable faults as defined in

" Seismic and Geologic Siting Criteria for Nuclear Power Plants," Appendix A, 10 CFR Part 100.

Considerable new regional geologic and seismic information has been obtained i since publication of the SSR, including new data regarding the Giles County i and Charleston earthquakes and theories about their source mechanisms. The applicants are assessing this new information relative to the proposed CRBRP

site. The staff has been following the development of new information and to date finds no reason to change our conclusions regarding the suitability of the site.

I Man-Made Conditions

{

A facility for injecting radioactive waste into subsurface strata is located on the Oak Ridge Reservation approximately four miles east of the proposed CRBRP.

l site. These injection wells have been used periodically since February 1954 to inject wastes mixed with a cement grout slurry into the Conasauga shale along cracks generated by hydrofracturing. Thus far, injections have been.into units l stratigraphically above the projection of the Copper Creek thrust fault.

j Many small seismic signals resembling earthquakes have been recorded on the i seismograph at Oak Ridge. These occur primarily during working hours. Moreover, they do not seem to occur any more frequently when injection is in progress.

The applicant has conducted analyses of these signals and has compared these results with the data obtained from Rangely, Colorado where earthquakes have occurred due to man-made causes. The applicant has concluded that the ORNL IV-4 l

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

injection wells are not inducing seismicity in the area; the staff also concurs in this assessment on the basis of a statistical comparison. The applicant has committed to restrict future hydrofracture operations within a defined set of parameters.

A new disposal well has recently been installed and will be utilized beginning June 1982. This new facility is located about 800 feet southwest of the well that has been used during the last few years. That well has been retired. Tests at the new location have demonstrated that the new disposal well penetrates essentially the same geologic horizon as the old well. The new well will be closely monitored using technique that have proved successful in the past. The staff therefore concludes that future waste injection will not have an adverse affect on the proposed CRBRP site.

Summary Based on the staff's review, this site has no geologic problems which are not amenable to established engineering solutions and is, therefore, a suitable location for the CRBRP.

IV.C.2 Seismology In arriving at the Safe Shutdown Earthquake (SSE), the proposed CRBRP site has been considered to be located in the Southern Valley and Ridge Tectonic Province.

The epicentral intensity of the maximum historical earthquake which has occurred in the province in which the proposed CRBRP site is located has been the subject of a reevaluation by the U.S. Geological Survey (Letter to Edson G. Case, USNRC from W. A. Radlinski, Acting Director, U.S. Geological Survey, February 12, 1976).

The conclusion of the reassessment of the maximum intensity of the Giles County, Virginia earthquake of May 31, 1897 was that, "Following past practice, there is no basis for revising the assigned maximum intensity of MM VIII." Following the tectonic province approach described in " Seismic and Geologic Siting Criteria for Nuclear Power Plants," Appendix A, 10 CFR Part 100, it is assumed that the  !

intensity at the proposed CRBRP site due to other Safe Shutdown Earthquake could  !

equal intensity MM VIII. Plots of measured peak ground acceleration values versus I observed intensity show a large variation.

Several authors have reported curves or correlations in the literature which in one way or another attempt to represent these data. The most frequently used curves are least squares lines which relate the logarithm of mean acceleration to intensity. Because the samples are varied from one study to another, the derived relationships have varied as well. The staff practice is to choose l

values which are representative of the trend of the mean of the data for the intensity of the SSE. On this basis, the staff considers a value of 0.25g to be appropriate for the SSE at the CRBRP site. The staff recognizes that the correlations relied upon in its assessment have been derived using data recorded in active seismic zones, primarily California. The staff is independently reviewing available strong motion data in an attempt to better identify the parameters affecting the vibratory motion earthquake size correlation and to assess any geographical dependence. Based on the preliminary results of these studies, the value of CRBRP is judged to be adequately conservative.

IV-5

In accordance with 10 CFR 100, Appendix A, the SSE is defined as the design response spectra. In the zero period limit, these spectra are normalized to the acceleration for seismic design corresponding to the design earthquake.

The seismic design response spectra for CRBRP will be reviewed against the existing staff positions and Regulatory Guides to assure that the seismic input, as defined by the design response spectra corresponding to the specified ground acceleration is acceptable.

Summary From its analysis and evaluation of available seismological data, including the results of investigations performed by the applicants' the staff concludes that there are no corresponding considerai. ions that would preclude the acceptability of the site for a nuclear power plant.

IV.D. Foundation Engineering The foundation rock for the proposed CRBR site consists of alternating layers and laminations of siltstone, limestone and shale with some chert. The bedrock is overlain by terrace deposits up to 40 feet thick, weathered rock, and zones of clayey residual soil. Overburden varies in thickness from 8 to 56 feet throughout the site area. The main seismic Category I structures, except for the Steam Generator Maintenance Bay, the Fuel Oil Storage Tanks, and the Cooling Tower will be founded on a single common structural mat called the Nuclear Island at elevation 715 feet located directly on a siltstone structure termed the "Chickamauga Unit A Upper Siltstone." The Steam Generator Maintenance Bay will be founded in a limestone formation termed the "Chickamauga Unit B Limestone." Two seismic Category I Fuel Oil Storage Tanks will be anchored to a common reinforced concrete mat with base at elevation 787 feet supported directly by compacted Class A structural backfill material overlying the Unit A Upper Sil stone. The seismic Category I Cooling Tower will be supported by a single mat founded at elevation 765 feet on the Unit A Upper Siltstone. Emergency plant and underground class IE electrical ducting will be founded on compacted Class A structural backfill materials.

The applicants have reported a total of 129 borings and 6360 linear feet of seismic refraction traverses have been accomplished to determine subsurface conditions at the site. Additional in situ testing accomplished included seismic up-hole surveys, seismic cross-hole surveys, continuous velocity logging and Goodman Jack testing. Laboratory testing of representative samples of the sub-surface rock has been accomplished to determine the static and dynamic properties of the materials. Other site investigative efforts reported as accomplished relevant to the geotechnical aspects of the site involved a comprehensive office review of available published data including reports, geologic maps, and previous construction data for the area.

Based upon the information presented by the applicants, it is the finding of the staff that the applicants' site investigation efforts provide adequate coverage of the site area in sufficient detail to provide a high level of confidence that specific subsurface conditions have been adequately defined.

The staff's review of the data presented reveals no evidence of significant zones of solutioning, caverns, or highly weathered areas in the foundation bedrock which could produce significant subsidence under the anticipated loads IV-6

to be imposed by the proposed structural mats. The staff, therefore, concludes that there are no subsurface conditions expected which would preclude the suitability of the site for the proposed plant.

IV-7

APPENDIX A CLINCH RIVER BREEDER REACTOR PLANT DESIGN CRITERIA ISSUED BY CLINCH RIVER BREEDER REACTOR PROGRAM OFFICE l OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION l

\

f CLINCH RIVER BREEDER REACTOR PLANT DESIGN CRITERIA CRITERION 1 - Quality Standards and Records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functons to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and com-ponents important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

CRITERION 2 - Design Bases for Protection Against Natural Phenomena.

Structures, systems, and components impostant to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect:

(1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

CRITERION 3 - Fire Protection. Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of structures, systems, and components.

CRITERION 4 - Protection Against Sodium Reactions. Systems, components and structures containing sodium shall be designed to limit the consequences of A-1

sodium chemical reactions resulting from a sodium spill. Special features such as inerted vaults shall be provided as appropriate for the reactor coolant system. Means to detect sodium or sodium reaction products and fire control (

systems shall be provided to limit and control the extent of such reactions to assure that the functions of components important to safety are maintained.

Means shall be provided to limit the release of sodium reaction products to the environment as necessary to protect plant personnel and to avoid undue risk to the public health and safety. Materials which might come in contact with sodium shall be chosen to minimize the adverse effects of possible chemical reactions. In areas where sodium chemical reactions are possible, structures, components, and systems important to safety, including electrical wiring and components, shall be designed and located so that the potential for damage by sodium chemical reactions is minimized. Means shall be provided as appropriate to minimize possible contacts between sodium and water. The effects of possible interactions between sodium and concrete shall be considered in the design.

The sodium-steam generator system shall be designed to detect sodium-water reactions and limit the ef#ects of the energy and reaction products released by such reactions so as to prevent loss of safety functions of the heat transport system.

CRITERION 5 - Environmental

• and Missile Design Bases. Structures, systems and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, anticipated operational concurrences, and postulated accidents. These structures, systems, and components shall be appropriately protected against dynamic effects such as the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

CRITERION 6 - Sharing of Structures, Systems, and Components. Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

CRITERION 7 - Sodium Heating Systems. Heating systems shall be provided as necessary for systems and components important to safety which contain, or may be required to contain, sodium. The heating systems and their controls shall be appropriately designed with suitable redundancy to assure that the temperature distribution and rate of change of temperature in sodium systems and components are maintained within design limits assuming a single failure.

The heating system shall be designed such that its failure will not impair the safety function of associated systems and components.

CRITERION 8 - Reactor Design. The reactor and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

• Natural phenomena are covered by Criterion 2.

A-2

l l

CRITERION 9 - Reactor Inherent Protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuc' ear feedback characteristics tends to compensate )

for a rapid increase in reactivity.

CRITERION 10 - Suppression of Reactor Power Oscillations. The reactor core and associated coolant, control, and protection system shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

CRITERION 11 - Instrumentation and Control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation for anticipated operational occurrences and for postulated accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

CRITERION 12 - Reactor Coolant Boundary. The reactor coolant boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

CRITERION 13 - Reactor Coolant System Design. The reactor coolant system and associated control, protection, auxiliary and sodium heating systems, snall be designed with sufficient margin to assure that the design conditions of the reactor coolant boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

CRITERION 14 - Containment Design. Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

CRITERION 15 - Electric Power Systems. An onsite electric power system and an offiste electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function function for each system (assuming the other system is functioning) shall be to provide sufficient capacity and capability to assure that (1) specified accept-able fuel design limits and design conditions of the reactor coolant boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled, and containment integrity and other vital functions are maintained in the event of postulated accidents.

The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the onsite electric distri-bution system shall be supplied by two physically independent circuits (not A-3

r -- - -

i i

necessarily on separate rights-of-way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under /

operating and postulated accident and environmental conditions. A switchyard (

common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following any postulated accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.

CRITERION 16 - Inspection and Testing of Electric Power Systems. Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full opera-tional sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.

CRITERION 17 - Control Room. A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions (including those conditions addressed in Criterion 4 - Protection Against Sodium Reactions). Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and with a design capability for subsequent control of the reactor at any coolant temperature lower than the hot shutdown condition.

CRITERION 18 - Protection System Functions. The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

A-4

)

CRITERION 19 - Protection System Reliability and Testability. The protection system shall be designed for high functional reliability and in-service test-I ability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation including a capability to test channels independently to determine failure and losses of redundancy that may have occurred.

CRITERION 20 - Protection System Independence. The protection system shall be designed to assure that the effects of natural phenomena and of normal operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demon-strated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

CRITERION 21 - Protection System Failure Modes. The protection system shall be designed to fail into a safe state or into a state such as disconnec-tion of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, sodium, sodium reaction products, and radiation) are experienced.

CRITERION 22 - Separation of Protection and Control Systems. The protec-tion system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

CRITERION 23 - Protection System Requirements for Reactivity Control Malfunctions. The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal of control rods.

CRITERION 24 - Reactivity Control System Redundancy and Capability. Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably control-ling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical for any coolant temperature lower than the hot shutdown condition.

A-5

CRITERION 25 - Combined Reactivity Control Systems Capability. The reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool (

the core is maintained.

CRITERION 26 - Heat Transport System Design. The heat transport system shall be designed to reliably remove heat from the reactor and transport the heat to the turbine generator or ultimate heat sinks under all plant conditions including normal operation, anticipated operational occurrences, and postulated accidents. Consideration shall be given to provisions of independence and diversity to provide adequate protection against common mode failures. The system safety functions shall be to:

(1) Provide sufficient cooling to prevent exceeding specified acceptable fuel design limits during normal operation and following anticipated operational occurrences, and (2) Provide sufficient cooling to prevent exceeding specified acceptable fuel damage limits and to maintain integrity of the reactor coolant boundary following postulated accidents.

Following the loss of a flow path, the heat transport system shall include at least two independent flow paths, each capable of performing the safety functions following shutdown.*

The system shall include suitable electric interconnections, leak detetections, isolation, and containment capability to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power systems operation (assuming onsite power is not available) the safety function can be accomplished, assuming a single failure.

CRITERION 27 - Assurance of Adequate Reactor Coolant Inventory. The reactor coolant boundary and associated components, control and protection systems shall be designed to limit loss of reactor coolant so that an inventory adequate to perform the safety functions of heat transport system is maintained under normal operation, anticipated operational occurrences and postulated accident conditions.

CRITERION 28 - Quality of Reactor Coolant Boundary. Components which are part of the reactor coolant boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

CRITERION 29 - Fracture Prevention of Reactor Coolant Boundary. The reactor coolant boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident

• This requirement is not intended to preclude two-loop operation provided the system safety functions can be appropriately met.

A-6

)

conditions (1) the boundary behaves in a nonbrittle manner and (2) the proba-

, bility of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures, service degradation of material properties, creep, and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of coolant chemistry and irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

CRITERION 30 - Inspection of Reactor Coolant Boundary. Components which are part of the reactor coolant boundary shall be designed to permit (1) periodic inspection and testing of areas and features important to safety, to assess their structural and leaktight integrity, and (2) an appropriate material surveillance p rogram.

CRITERION 31 - Intermediate Coolant System. The intermediate coolant system shall be designed to transport heat reliably from the reactor coolant system to the reactor residual heat extraction systems as required for the reactor coolant system to meet its safety functions under all plant conditions including normal operation, anticipated operational occurrences, and postulated accident conditions. The intermediate coolant system shall contain coolant that is not chemically reactive with the reactor coolant.

A pressure differential shall be maintained across a passive boundary between the reactor coolant system and the intermediate coolant system such that any leakage would flow from the intermediate coolant system to the reactor coolant system unless other provisions can be shown to be acceptable on some defined basis.

CRITERION 32 - Fracture Prevention of Intermediate Coolant Boundary. The intermediate coolant boundary shall be designed with sufficient margin to assure that when stressed under normal operation, anticipated operational occurrences and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.

The design shall reflect consideration of service temperatures, service degrada-tion of material properties, creep, and other conditions of the boundary material under normal operation, anticipated operational occurrences, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of coolant chemistry, (3) residual, steady-state and transient stresses, and (4) size of flaws.

CRITERION 33 - Inspection and Surveillance of Intermediate Coolant Boundary. Components which are part of the intermediate coolant boundary shall be designed to permit (1) periodic inspection of areas and features important to safety, to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the intermediate coolant boundary.

Means shall be provided for detecting intermediate coolant leakage.

CRITERION 34 - Reactor and Intermediate Coolant and Cover Gas Purity Control. Systems shall be provided to monitor and maintain reactor and inter-mediate coolant and cover gas purity within specified design limits. These limits shall be based on consideration of (1) chemical attack, (2) fouling and A-7

plugging of passages, (3) radioisotope concentrations, and (4) detection of sodium-water reactions.

CRITERION 35 - Reactor Residual Heat Extraction Systems. The reactor residual heat extraction systems shall be provided to transfer residual heat from the reactor heat transport systems to ultimate neat sinks under all plant shutdown conditions following normal operation, anticipated operational occurrences and postulated accident conditions. A passive boundary shall normally separate heat transport system coolant from the working fluids of the heat extraction systems. Suitable redundancy in components and features, and suitable inter-connections, leak detection, and isolation capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

CRITERION 36 - Inspection of Reactor Residual Heat Extraction Systems.

The reactor residual heat extraction systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of their components, (2) the operability and the performance of the active components of the systems, and (3) the operability of each complete system, and under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation for reactor shutdown and following postulated accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.

CRITERION 38 - Additional Cooling Systems. In addition to the heat rejection capability provided by the reactor residual heat extraction systems, systems to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided, as necessary. The system safety function shall be to transfer the combined heat load of these structures, systems, and components as required for safety under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

CRITERION 39 - Inspection of Additional Cooling Systems. The additional cooling systems shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the systems.

CRITERION 40 - Testing of Additional Cooling Systems. The additional cooling systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of their components, (2) the operability and the performance of the active components of the systems, and (3) the operability of the complete systems and, under A-8

conditions as close to design as practical, the performance of the full opera-tional sequence that brings the systems into operation, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.

CRITERION 41 - Containment Design Basis. The reactor containment structure, including access openings and penetrations, and if necessary, in conjunction with additional post accident heat removal systems including ex-vessel systems, shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate, and with sufficient margin, the calculated pressure and temperature conditions resulting from normal operation, anticipated operational occurrences and any of the postulated accidents. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as decay heat in released fission products, potential spray or aerosol formation, and potential exothermic chemical reactions; (2) the limited experience and experimental data available for defining accident phenomena and containment responses; and (3) the conservatism of the calculational model and input parameters.

CRITERION 42 - Fracture Prevention of Reactor Containment Boundary.

The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its metallic materials behave in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws.

CRITERION 43 - Capability for Containment Leakage Rate Testing. The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

CRITERION 44 - Provisions for Containment Testing and Inspection. The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows.

CRITERION 45 - Piping Systems Penetrating Containment. Piping systems penetrating reactor containment shall be provided with leak detection, isola-tion, and containment capabilities having redundancy, reliability, and perfor-mance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

CRITERION 46 - Reactor Coolant Boundary Penetrating Containment. Each line that is part of or directly connected to the reactor coolant boundary and A-9

that penetrates reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are accept-able on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment, or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment, or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment, or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appro-priateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

CRITERION 47 - Primary Containment Isolation. Each line that connects directly to the containment atmosphere and penetrates primary reactor contain-ment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment, or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment, or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment, or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

A-10

1 Isolation valves outside containment shall be located as close to the con-tainment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

CRITERION 48 - Closed System Penetrating Containment. Each line that penetrates primary reactor containment and is neither part of nor directly connected to the reactor coolant boundary, nor connected directly to the containment atmosphere shall have at least one containment isolation valve, unless it can be demonstrated that containment isolation provisions for a specific class of lines are acceptable on some other defined basis. The isolation valve, if required, shall be either automatic or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

CRITERION 49 - Containment Atmosphere Cleanup. Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. The necessity of such systems should consider the effects of sodium leakage and its potential reaction with oxygen and its potential for hydrogen generation when in contact with concrete.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capa-bilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.

CRITERION 50 - Inspection of Containment Atmosphere Cleanup Systems.

The containment atmosphere cleanup systems shall be designed to permit appro-priate periodic inspection of important components, such as filter frames, l ducts, and piping to assure the integrity and capability of the systems.

{

CRITERION 51 - Testing of Containment Atmosphere Cleanup Systems.

The containment atmosphere cleanup systems shall be designed to permit appro-priate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and i valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational i sequence that brings the systems into operation, including operation of applicable  !

portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems.

CRITERION 52 - Control of Releases of Radioactive Materials to the Environment. The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, A-11

4 including anticipated operational occurrences. Sufficient holdup,ca,6acity (i

shall be provided for retention of gaseous and liquid effluents contain#ng radioactive materials, particularly where unfavorable . site environmental ,,

conditions can be expected to impose unusual operational limitations upon the _

release of such effluents to the environment. ,

CRITERION 53 - Fuel Storage and Handling and Radioactivity Centrol. l The  !,

fuel storage and handling, radioactive waste, and other' systems which may contain radioactivity shall be designed to _ assure acequate safety under norcal s' and postulated accident conditions. These' systems shall be designed (1) with a,s capability to permit appropriate periodic inspection and. testing of components g important to safety, (2) with suitable shielding fi,7 radiation protection, (3) ., ~ ' .

with appropriate containment, confinement, and filtering, systems, (4) with a residual heat removal capability having ~' reliability ar.d testability that reflects the importance to safety of decay heat and other re'sidual heat removal, ,

and (5) to prevent significant reduction in fuel storage coolant inventory , .s under accident conditions.

. 1-CRITERION 54 - Prevention of Criticality in Fuel Storage' and Handli.gj.

s-Criticality in the fuel storage and handling systen shall De prevented by ,~ ,

physical systems or processes, preferably by use of geometrically safe, . <

configurations.

. , n. ,

CRITERION 55 - Monitoring Fuel and Waste Storage. s pprope.iates$ftcds 9 -

F shall be provided in fuel storage and radioactive waste syst6ms.and as,0ciated, ,

handling areas (1) to detect conditions that may. result in loss of!rer.idual "

heat removal capability and excessive radiation leve'Is and (2) to ' initiate' 4: 1 appropriate safety actions. s y- _

CRITERION 56 - Monitoring Radioactivity Releases. iMeans shall be provided i for monitoring the reactor containment atmosphere, effluent dischar'ge paths, \

/,

and the plant environs for radioactivity that r.ay be released frcm normal l operations, anticipated operational occurrences, and frori postulated \ accidents. ,

9 ,' ,

4 a

! 's N s ._

! .I ,

\ ,

s . >

• ~.

a s 4

( 1

. t h

a s

A-12

[RC r oRu 335 - U S NUCLE AR REGUL ATORY COMMISSION BIBLIOGRAPHIC DATA SHEET fiUREG-0786 4 TIT LE AN P 5UfiTIT LE (Ao1 volume No,,f avorcer,arrt 2 (L *me e/me)

Site Suitability Report in The Matter of The Clinch River Breeder Reactor Plant, Revision to March 4,1977 a RECIPIENT S ACCESSicw NO J port (Docket No. 50-537)

1. AU T Hoh tSi 5. DATE REPORT COMPLE TED

' Clinch River Breeder Reactor Program Office May l 52 9 PT:HF OHMING OHGANI/ATION N AME AND MAILING ADDRESS (inctu* I,a Code / DATE REPORT ISSUED Clinch River Breeder Reactor Program Office I"j"28 Office of Nuclear Reactor Regulation , g,,fgi, U.S. Nuclear Regulatory Commission Washington, D. C. 20555 a a ,,, .;

il SPONSOHING OHGANil A t lON N AME AND M AlltNG ADDRE SS I/nclu* I,o Co*1 p

Clinch River Breeder Reactor Program Office , , , , ,y go Office of Nuclear Reactor Regulation 13 T Y PE OF At POH T et RIOD COVE RE D (inclus<re dJes)

Site Suitability Report (Regulatory Report) March 1977 through June 1982 15 SUFPLE M N TAHY NO TE S ,4 (te ,y, of e g )

16 ABSTH AC T (200 words or less/

The Office of Nuclear Reactor Regulation issued a Site Suitability Report (SSR) for the proposed Clinch River Breeder Reactor Plant (CRBRP) in March 1977. That report documented the results of the staff's evaluation of the suitability of the proposed CRBRP site for a facility of the general size and type as the CRBRP from the standpoint of radiological health anJ safety considerations. The staff concluded in that report that the proposed CRBRP site was suitable for such a facility.

This report supersedes the March 1977 report. Although a number of changes have occurred since the March 1977 Site Suitability Report was issued, the staff's con-clusion in this report remains unchanged. The proposed CRBRP site is suitable for a -

facility of the general size and type as the CRERP from the standpoint of radiolo-gical health and safety considerations.

17 KE Y WORDS AND DOCUME NT AN ALYSIS 17a DE SC HIP TORS 1

Clinch River Breeder Reactor Plant '

(CRBLP)

Site Suitability Report 17ti IDE N TIFIE HS OPE N E NDE D TE RYS 18 AV AIL AplLITY ST ATEVE NT 19 SE CURITY CLASS (TAs report) 21 NO OF PAGE S Unclassified Unlimited 20 SE CURITY CLASS (TAs pap / 22 PRICE S

N RC F ORM 3JS Mt 34

w f ousYn cL ass gait Posta6s a >s es ta.o Z UNITED STATES C C:UCLEAR NEGULATORY COMMISSION .usent asa o e :D Pl **d ** lil WASHINGTON. D C. 20%6 ]

6 s

Of fiCIAL susNISS PENALiy FOn PmivATE USE, s300 M

'a '19 ') 9 H '

12 05 'i S O I H a l I 1

• US NPC ADM l' l V Ut IlJC E[

P f'l l L Y t. P u t' L I L A l l C N '> NI BR d y e Pi;R isUR L G C f.PY m .

L t. 212 oc wss5 ~

a. so u.w n f-k

D m

x H

Z H

Z m

E D

d m

M O

n O

C M

Z m

D EN 3

m m

C m

D

D m

0

-4 O

D 1

r" Z

-4 e

C 2

m a

8

~