ML20063E912

Forwards Accident Evaluation Branch Revised Evaluation of Four Classes of Core Disruptive Accident,Included as App J to Proposed Fes Addendum,Per B Morris Verbal Request. Calculated Risks of Accidents Not Different from LWR Risks

ML20063E912
Person / Time
Site:	Clinch River
Issue date:	07/01/1982
From:	Houston R Office of Nuclear Reactor Regulation
To:	Check P Office of Nuclear Reactor Regulation
Shared Package
ML20062N348	List: ML20062N347 ML20062N378 ML20063A981 ML20063B672 ML20063B783 ML20063C455 ML20063E912 ML20063N719 ML20063N720 ML20063N721 ML20063N722 ML20063N724 ML20063N725
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FOIA-82-344 NUDOCS 8207140138
Download: ML20063E912 (42)

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r DISTRIBUTMN W3 DOCKET FILE AEB R/F JUL 1 19 82 6

WPasedag LGHulman ucchat "c.: 52-3;7 AD/RP RF rd;".0 P.r.:3 D L):'. F0n: Paul S. Check, Director Clinch River Breeder Reactor Progran Of fice, f1RR FR0f;:

R. W. Houston, Ass *stant Director for Radiation Protection Division of Systems Integration SLUJECT:

REVISED EVALUATI0fl 0F ACCIDE!JTS FOR THE CRCR E"VIR0t:i ENTAL REVIEW In response to your recuest to L. G. Hulnan, dated 3/31/82, the Accicent Evaluation Oranch (AEE) re evaluated the risks resulting from a Class 9 accident at CRGR site, and provided you with our evaluation of the risks from a single Class 9

~

accident discussed in the FES. Subsequently, GILL f~ orris verbally requested us to perform additional risk analyses for a spectrum of four categories of severe accioents. Our revised evaluation of these accidente which is being included as i

r i

Appendix J to the proposed FES Addendum, is enclosed.

l Our analysis is based on the information for a spectrum of four Core Disruptive j

Accident (CDA) classes, their probabilities, and the associated release fractions provided by SILL florris and Ed Rue:ble (SAI). In this evaluation ue have added a new section discussing the economic risks of the accidents on the facility itself.

l Since our evaluation is based on the methodologies of the Reactor Safety Study i

and the related follow on work on calculation of Light water reactor (LVR) l consecuences, our methods at present do not account for the large quantities of sodium present in the CRBRP in place of the large quantities of water present j

in the LWRs. We have, however, bounded the consequences of sodium in our j

assessment.

The results of the AEB analyses indicate that the calculated risks for the selected CRERP accidents are not different from the risks that the staff has presented in the environmental statements of light water reactors which have been Licensed since the issuance of the Concission's June 1980 Statement of Policy.

The accident probabilities and release fractions were provided by DILL Norris and Ed Runble, input for the econcaic ic: pacts of the facility loss uere provided by Argil Toals L n'f provided population distribution infor.Tation. The o 2WNSE j

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Ta evaluatien of the accident risks was perfort.:ed by "ohan Thec ani, x22%1, whc also cocedinated the preparation of attached Ap.nendix J.

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t. Wayne Houston, Assistant Director' for Radiation Protection

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APPENDIX J t

ACCIDENT EVALUTION BRANCH INPUT TO THE FINAL I

ENVIRONMENTAL STATEMENT UPDATE FOR CLINCH RIVER BREEDER REACTOR PLANT Addendue to Section 7.1

[ f.1 PLANT ACCIDENTS INVOLVING R ADIO A CTIVE MATERIALS The staff has examined the Clinch River Breeder Reactor

[

Plant (CRBRP) Final Environmental Statement (FES) with a view to updating the FES to reflect any plant-site-feature or regulatory framework changes that have occurred since l

t the FES was issued in February 1977.

The staff finds that since the issuance of the FES, no significant plant-site

{

changes have occurred which are relevant to environmental 1

is concerns, nor their significant new information relevant i

g to environmental concerns which bears on the project or the environmental impacts or risks of accidents as reported in the FES.

Since the issuance of the FES, however, the Commission has issued a Statement of Interim Policy (J une 13,1980) that provides guidance on the considerations to be given to nuclear power plant accidents under NEPA.

l 1

Among other things the Commission's statement indicated that:

i "this change in policy is not to be construed as any lack of confidence in conclusions regarding the environmental risks of accidents expressed in any previously issued (Environmental Impact) Statements, nor, absent a showing of...special circumstances, as a basis for opening, f

reopening, or expanding any previous.or ongoing preceedingY i

s s

2-jfThe staff in its environmental review of the CRBRP application concluded that the CRBRP did constitute a special circumstance that warranted consideration of Class 9 accidents in the Environmental Statement.

i Because the CRBRP reactor was very different from the conventional Light water reactor plants for which the safety experience base is much broader, the staff included in the CRBRP FES a discussion of the potential impacts and risks of such accidents.

As noted in the Statement of Interim Policy, the fact that the staff had identified this case as a special circumstance was one of the considerations that led to the promulgation of the June 13, 1980 Statement.

jf'(Inexamining 4

i the CRBRP FES, as issued in 1977, the staff has considered the guidance of the Interim Policy Statement which was provided for " Future NEPA Reviews."

We have concluded that the discussion of accidents as presented in the FES generally meets that guidance, Jk except for consideration of the risks due to Liquid pathways.

A discussion of the liquid pathway risks X

b SM k%- I*52.

is included '-'

^

A

[/.1.1 Desian Basis Accident W M 1A

' the results of the staff's analyses of the realistic consequences of design-basis accidents were presented in the FES Table 7.2.

The reported values appear to the staff to be reasonable.

This conclusion is based 9

a

.i

, upon comparison of the realistic dose consequences of CRBRP design-basis accidents with the j

corresponding doses for some recently evaluated Light water reactors (LWRs) such as Comanche Peak, CalLaway, and Palo Verde plants, as shown in Table J-1.

The CRBRP doses are within the range of dose values of some of the LWRs, and the radiological health effects and the environmental impacts of such postulated accidents would be comparable to those'from postulated LWR accidents.

Although the staff analysis of the design-basis accidents does not treat in detail the probabilities of accident occurrence, except as implied in a general way in the development of the accident classification scheme of the previously proposed annex of Apendix 0 to 10 C FR 50, the estimated doses are so smalL that in the staff's judg ment no unreasonable radiological risk to the public health and safety, and to the environment could arise as a result of these accidents.

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s

1 I

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-4 I

/

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-.M Table J.1.#/ Comparison of p'esign-basis fccident 3gW I

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( (C las ses 2-8) 'E i t e b'ounda ry) dos es i - 2 "' ';L/f s'

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w

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3 i/eported in the CRBRP FES with L

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,A

,1 pCorresponding, doses reported in the

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41

/f fj/environmentalftatementsoflome j

LWR a/

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i 1

perating license jrecent p

reviews.

Comanche -

CR8RP Peak CalLaway Palo Verde Accident FES FES FES FES Fuel-handling accidents l

Rems thyroid 0.4 2.0 4.0 0.002 Rems whole body 0.5 0.05 1.0 0.07 Large-break LOCA or site suitability source ters Rems thyroid 1.0 85.0 91.0 8.0 Rems whole body 0.1 1.2 2.2 0.4 Rees Lung 0.2 Rems bone 1.2 F.

-e in this judgyient is acknowledgment that g

Included

-~

accidents of the types represented by those described in FES Table 7.2 for Classes 2-8 have a finite and relatively larger likelihood of occurrence during the operating lifetime of the CRBRP than the occurrence of Class 9 accidents.

Furt he rmo re, their consequences are required not to exceed the dose guideline values of 10 CFR 100.

This acknowledgment ensures that an i

assessment of the adequacy of the engineered safety l

features and operating requirements to mitigate and i

P a

t 8

0 L

(

5-Limit the consequences of such accidents wilL be considered in the safety evaluation of the CRBRP.$veh pq f

considerations at atL contemporary LWRs have resulted in a combination of engineered safety features and A a.

operating procedures so that contribution of these

)(

4 accidents to total risk to the environment is judged to be neg Ligib Le.

The staff wilL reexamine the radiological risk contribution of the design-basis accidents at both the construction permit stage and the operating License stage of CRBRP, giving consideration to the probabilities of occurrence of accidents and to l

their consequences.

The purpose of this reexamination at each stage of Licensingg wilL be to require that the

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plant safety and mitigation systems be designed

[

G equate h nd operated [ o offset the uncertainties arising from a Limited national and international LMFBR operating experience base, and to ensure that the radiological risks of accidents up to and including the severity of design basis events are not greater than i

those of the LWas.

d p

$ C%*

$bkI1.2 Evaluation of Class 9 Accidents l

(

The staff has also performed some new calculations to l

provide additional perspective on the risk associated l

l with the hypothetical Class 9 accidents at the CRBRP.

Presented below is a discussion of the Class 9 accident sequences, estimates of accident l

l

6-i probabili ties, release of raoioactive material to N NSU risks due to the atmospheric and Ip'd pathway exposures, economic costs of the loss of the facility, the uncertainties in predictions, I

and the conclusions.

y All C%f *

(g " Probabilities of Severe Accidents b The Class 9 accident discussed in the FES involved t

a sequence and release representative of possibLe l

t core disruptive accidents (CDAs).

Additional sequences are included here to provide better perspective regarding the risks of CRBRP severe accidents.

LC gp The frequencies of severe (Class 9) accidents at CRBRP involving potential core disruption and containment failure are related to three phases of such accidents.

First, initiation of core disruption must be considered, and this typically requires simultaneous failures of redundant safety systems.

Secondly, there are t

I variations in the release to containment that are dependent on the energy associated with core disruption and the nature of the response of the primary coolant boundary.

F i n a l L y,- the potential for containment failure must be considered.

The probabilities of such events are discussed below.

i

. Initiators of Core Disruptive Accidents Core disruption could be initiated by:

(1 ) failure to adequately cool the fuel as exemplified by a loss of accident (LOCA), or heat sink (LOHS), loss ofgcoolant p

massive flow blockage; (2) failure to terminate the fission chain reactions when necessary, as exemplified by a failure to scram during a loss of flow event (ULOF) or a transient overpower event (UTOP); and (3) core-wide fuel failures as ex emp li fi ed by ropagation of local fuel faults (FFP).

g As discussed on pages 7-2 and 7-7 of the

FES, UK requirements

• for prevention of severe accidents wilL e

be imposed on the CRBRPdesign to ensure that initiation

)(

of core disruptive accidents is made very improbable.

Consequently such accidents are not included in the CRBN' des ign-ba si s ac cident spectrum.

gf.kLOHS events at CRBWPwould have to involve simultaneous A

Loss of availability of the main condenser-feedwater train, of alL three trains of the steam generator-auxiliary heat removal system (SGAHRS), and i

of both trains of the direct heat removal system (DHRS).

l The CRBRP SG AHR S system, which is similar in many l

>k respects to the steam generator / auxiliary feedwater The staff has required in the FES that the design basis accidents envelope extend to accidents with probabilities of one chance in one milLion per reactor year.

p

--n-e

+m-

~-+we-n-e

=e,-

n m

. systems included in PWR designs, consists of one steam-driven and two electrically driven auxiliary feedwater trains.

The DHRS employs a diverse heat removal concept.

Although the staff review of these systems is not complete, it is the judgment of the staff that there is sufficient inherent redundancy, diversity, and independence fn the SGAHRS and DHRS systems to achieve a core degradation frequency due to LOHS events of less than per 10gp/ reactor year.

This estimate is based on a general consideration of typically achievable PWR auxiliary feedwater system reliabilities, the l

potential for common cause failures, and the potential for achieving high reliability in final design and operation through an effective reliability program.

A significant contributor to the LOHS pr([bability for

,q the CRBRP would be from simultaneous Loss of offsite and onsite ac electrical power and the steam-driven 9

auxiliary feedwater train.

&}~

Because of the high boiling point of sodium, the CRBRP primary coolant system wilL operate at significantly lower pressures than LWR primary coolant systems.

This reduces the frequency of Large ruptures in the primary coolant system.

To further ensure that large breaks cannot occur and cause core damage, implementation of preservice and inservice inspection of the primary coolant boundary and a leak detection system wilL be 1

..... ~

7

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required.

In addition, a guard vessel wilL be included t

to prevent unacceptable Leakage from Large portions of

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I the primary coolant system.

For these reasons LOCAs l

t-sre not considered credible (i.e. design basis) events j

l at CR8RP.

The frequency assumed for LONS adequately I

bounds the LOCA contributions to core disruption l

, frequency.

IX coolant inlet region of the CRBRP core is being f

gkThe designed to prevent Large sudden flow blockage such i

as that which Led to extensive damage to two subassemblies in the Enrico Fermi reactor.

Multiple j

intet ports at different planes with interposed

{

strainers wilL prevent large pieces of debris from I

significantly reducing coolant flow to a subasaembly module.

Although sources of particulate debris in sufficient quantity to produce significant flow blockage j

i have not been mechanistically identified, it may be I

postulated that this might occur.

Such debris would be l

expected to be distributed rather generally throughout a large region of the core and would be detectable by I

the core outlet thermocouples if significantly j

reduced core flow were to r e s'u l t.

The frequency assumed for LOHS core degradation sequences adequately.

bounds the flow blockage contribution to core disruption l

f,requency.

I GA I

UTOP and ULOF events involve simultaneous failure of both of the reactor shutdown systems.

Each of these t

i 4

I

-i

~

. systems wilL be required to meet the high standards normalLy applied to LWR shutdown systems.

For example, as specified by IEEE Standa-d 279, each shutdown system wilL be automatically initiated, wilL seet the single gailure criterion, and wilL be tested regularly.

Each system consists of three independent electrical actuation channels of diverse Logic and diverse components.

The mechanical portions of the two systems employ diverse mechanisms and materials.

Although the staff review of these systems is not complete, it is the ju e staff that there eee sufficient inherent redundancy, diversity, and 4

N inde[hndence in the over L L shutdown system designs to X

v expect an unavailability of less than 1Dgp/per demand.

This estimate is based on a general consideration of LWR shutdown system unavailability rates, ATWS precursors, potential for common cause failures, and the feasibility of implementing an effective reliability program to achieve high reliability in the final design and in operation.

Using the assumption, based on LWR experience, that an average of about 10 transients (Yequiredscram) might occur per year of operation x

over the life of the plant, the staff concludes that the combined frequency of degraded core accidents initiated by ULOF and UTOP events is less t h a n 10gp/

per reactor year.

The CRBRP fuel design wilL be required to have an 1

]

inherent capability to prevent rapid propagation of fuel t

2 failure from local faults.

Systems to detect more slowly I

developing faults witL also be required.

Each of these

?

i I

features is considered feasible and i n fact has been

+ka achieved on fuel designs similar to that of CRSRP.

)(

Therefore, the frequency of fuel failure propagation is considered very low.

The frequencies attributed to i

LOHS, UTOP, and ULOF events adequately bound the contribution to core disruption frequency from fuel failure propagation.

M j

f In cummarp the frequencies of core disruption from LOHS, y

UTOP, ULOF, LOCA, and FFP events are alL considered Even when to be Less than 1Qgp/ per reactor year.

combined, tne overalL combined probability of these types of events are estimated to have afrequencyof1&qfper reactor year or less.

This net frequency does not 4

reflect the variations in response of the primary coolant system that might be associated with the various

!j initiators.

Some initiators may result in more severe j

response than others.

This is taken into account as described in the folLowing paragraphs.

1;

<1 /(

f Response of the Primary Coolant System 1

4 q

gf

• he response of the primary coolant system to core L

disruption depends on the amount of energy associated F

with the disruption.

Three categories have been identified and are Listed here in order of increasing i

i l

7 r,-

,,y-

,y--,.--%--,

,,c,,y%--p

.,iy p

_,ep,y

,. +,-

-,_,,.y

,_-,,-,,-,.--,-,,g,-.-,+---r,

. potential threat to containment integrity and increasing release of radioisotopes into containment:

.t R

'I.

Primary system remains intact; no significant

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release of radioactive materials to the i

9 containment atmosphere.

,A 2

1 II.

Primary system initially intact,but ultimately

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4 fails due to ineffective Long term decay heat removal (of the order of hours or more).

Core

^

debris and sodium are initially released into end reach the

(

cav i ty) bg4, ev ent ua l Ly the reactor containment atmosphere through the reactor cavity vents at a slow rate relative to the initial releases of Category III below.

ON

]

III.

Primary system seals fait ue to excessive M(

mechanical and/or thermal loads.

Core Pu, f

solid fission products, noble gases, and i

volatile material would be released into upper containment immediately.*

%Most core disruptive accidents are expected to be iM h 4 I

nonenergetic and to culminate in effects such as Categories I and II above.

Eh e applicant s have proposed to incorporate features to mitigate the above behavior indicated in Categories II and III to reduce the probability of subsequent

• Note:

Longer term release to containment via the reactor

/'

$W'

~] d vents would be an as in II.

ca sityJ

V l

. containment failure.

These include a filtered vent system to relieve containment pressure, a containment purge system to reduce the potential for hydrogen explosions, fans in the annulus between the steel containment shelL and the confinement structure to cool i

the two structures, and vents to relieve pressure from gases generated behind the reactor cavity cell Liners.

These provisions are currently under review by the staff.

.n '

The Class 9 accident release described in Category III corresponds to a core disruptic" of sufficient energy, due to recriticality, to cause mechanical damage to the primary coolant system.

The staff is reviewing the potential for energetic recriticalities to determine the magnitude of energy release anticipated.

If the conclusion of~this review is that an energy release beyond primary system capability cannot be precluded, the staff wilL require some action be taken (e.g.,

that the vessel be strengthened or that head restraints and sodium spray deflectors be installed) to prevent early containment failure from missiles or spray fires.

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 Atomic International's design studies of a 500 MWe LMFBR demonstration plant.

A f'l(Assuming that a core disruptive accident occurs, the conditional frequencies of event Categories I through

. III subsequently occurring are estimatyed as folows:

Primary System Failure - Category I & II combined:

0.9 per CDA Primary System Failure - Category III:

0.1 per CD A A,

X jfThese estimates reflect the lower frequencies expected for coae disruption accidents of increasing energetics.

If of Resoonse M ontainment 4

'1K gf For the purpose of estimating risk given the threats to containment identified above, the folLowing two containment failure modes leading to airborne releases are identified:

!(kA) Failure of Containment Caused by Overpressure.

(B). Failure of Cont ai nment to Isolate.

$$The frequency and consequences of releases to the ground by basemat penetracion are considered to be ove rs h ad ow ed by airborne releaser, as discussed under the subsection entitled " LIQUID PATHWAYS."

9ThestaffwilL require that the containment annulus cooling and vent / purge systems be designed with sufficient redundancy and quality, and be tested and inspected during operation with sufficient frequency, that it can be assumed that their unavailability for anticipated mission times wilL not ex ceed 10g}/ pe r y

d[ mand.

Such systems wilL not be needed to prevent averpressure conditions until many hours after initiation

=

m

.m

-+

iI of a CDA, and would not be expected to be affected by loss of offsite and emergency onsite power unless such power loss should be a long-term outage.

Should the containment systems be required ater a temporary loss of aLL ac power initiating event, failure to recover jpower h before containment failure occurs is estimated 1y/ demand.

, o have a frequency of about Containment isolation is an engineered safety feature at C RBRf?

Such systems are designed to high quality standards and with redundancy.

An unavailability of Less than 10pper demand is feasible for such systems and is y

expected to be attained at CRBRf given that implementation of an adequate reliability program wilL be required.

In summary, the conditional unavailabilities for the containment failure modes are as follows:

Containment Failure Mode A (Mitigating System Failure):

11 per demand Containment Failure Mode B (Containment Isolation Failure):

11 per demand.

g g

Release of Radioactive Ma t e ri a Estimations of the release fractions of the various isotopes which canefcape from the CRBRP are made using the isotope groups defined in WASH-0.

As shown in K1 s+d I

l Table J four release classes are considered and releases to the environment are defined for three c,ontainment modes:

<W 1.

Design Leakage and filtered venting.

16 -

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)30mgd p

2.

Overpressure failure (at about 3.

Containment isolation failure (24" diameter ventilLation Line),

3 /.

/(

Releases from the primary system to the RCB can potentialLy occur by either leaking through the vessel head seals immediately folLowing an energetic CDAjor p(

release from the sodium pool (which forms in the reactor cavity after reactor guard and vessel meltthrough) through the reactor cavity vent system.

1 jfChemicalLy inert noble gases (X e-K r) are not removed from the RCS other than by decay or leakage to the environment.

The remaining fission products, however, can be removed from the RCS by decay, Leakage, filtered venting, and also by naturally occurring depletion mechanisms such as:

M d.

settling;

)[

" e Aerosol agglomeration and

.Thermophoretic deposition on cooler surf aces' a==Jk 3(

ePlate-out Isk The fraction of airborne material which Leaks to the environment in the Long term, depends on the ratio of the Leakage rate to the total removal (Leakage, fit tion, decay, and deposition) rate.

Removal by

)kt L

aerosol agglomeration and setting, considered the JC g

dominant deposition mechanism, is modeled as an exponentially varying time dependent process.

.)Is g

Primary system sodium plays an important role in

i 1

1,

s i

removing fission products in CRBRP.

First, sodium N

1 chemically combines with fission products such as iodine and bromine to form Less volatile compounds.

Second, sodium is maintained welL below its boiling j

point during normat operation and thus fi ss ion -p roduct

's release to the RCS is retarded by the liquid sodium.

J

.1 Third, sodium vapor, after it becomes airburne,

.f becomes an aerosol.

When sodium vapor enters the

}

RCS, for example, a sodium oxide aerosol is formed.

'y i

3

t Since there are over 1 milLion pounds of primary CN>o [q,g.[

a

-t sodium, a dense aerosol (10-100 ug/ cc) wilL be airborne i

i in the RCB.

The airborne fission products witL interact l

with and essentially.espond as sodium oxide aerosols.

4 For the purpose of analysis, therefore, the airborne i

fission products (less noble gases) are considered to be removed at the same rate as the sodium aerosols.

dk T$4eJt g

jfEReferring to Table J./, the variation in release fractions among isotope groups and CDA classes Sb4 depends on the magnitude of competing, concomitant, rate 4

processes (Leakage from the RCS, release to the RCB, and deposition in the RCB).

It should be emphasized that the indicated release fractions do not include removal by decay; this is accounted for in the consequence

[

calculations.

0 l

- Leakage From the RCS b Leakage from the RCB considering CDA Class 1 involves

+ - -.

-y-u-

---w-,

9--

y,-a r--

1 1

1 l

18 -

.ij design Leakage at ratesof10gp/to1Dg//hourand 3

1 filtered venting which is 97% to 99% efficient.

Approximately 57% of the RCB atmosphere wilL be

.j released soon after failure by overpressure (CDA q

j Class 2) since the RCB pressure wilL drop from 1

}

about 2.3 atmosphere (abs) to 1 atmosphere (abs).

l Thereafter leakage through the RCB breach is about I'

equal to the release rates of fission products (1 Ggpf t o 1bq)// hou r).

and other gases into the RCB The a

leakage rate to the environment considering failure of the containment to isolate a ventilation supply or exhaust Line (CDA Classes 3 and 4) is estimated the order of 10gf to 10phour(similar k

o.V.

to be on 3

to the rates after overpressure failureL Thusj or

)q f

each release classy several exchanges wilL occur

(

during the estimated 100-200 hour period in which the sodium pool boils.

Release to the RCB

• f For the purposes of this analysis head release fractions were selected as indicated in Table b ')A b

The fission product inventory remaining in the vessel after the release)(constitutes the pool inventory after y(

head vessel meltthrough.

~

1

, b Table 3*h l.

O

[

H11Nr0 RELEASE SELECTED FOR SOURCE TERM ANALYSIS i

A PRIMARY SYSTEM PRECENT OF CORE INVENTORY p FAILURE CATEGORY RELEASED FROM THE HEAD (%)

X Xe-Kr I

Cs-Rb Te-Sb Ba-Sr Ru La 100 30 30 10 10 3

3 25% I 100 3

3 1

1 0.1 0.1 Pool releases were estimated by considering the relative volatilities of the fission products compared to s dium.

Alkali metals such as Cecium, for example, boi y f 10 to 20 times t h e 4* wee 99lPwedg.ate of sodiu Halogens such as b

and thus fre rete ed

)(,

iodine form compounds with sodium y

$d from the dewium pool at a slower rate than the sodium.

y The rem ning semi-volatiles and solids are released 1L considerably slower than sodium.

Insignificant amounts of the non-volatiles (including fuel) are released to the RCB before cavity dryout.

Once the sodium pool has boiled-off, the remaining dry debris wiLL increase in temperature and attack the concrete bases t.

Additional release of a

%/

fraction of the remaining fission products and fuel is then possible and may be exacerbated by sparging effects caused by off gasing from the concrete during thermal decomposition.

h.Deposition in the RCB_

eposition rates-for airborne fission products are a function of the assumed particle shape and size as welL l

• See. 4.c4wo4e.s &o Talot., 3 4 1

as concentration.

Typical analysis for similar sodium aerosol conditions indicate deposition rates in a single chamber of between 0.5 and 1.0 per hour.

Considering Leakage rates between 10-2 and 10-1 per hour, therefore, indicates that between 1% and 20% of the airborne fission products may eventually be released to the environment.*

The,,, overpressure failure mode drops the containment

)(

pressure to 1 atmosphere thereby releasing 57% of its Jk j

(ne(NA atmosphere.

Since this release emon not occur until g

4 about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the head release and about 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> JC j

af ter pool boiling begins, considerable deposition of Would the airborne material occur [.

The remaining releases p(

g after overpressure relief are similar to those occurring after containment isolation f ailure.

W ffIn addition to the RCB, further deposition wilL occur in i

the reactor cavity and its vent system, the annulus between the containment and confinement (over pressure failure), and the ventilation system (containment isolation f ailure).

Each of these features present a to rtuous flow path and surface area enabling condensation, plate out, and settling.

'he noble gases are conse rvatively estimated (de y not included) to completely escape to the environment for SIS cat This is deemed appropriate 4++

no each CDA c la s s.4 X.3 m.,

w.ut n

deposition occur) and several exchanges of the RC3 atmosphere wilL occur.

rates of 10-4 to 10-5/ hour correspond to

• Design L e aI' a g e 10-5 to 10-f long t e rm relese fractions.

Filtered venting is 97% to 99% efficient.

0 3

Af ter considering the above factors, releases to the I

environment for each CDA Classweree/htmatedforvessel X ii head releases, pool releases and dry cavity releases.

These three release components for each CDA class were then combined into a single set of constant rate releases for input into the consequence model.

The results of this analysis shown in Table J-I dA Comparison of Accident Sequence Frequencies 10he most T

probable class of CDA accident sequences is that in which containment systems function s designed.

)(g Releases to the environment wouldoccurbecdhh,e of design leakage and controlled, filtered venting at about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after CDA initiation.

The likelihood of this accident class is estimated to be less than 10$g/per If[

reactor year.

The doses associated with this accident ents.

class.{3,not expected to exceed 10 CFR 100 guidelines.

)C The two most probable classes of CDA accident sequences y

for which the doses are expected to exceed 10 CFR 100 guidelines are as folLows.

First, a CDA is initiated d [per reactor year),

(Less than 10 a primary system g

failure of Category I and II or III (combined conditional frequency ew1) occurs, and containment failure mode A, containment cooling or vent / purge failure at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (lessthan10gf

)(-

3 demand) folLows.

This class of CDA accident sequences c'cresponds to the FES Class 9 accident.

L Second, a CDA is initiated (less than 103/per Ik-reactor year), a primary system failure of Category Qw-

1 i

k I and II (combined conditional frequency ~1) occurs, and containment failure mode B, failure to isolate (less than1p/ demand) follows.

Both of these classes of CDA accident sequences would therefore, have an estimated 4

bounding frequency of less than 10gvper reactor year.

Furthermore, the frequency of 10pper reactor year bounds each CDA accident class sufficiently such that the combined frequency of the two classes is estimated to be less than 10gper reactor year.

gA Less probable class of CDA sequences for which doses M K

.:x;::::' " exceed 10 CFR 100 guidelines would be initiation of a CDA (less than 10$/ year), primary system QM V it h

failure category I and II (A g Woy III (10g

), and

/uw wpr containment failure mode B,,f ilure to i so la t ey (less than h t.W4 i

1g/ demand) r-co bined frequency of less than 10 "NM to 1p per reactor year.

releases to the N

corfpondto These C A sequence classes environment of four different magnitudes,en desses and

% 44r o r M L* +id s 4 represent an estimate of the frequency of each release g

mode.

Table J-[ gives the inventory of activity of radionuclides timeofshutdown.hheCDA in the CRBRP core at the sequence classes and their releases to the environment are Nb summarized in Table J-J. &The first class in the table, r

T A

which involves no containment failure, is expected to l

s

e. a v.

I,

~

Ttble J.

CRBR CDA stquence clas.T gs j

Boundinggistimate of gontainment Percent of for ventorygeleasdto COA Primary gystem Containment

,;eleasefrequency j

class initiation nv ent1 Xe-Kr I CdRb 2

Las failuregategory failuregode

,per reactor year)

-Sb Ba-Sr Ru 1

Generic Core I and'Il or III None 10 4 100 0.01 0.01 0.01 0.01 0.001 0.001 Olsruption 2

Generic Core I and II or III A

10.s 100 1.0

1. 0 0.6 0.6 0.08 0.08 Disruption (Overpressure) 3 Generic Core I and II 8

10.s 100

1. 3 1.3 0.8 0.8 0.06 0.06 Disruption (Containment Isolation) 4 Generic Core III 8

10 7 100 4.0 4.0

1. 7 1.7 0.35 0.35 Disruption (Containment Isolation) 1 1

• Background on the isotope groups and release mechanism is presented in Appendix VII of " Reactor Saf WASH-1400, NUREG-75/014, October 1975.

2 Includes Ru, Rh, Mo, T.

L 3 Includes Y, La, Zr,

, Ce, Pr, Nd, Np, Pu, Am, Ca.

k i

-l l

\\

i

,1_,

?

TableJ.[goreat112 Activity of pdionuclides in the CR MWt 1

Radioactive inventory

)

Group /,radionuclide in millions 5f curries Half-life (days)

A.

NOBLE GASES Krypton-85 0.1 3,950 Krypton-85m 5.0 0.183 Krypton-87 8.0 0.0528 Krypton-88 11.4 0.117 Xenon-133 52.3 5.28 Xenon-135 56.5 0.384 B.

10 DINES Iodine-131 30.0 8.05 i

Iodine-132 40.8 0.0958 Iodine-133 51.5 0.875 Iodine-134 54.7 0.0366 Iodine-135 50.4 0.280 C.

ALKALI METALS Rubidium-86 0.14 18.7 Cesium-134 0.66 750 Cesium-136 2.7 13.0 Cesium-137

1. 7 11,000 D.

TELLURIUM-ANTIMONY Tellurium-127 3.7 0.391 I

e Tellurium-127m 0.54 109 I i j

Tellurium-129

9. 7 0.048 Tellurium-129m 2.7 34.0

/M-,

Tellurium-131m 4.5 1.25 Tellurium-132 40.0 3.25 Antimony-127 3.8 3.88 Antimony-129 10.3 0.179 E.

AKALINE EARTHS Strontium-89 16.0 52.1 Strontium-90 0.7 11,030 Strontium-91 21.0 0.403 Barium-140 42.0 12.8 06/23/82 0011.0.0 CRBR SPECIAL REPORT APPENDIX J i

k(Continued)

TableJ./

Radioactive inventory Group /gadionuclide in millions If curies Half-life (days)

F.

NOBLE METALS Molybdenum-99 46.6

2. 8 Technetium-99a 40.3 0.25 Ruthenium-103 52.6 39.5 Ruthenium-105 38.5 0.185 Ruthenium-106 19.6 366 Rhodium-105 38.5 1.50 1

G.

RARE EARTHS, REFRACTORY I

OXIDES, AND TRANSURANICS Yttrium-90 O.71 2.67 Yttrium-91 20.4 59.0 Zirconium-95 36.2 65.2 Zirconium-97 40.9 0.71 Niobium-95 34.8 35.0 Lanthanum-140 42.2 1.67 Cerium-141 42.9 32.3 Cerium-143 34.8 3!Hr-I 1 I Cerium-144 20.2 284 7

Praseodymium-143 34.8 13.7 Neodymium-147 17.0 11.1 Neptunium-239 1100 2.35 Plutonium-238 0.38 32,500 Plutonium-239 0.11 8,900,000 Plutonium-240 0.10 2,400,000 Plutonium-241 13.0 5,350 Americium-241 0.16 150,000 Curium-242 14.0 163 Curium-244 0.01 6,630 i

Note:

The above grouping of radionuclides corresponds to that in q;r Tabie a e3 M AG A h% 'f i

i l

l 06/23/82 0012.0.0 CRBR SPECIAL REPORT APPENDIX J j

s

. =.

~

pr duas.aloses ne+ e.-4Q es. sMeQ g

'-----*- of 10 CFR 100.*

The second class in 4

the table is the FES Class 0 accident sequence.

Although the sequences represented by the third and fourth classes

[

would involve earlier releases than the FES Class 9 l

accident, it is expected that they would involve risks (product'of probabfLity and consequences)j(about e

the same and44

/k as the FES Class 9 accident risk.

th Risksj40 & *.

l

(,]Q A, tmospheric Pathways n tL

• The potential atmospheric pathway radiological consequences I

of these accidents have been calculated by the consequence model used in the RSS (NUR EG-03 0) adapted and modified to l

the

---##ia CRBRP site.

The model used 1 year of site M

meteorologic data, projected population for the year 2010 whens CL extending throughoutb :;'- :

"

• 0 ': :

'??-

radius end.

563-km (?50-mi)

d'_;

from the site, and habitable Land fractions within the 563-km (3 50-m i) radius.

The t

essential elements of the atmospheric pathways model are l

shown in schematic form in Figure J.1.

GA

{

To obtain a probability distribution of consecuences, the calculations were performed assuming the occurrence of l

each accident-release sequence at each of 91 different j

" start" times throughout a 1 year period.

Each t

catculation utilized the site-specific hourly

• The comparison to 10 CFR 100 guidelines it made to indicate that this class of CDA does not Fave such severe consequences as other class 9 accidents.

The 10 l'

CFR 100 guidelines were developed for siiing analysis and are often applied in design basis accident analysis.

e-

" + * -

hmG eg 9 h,^ _

g 7

1 i

I i

I e

- I k

b E

r 5

6 i

e 1'

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i i

i I

f i

a r

a e

a f

(

4 t

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I I

t 6..,._>,...,_...,4._,._.,_.....,

P h

i m.

06/23/82 00 0.0 CRBR SPECIAL REPORT APPENDIX J

<,y, 4

_ m -m a>

e 0

e D-e y

.m

-.--e=w g

e m-

,. -. +

p wg -p une-9 y

pw w

eae-- ga og 3 pgg--t-gw--3e-%ym-gqg-

g

.e vuember ose m

ae cameer=s on i

l 1 r De===.sv m

ca.us I

Eh Dianamen Pres =riv 1 r Popsisd.m Emes I

Gsuund j g l

con==.nso n I

s-3 l i

l Figure 7.1 Schematic out ine of atmospheric pathway consequence soul I

{

k e

e

~

~

g

8 meteorological data and seasonal information for the time Q*

period f ollowing each "staM time.

The consequence model y

('

also contains provisions for incorporating the

]

consequence-reduction benefits of evacuation, relocation, h

and other protective actions since early evacuation and K

g relocation of people would conside rably reduce the ws og g exposure from the radioactive cloud and from the U

contaminated ground in the wake of the. cloud passage.

The evacuation model used has been revised from that

• j used in the RSS for better site-specific application.

d[$

The quantitative characteristics of the evacuation model w u,mm g.t '.a used for the CRBRP site see. es tima t es ymade by t he staf fg%

4' 4

N S.

j 3g

'j Ahe applicant's estimates are in a prelimi ary stattof p repa rat io n.

here normally would be some facilities T * ** **

n ea r a plant--such as schools or hospitats-where special equipment or personnel may be requi red to effeet f

evacuation, and there may be some people near a site who may choose not to evacuate.

Seve ral facilities of this type have been identified near the CRBRPsite, such as y

the Lofdon County Memo rial Hospital, Roane County High T

School, and facilities related to national security.

(,

[

Therefore, actual evacuation ef fectiveness could be t

greater or less than that characterized but would not y

l s; p.{ ca.d29 l

y

,b e expectedtobe[:.y a.

Less.

The other protective actions include:

(1) either complete f

denial of use (interdiction), or permitting use only at a sufficiently later time after appropriate

i i :

decontamination of foodstuffs such as crops and milk, (2) t decontamination of severely contaminated environment (Land and property) when it is considered to be

[

i economically feasible to Lower the levels of contamination i

to protective action guide (PAG) levels, and (3) denial

(

of use (interdiction) of severely contaminated Land and p rope rty for varying periods of time until the contamination levels are reduced by radioactive decay Am4Ag and weathering that land and prope rty can be economically 4

decontaminated as in (2) above.

These actions would reduce the radiological exposure to people from immediate and/or subsequent use of or Living in the contaminated environment.

Early evacuation of people f rom the plume exposure pathway sonA X

+

(EPZ)andprotectiveactions as mentioned above are considered T

essential sequels to tevere nuclear reactor accidents involv-ing significant release of radioactivity to the atmosphere.

i Therefore, the results shown for CRBRP include the benefits of these protective actions.

There are uncertainties in each facet of the estimates of e

consequences (See Figure J.1) and the error bounds may be l

.n.44 i

la rge as they a re f o r[p robabili ties.

The result s of as the calculations, based on conservative assumption of 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> delay in evacuation, are summarized and compared i

- e

. I l

I with those f or Midland Plant (LWR) in Table J.5 as expectation values, or averages of environmental risk per year of reactor operation.

These averages are instructive as an aid in the

[

comparison of radiological risks associated with potential CRBRP 1

accidents and those risks calculated for recently evaluated LWRs, (e.g., Midland) f or which calculations of radiological risks were made in essentially the same manner.

The table shows the average j

risk associated with population dose, early fatalities, latent I

fatalities, and costs of protective actions and decontamination.

f-g

~

Tab Le J.)3 compa rison of average values of environmental risks t

/

a T,

t e

t

-gq[*{

to gelected CRBRP gccidents with due those f or Midland Alant a

Environmental risk (pe r reacto r year)

CRBRP Mid la nd l

Population exposure i

Person-rems within 80 km 3.5 26 Tot al pe rson-ress 5

130 Early fatalities 6x10-6 1.5x10-5 l

t Latent cancer f atalities AlL organs excluding thyroid

0. 3 x10-3

7. 2 x 10-3 Thyroid only

0. 0 4 x 10-3

1. 8 x 10-3 Cost of protective actions

$690*

S4,800*

and decontamination

• 1980 dolLa rs i

l I

L i

~ ^^

. The population doses and latent fatality risks may be compared with those for normal ope ration population doses given it, t ab le 5.13 of the FES.

The comparison shows that the ac ci dent risks are comparable to operating risks.

For perspective and understanding of the meaning of the early fatality risks of 6x10$ pe r reactor year, howeve r, the staf f notes that to a good approximation the population at risk within Ot90LM k k about 16 km (10 miles) of the plant is about 80,000 ge rsons h$e.d upon 4his,eoh h4.g* gf.34gdj e

in the year 2010 4 Ac ci dent al f atal ties per year for a population of this size, based upon overalL averages for the United States

• are approximately 18 from moto r vehic le ac ci de nt s, P

6.2 from falls, 2.5 from drowning, 2.3 from burns, and 1.0 from g,

j t

fire arms.

T,

[L() l Liquid Pathways "i

9 Surf ace water hydrologic prope rties at CRBRP should be s *, m i l a r t o k

those used for the Liquid Pathways Generic Study (LPGS) smali river site which was based on the Clinch - Tennessee - Ohio -

h Mississippi ri ve rs system, although the river uses and population in the LPGS were based upon national averages and have not been directly compared to the CRBRP.

The groundwater characteristics at Clinch River do not indicate any unusual adve rse transport characteristics.

ased on risk to individual in "CONAES Final Repo rt," National Research C oun ci l, Chapter 9, pp 577-534, 1979.

s, Ad di t ional Ly, the CRBRP is a considerably smalLe r plant than l

LPGS case (CRBRP is 1121 MWt vs. 3425 MWt assumed f o r LPGS),

and contrary to the Light Water Reactors characteristics, 1

CRBRP does not contain any large storage of water which could l

se rve es a potential " prompt source" to the environmental liquid pathways.

Therefore, only the radioactive material Leached from the core debris by the local groundwater is Likely to be transported to the Clinch River.. This source was found in the LPGS to be considerably smaller than the

" prompt source."

Theref ore, based on the p r eli mi n a ry f

i

~

appraisal of the liquid pathways, the staff concludes that the liquid pathways impacts of CRBRP would be probabLy smalLe r than those for the LWRs analyzed in the LPGS "SmalL River" site case.

f (g) Other Economic Risks i;

The re are ot her economic impact s and ri sks which are not included in the costs that can be given a monetary value.

j These are accident impacts on the facility itself that result in added costs to the public, primarily t axpay e rs.

These i

costs would be f or decontamination and repair or rep la c eme nt l

[

of the facility, and replacement of power.

Although it is possible that the facility would simply be decommissioned L

rather than restored following a se ri ous (core-melt) accident

)(

an assumption of restoration is conside red cons e rv a t ive l

1

. (high cost) in reflecting the cost impact of an ac cident.

If the worth of the facility at the time of an accident is perceived to be more than the cost of restoration of the facility, then presumab Ly t he facility would not be restored and the cost impact would be less than the restoration cost, so that use of the restoration cost would represent a high side estimate.

Because the worth of the f acili ty is primarily in the nature of research and development, the actual value cannot be quantified any more accurately than it is perceived at the time.

Expe ri ence with such costs is currently being accumulated as a result of the Three Mile Island accident.

Although CRBRP is considerably smaller in electrical output than the Three Mile Island pla nt, the physical size and complexity of CRBRF

/

is comparable and the cost of decontamination and restoration is estimated to be about the same as that for Three Mile Island.

If an accident occurs during the first full year of CR3RPope ration (1989), the econcaic penalty associated with the initial year of the unit's operation is estimated at

$2250 mi t Lion f or decontamination and r es to ra tio n, including replacement of the damaged nuclear fuel.

This is based on a $952 mit Lion value in 1980 dolla rs as reported to Congress by the Comptroller General (1981).

The S952 milLion in 1980 dolla rs has been escalated at 10% to 1989.

Although prope rty damage insurance would cover part of this, the P

j l

insurance is not credited because the insurance payment times the risk probability would theoretically balance the insurance premium.

1 In addition, staf f estimates average additional production costs of $25 milLion (1989 dollars) f or replacement power during each yene the CRBR is being restored.

This is based on applicant's net projections of operating savings during the first six years of operation, discounted at 10% to 1989.

Assuming the nuclear unit does not operate for 8 years due to shutdown, the tot al addi tional replacement power cost should be approximately $200 mit Lion in 1989 dollars.

The probability during each year of the units service Life of sustaining a total loss of the original f acility as a result of a disab Ling accident is taken from Table J-3 as 1.0 x 10-4 Multiplying the previously estimated costs of $2450 mit Lion f or i

an eccident to CRBRPduring the initial year of its operation by T'

theabove(1.0x10-4) probability results in an economic risk of

_L-approximately $250,000 (in 1989 dolla rs) applicable to CRBR 9 d-during its first year of operation.

This is also approximately

~~

the economic risk (in 1989 dolLa rs) to CRBRfduring the second and each subsequent year of its operation.

Although CRBRf would depreciate in value such that the economic consequences of an ac ci de nt becomes less as the unit becomes older, this is considered 'co be offset by a higher cost of decontamination of the unit in later years.

..e

- 3t

/

of 7'vironne t ction/

i risks of mos#

ric p

hways, 1

The assessment s

, show at t r

ks re assue g

'easo ble protec ve s gnifi tly owe ha similarly calcu ted va es for ht water react es curre Ly being/ cens for o ratio

See, r

/

l ple, lNUREG-0894 ES f r llava

(

REG-0813), D or Seab ok ta ex on j f

' QiUREG 564),

DES fo git for he envi onnental kg of light water r,eactos.

[f]& e foregoing estimates of f requencies and risks associa UNCERTAINTIES Th P

CRBRPhave included allowances for uncertainties.

For example, i

unavailability estimates for shutdown and heat removat systems have been set high enough to include allowances for potential common cause failures.

Howeve r, t he ri sk s f rom sabot age or f rom external natural events such as earthquakes, to rnadoes, and floods beyond design bases for such events are difficult to quantify.

This situation is generic to LWRs and advanced r e a c t o rs such as CRBRB NRC is presently devoting significant effort to developing methods for quantifying risks from such events.

Compliance with current NRC siting structural, and seismic design criteria, and wi th 10 CFR 73 @o n physical security provides assurance that reactor related risks from external events are adequately low.

The CRBRP design wiLL be required to meet alL these criteria.

Risks and the uncertainties in risks from the CRBRP related to sabotage and external events are not expected to differ

~

l

)

A-21 !

significantly from such risks and their associated uncertainties at LWRs.

One additional potential containment failure mode not q ua nt i f i ed

{

above involves early containment failure and release caused by either a spray fire or missile generated from a very energetic CDA.

The staf f wilL review the potential f or CD A energetics to ensure that necessary design enhancements of'the primary coolant l

system are incorporated such that the probability of prisary l

coolant system failure as a result of physically reasonable core rearrangement of sodium, cladding, or fuel wilL be very smalL.

However, because it is possible to hypothesize nonsechanistic I

?

and speculative coherent and rapid core reconfigurations le adi ng 4

to high reactivity ramp rates, high energetics cannot be entirely precluded.

Quantification of the frequency of this very improbable event at this time would involve such la rge uncertainties

[

l that the results would have no reat meaning.

i i

It should also be noted that the results do not fully account l

for the effects of the sodium coolant on the radioactive source I

term.

For example, inclusion of the effects of sodium is i'

expected to reduce the quantity of iodine available for leakage.

The large mass of sodium aerosol also contributes to the agglomeration l

and settling of aerosols in the primary containment.

On the other l

i i

hand, the sodium activation product s would be released together l

with the primary coolant, thereby adding to the amount of radioactive li l

l :. -- -.

..r

., ~. '

~

! ma t e ri al released to the containment.

On balance, it is expected that the risk contribution of the presence of radioactive sodium i: :ignif'::nt,

...J. ;m. _ ! :- : :Awould 5

k not invalidate the conclusions of these calculations.

Further i

consideration of this subject wilL be included in the staff's review of the Probabilistic Risk Assessment for this plant, and in the staff's Safety Evaluation Report.

i In summary, from the limited quantitative analyses discussed I

above, it is the best estimate of the staff that the frequency of individual classes of severe accidents resulting in 1

fatalities or even doses exceeding 10 CFR 100 guidelines i s l;

fhf Less than 10 /pe r reactor year.

Compli ance wi th. current design criteria wilL ensure that risks from external events and sabotage are acceptably low.

The ri sk s estimated for I

c obui(A CRBRffrom t?.e F Eg Cla s s 9 a c ci de nts ap pe a r i n Tab te J -5.

The estimated probabilities of seve re ac cident s for CRBRf do j(

not depend in a significant way on the Reactor Saf ety Study l

(RSS) which was published i n 1975.

However, the RSS has been reviewed to gain perspective regarding r ep r es e nt a tive system unreliabilities and general aspects of methodology and f

un c e rt a i nt i es.

For that reason the folLowing di scussion of r

the current status of WASH-1400 is provided.

l a

i

+*- -

v---

a m

g y-y w-w w-e

,~,y

--m-

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w-r p-

g a S E

- i i

7

" l 0

In July 1977, the NRC organized an Independent Risk Assessment Review Group to (1) clarify the achievements and limitations of the Reactor Safety Study, (2) assess the peer comments l

i thereon and the responses to the comments, (3) study the current state of such risk assessment methodology, and (4) recommend to i

the Commission how and whether such methodology can be used in i

the regulatory and Licensing process.

The result s of this

[

t study were issued in September 1978.

This repo rt, commonly called the Lewis Report, contains seve ra l findings and recommendations concerning the RSS.

Some of the more significant l

[

findings are summarized below.

(1)

A number of sources of both conservatism and non-conservatism i

in the probability cattulations in RSS were found, which were very difficult to balance.

The Review Group was unable to r

determine whether the overalL probability of a core melt given in the RSS was high or Low, but they did conclude that the l

error bands were understated.

f (2)

The methodology, which was an important advance ove r earli er

{

t p

methodologies that had been applied to reactor ri s k, was

[

k I

sound.

l (3)

It is very dif ficult to follow the detailed thread of l

calculations through the RSS.

In particular, the Executive I

l Summary is a poor description of the' contents of the repo rt, r

l y

,-sg w.

3-,e

,_,-,-__-,.3,....,,.,.,.a

-.-r,,

,y7_

--.._,-.e--_,.,,,g.,,

,,--+

n,,,. -

.s!

........ i 1

should not be used as such, and has Lent itself to misuse in the discussion of reacto r ri sk.

On January 19, 1979, the Commission issued a statement of policy concerning the RSS and the Review Group R ep o rt.

The Commission ac cepted the findings of the Review Group.

These findi ngs have been considered in evaluating the potential risks from CRBRR jf/,3 CONCLUSION Ob The foregoing sections have evaluated the envi ronmental impact s of severe ac cident s including potential radiationexposurftothe l

j population as a whole, the ri sk of near-and long-term adve rse I

health effects that such exposures could entailjand the potential economic and societal consequences of accidental contamination of the environment.

The assessment of environmental risk from several I

categories of accidents, assuming reasonable protective action, provides perspective on the overall risk from CRBRP accidents in comparison to those from LWRs.

From this comparison it is concluded that there is no basis for disagreement with the FES conclusions (that the CRBRP a c ci d e nt ri sks wil L not be different from those of

en; LWRs).

If l

~

~

-.