ML19221A814
| ML19221A814 | |
| Person / Time | |
|---|---|
| Issue date: | 02/28/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0555, NUREG-0555-07.1, NUREG-555, NUREG-555-7.1, SRP-07.01, SRP-7.01, NUDOCS 7907090176 | |
| Download: ML19221A814 (12) | |
Text
Section 7.1 February 1979 ENVIRONMENTAL STANDARD REVIEW PLAN FOR ES SECTION 7.1 ENVIRONMENTAL IMPACTS OF POSTULATED ACCIDENTS INVOLVING RADI0 ACTIVE MATERIALS:
PLANT ACCIDENTS REVIEW INPUTS Environmental Report Sections 2.1.2 Geography and Demography: Population Distribution 7.1 Environmental Ef fects of Accidents: Station Accidents Involving Radioactivity Environmental Reviews
- 2. 5.1 Socioeconomics: Demography 2.7 Meteorology 3.2 Reactor Steam-Electric System 3.3 Plant Water Use 3.4 Cooling System 3.5 Radioactive-Waste Systems 3.8 Radioactive Material Movement Standards and Guides Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radio-logical Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors" Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radioiogical Consequences of a Loss-of-r.calant Acc dent for Pressurized Water Reactors" Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants" 10 CFR Part 20, Section 20.105, " Permissible Levels of Radiation in Unrestricted Areas" Other Responses to requests for additional information Applicant's Preliminary Safety Evaluation Report (if tendered with the ER)
REVIEW OUTPUTS Envircnmental Statement Sections 7.1 Environmental Impacts of Postulated Accidents Involving Radio-active Materials: Plant Accidents 109 029 7 ' '-
7 907090\\%
February 1979 Other Environmental Reviews None I.
PURPOSE AND SCOPE The purpose of this environmental standard review plan (ESRP) is to direct the staff's evaluation of the environmental risks of accidents involv-ing radioactivity that can be postulated for the plant under review.
This is accomplished through evaluating the calculated realistic individual population exposures for these accidents. Thc applicant's analysis and identification of accidents will be reviewed and compared with the requirements for analyses of accidents given in Appendix A of this ESRP (the staff's proposed Annex to Appendix D, 10 CFR Part 50).
In addition, an independent staff analysis will be conducted and reported.
The scope of the review directed by this plan will include consideration of a limited amount of plant specific data (demography, reactor power level, and exclusion radius) in sufficient detail to calculate realistic dose and man-rem value estimates for specified a cidents.
II.
REOUIRED DATA AND INFORMATION The kinds of data and information required will be affected by site-and station-specific factors, and the degree of detail will be modified according to the anticipated magnitude of the potential impact.
The following data or information will usually be required:
A.
Identification of 3pplicable accidents and analysis of the conse-quences of these accidents (f rom the ER).
B.
The projected demographic data to an 80-km radius from the plant for the year 2000 (f rom the ESRP for ES Section 2.5.1).
7.1-2
February 1979 C.
Plant data such as power level, safety equipment details, and exclu-sion boundary distance that are used as input to the dose calculations (from the ER).
D.
Relative concentrations (X/Q) at appropriate distan:es f rom ef fluent release points.
X/Q values may be determined from or. site meteorological dat: 1 at the 50% probability level (Section 2.3.4 of Regulatory Guide 1.70) or at 10% of the levels given in Regulatory Guides 1.3 (for a BWR) or 1.4 (for a PWR) (f rom the ER).
III.
ANALYSIS PROCEDURE The reviewer will examine the applicant's description of accidents con-sidered, the assumptions made, and the environmental consequences and compare them with the requirements given in Appendix A of this ESRP.
The reviewer for ES Section 2.5.1 will be consulted, and other appropriate sources of demographic data should be reviewed to ascertain that sufficient population data are presented for the reviewer to calculate man-rem values.
In each case, the reviewer will select and emphasize various aspects of the material covered by this ESRP.
The areas seiec+9d for emphasis during the review will be based on an inspection by the reviewer of the material presented to determine (1) whether the material is similar to that recently reviewed on other plants and (2) whether items of special safety significance are involved.
The following is a summary of the normal review procedure.
A.
Review the applicant's environmental report to determine that all accidents identifica in Appendix A are described as they might occur at the proposed plant.
If the applicant claims a given sequence of events as unlikely at the proposed plant, technical justification must be p.ovided for not estimating the consequences of a given accident using the assumptions provided in Appendix A.
The consequences will bc ated for an individual exposure at the exclusion boundary and for the integrated man-rem to 80 km.
Use the projected population for the year 2000.
]00 0l'
/
Gv s 7.1-3
February 1979 B.
Prepare the text for the environmental statement and make dose esti-mates.
The dose estimates are made using the assumptions in Appendix A.
To represent more realistic dispersion conditions than assumed in the safety r,e v i ew, the reviewer will use either X/Q values that are one-tenth the values in Regulatory Guides 1.3 and 1.4 or the 50% X/Q values based on onsite meteoro-logical data. Adjustments in assumptions will be made based on the engineered safety features at each plant. Dose estimates are made for an individual expo-sure at the exclusion boundary and for an integrated exposure for the projected population for the year 2000 to an 80-km radius.
IV.
EVALUATION The reviewer must determine if the environmental risk from accidents is acceptably low.
Although no quantitative criteria exist, comparisons of the calculated impacts with allowable and experienced exposures to the population and individuals will permit a judgment as to acceptability of the proposed project.
The reviewer will consult with the reviewers for ES Sections 2.5.1 and 2.7 to ensure that appropriate demographic data and X/Q values were used in the analysis.
The reviewer will compare the dose estimates with the gu.delines of 10 CFR Part 20, the man-rem value estimates with normal environmental radiation levels, and both the dose estimates and man-rem value estimates with estimates that have been made for previously reviewed nuclear power plants. The reviewer may conclude that the calculated environmental impccts of realistic accidents are acceptable if:
A.
The realistic doses of all accidents more likely than the Class 8 accidents are less than, and in general are much less than, Part 20 guidelines.
B.
Doses resulting from the worst postulated accident are comparable to Part 20 guidelines.
e m *c 7.1-4
February 19/9 C.
The man-rem value estimates are small in comparison to normal environ-mental radiation levels and are comparable to man-rem value estimates made for previously reviewed nuclear power plants.
The reviewer wiIl determine if there are any plant-or site-specific condi-tions that could affect the staff's generic conclusion regarding the low degree of environmental risk associated with accidents (for example, a postulated accident with unusually severe consequences that is unique to the proposed project).
For any such condition, the reviewer will evaluate the likelihood of the condition, will estimate the level of environmental risk, and will reach a conclusion as to risk acceptability.
If the predicted risk is unacceptable, the reviewer will recommend consideration of alternative designs, locations, or operating practices to reduce the risk to acceptable levels or to correct adverse unique site-or station-specific f actors or conditions.
V.
INPUT TO THE ENVIRONMFNTAL STATEMENT ES Section 7.1 w111 contain the following information:
A.
A general discussion of plant -idents and the methodology used to calculate realistic doses.
B.
A section that presents the staff's findings, relative to this plant, as to representative accidents and the estimates of doses and man-rem values for these accidents.
C.
A conclusion as to the degree of environmental risks due to radio-logical accidents at this plant and the level of significance of the resulting environmental impacts.
V.
REFERENCES None 109 035 7.1-5
Appendix A to ESRP 7.1 February 1979 ENVIRONMErlTAL STANDARD REVIEW PLAN FOR ES SECTION 7.1 EfWIRONMENTAL IMPACTS OF POSTULATED ACCIDENTS INVOLVING RADI0 ACTIVE MATERIALS: PLANT ACCIDENTS DISCUSSION OF ACCIDENTS IN APPLICANTS' ENV'FONMENTAL REPORTt ASSUMPTIONS Ihe comp!cte text of theproposcJ Armex to Ap;n,Jix D,10 CFR Part 50,Jd!an s. It ws originally published in the Federal Repster De, embcr 1,19 71 (36 i R 2:851).
This Annex requires certain assumptions to be made Those classes of accidents. other than Classes I and 9, in discussion of accidents m I nvironmental Re{nts found to have sigmficant adserse envuonmental effects submitted pursuant to Appenda D by apphcants for shall be evaluated as to probabbt), or fiequency of construction per: nits or operating hcenses for nuclear ociurrence to penmt estimates to be made of envuon-power icactors) mental nsk or cost ansing from accidents of the gnen dass.
In the consideration of the environmental risks associated with the postul.ited acciJents, the probabd.
Class I events need not be considered because of their ities of their occurrence and their consequences must tovi.I consequences.
both be taken into account. Since it is not practicable to consider all possible accidents, the spectrum of acci-dents, ranang m seventy from troial to very senous. is t
8 events am snoye considercJ m safety analpis divided inia classes.
wports and Al C staf f safety evaluatmas They are used, together with highly conservative assumpoons, as the LUch class can be charactennd by an occurrence rate daign-basis events to estabbsh the performance iequue-and a set of consequences.
ments of cngneered safety features. The highly conser-vatne assumptions and calcularmns uses in Al C sifety Standardized examples of classes of accidene to be evaluations are not suit able for ensuonmenta nsk considered by applicants m prepanng the sectmn of evaluatmo, because their use would result in a 3ubstan-Environmental Reports dealing with accidents are set ual osewnimate of the ensuomnental nd. Nr on-out in tabular form below. The spectrum of accidents.
wason. Class X cvents shall be evaluated ea hstit all).
from the most trisial to the most sesere, is divided into Consequences predicted m this way w!l1 be fjr less severe nine dasses, some of which hase subclasses The acci-than those psen for the same esents in s?cty anah us derts stated in each of the eight classes in tabular forrn kP'uts where nune consersative evaluationare used.
below are representatne of the types of accidents that must be analyzed by the appheant in Ermronmental Re oaunences m Chss 9 invohe sequences of Reports; however, other accident assumptions may be p stulated successive failures more severe than those more suitable for individual cases. Wher assumptions postulated for estabbsfung the design baas for protective are not speciGed, or where those specified are deemed systems and enpneered safety features. Their conse-unsuitabic, assumptions as realistic as the state of quenas could be severe. Iloweser, th probabihty of knowledge permits shall be used, takm their occurrence is so small that then environmental nsk specific desmn and operational char'g into account the acteristics of the is extremely low. Defense in depth (muhiple physical plant under consideration.
barners), quality assurance for design, manufasture, ar '
operatmn, continued surveillanse and testing, and con-For each.iass, acept Classes I and 9, the ensuon-servative design are all apphed to prmiJe and mamtain mental consequences shall be evaluated as indicated.
the required high degree of assurance that potent al accidents m this class are, and will remam, sutheten ly remote m probabihty that the envuonmental nsk is ext.emely low. For these reasons, it is not nece,sary o I Although this Aane s refen to applx ant s' E nvironmen tal discuss such even ts in apphean ts Er viromnen t al Reports. tt.e current asmmptmns and other pronuons thereof are applicatJe, except as the content rnay otherwise require, to
- P" Al C draft and final Detailed Statements.
Furthermore. it is not necessev to take into account
'Prehn.u.ary guidance as to the centent of arrbcants' I nuron-these Class 8 accidents fc r wtuch t':e apphcant can mental Reports was prouded in the Draf t ALC Gosde to the Prcraration of Enuronmental Reports for Nuclear Power Plants ggP g
g g dated f ebruary 19, 1971, a document rnade available to the thereby the calculated risk to the ensircnment made pubhe as wen as to the appheant. Guidance concerning the equivalent to that which might be hypothesi /ed for a discusuon of accidents in environmental reports was prouded Class 9 event.
to app hean t s in a September 1,1971, document e n title d
" Scope of Apphcants' f nvironmentat Reports wi'h Respect to Transpertauon, Transmission tanes, and Accidents," also roade Applicant may substitute other amd< nt class heak-available to the public.
downs and alternatne values of radio, ctiv ' material 7.1-A-1
Feb.> ry 1979 rho and anWm il assumptions. if such substitution of radio gtne n ote: 4 ou, f contaunnent. i hese is justilled in t 2 n$it nmental Report.
releaws slu!! bt inclaued and e aluated under routme releases m actotdance witi, pmpus,4 Appenda 1.
ACCIDLNT -3.0 RWWASTE SYSI CJ raunE ACCIDENT ASSUMPTIOds TABLE OF CONTENTS maljimctwn (meludes 3.1 Epi;mwn!
o J..
- 'r operator en or).
A cijcru 1.0 linal incidents.
(a) Radatne pses anJ hquiJs. 25" of average 2.0 Small releases outside wntamment.
i"sentory m the largest storage tank shall bo usumed to 3 0 Radwaste system failures.
be releavd.
3.1 Equipment leakaec or ma! function.
3.2 ReIcaw of waste ps storap rank contents.
(M Meteorology assumpt:ons A ulues are to be 3.3 Release ofliquid waste storage tank contents 1/10 of those pnen m Al C Safety r aide No. 3 or 4.2 4.0 lission products to puman system (ItWR).
4.! Fuel claddmg defects.
h 1 Gecquences should be calculated by weighting 4.2 00 design tranuents that induce fuel fadures the en ets in diff erent Jues tions bv the frequency the abose those expected.
wuJ Rw s in t ach ducction.
5.0 Fission products to pomary and secondary sy stems
( P% R L 3.2 &lcw of waste ps storau t.vrA contents 5 I Fuel cladding defects and steam generator leaks.
( n'ludes fadu:e of release uhe and rupture dakst 5.2 Off-design tran sien t s that induce fuel tailure abose those expected and steam generator leak.
. 100i of the aserage tank inventory shall be Ia) 5.3 Steam generator tube rupture.
auumed to be released.
6.0 Refucimg accidents.
6.1 Fuel bundle drop.
(b, Meteoro!ogy awumptions: 60 uf ues shall be 6 2 lieny object drop onto fuelin core.
7.0 Spent fuel handhng accident.
1/10 of those gnen m Satety Guido No. 3 or 4.
7.1 Fuel assembly drop m fuel storage pool.
(c) foi <:quences should be calculated by weighting 7.2 IIcav" object drop onto tu ' rad.
the ef fects in ihtferent ducetmns by the f requency the 7.3 Fut asx m op.
w md b* ws m each duceuon.
5.0 Accident initiation events considered m design bass evaluation in the safety analy sis repo-t.
3.3 Rclene ofliquid waste storage tJHN entdents h 1 Ins-of-coolant acciJents.
8.l(a) Break in instrument line from pnmary system (a) Radmactise hymds: 100 7 of the averate storac that penetrates the containment.
tank msentory shall be assumed to be spille'd on :5e S.2(a) Rod gection actident (PWR)
C(b) Rod drop accident (BWR).
tbor of the bulJ ng.
8.3(a) Steamhne breaks ( PWRs outside c an t a.n -
(b) Buddmg structure sha!! b - mmed to remin men t ).
mtact.
8.3(b) Steamhne breaks (BWR).
(c) Metecrologo assumptions: x Q ulues sha!! be ACCIDENT ASSUMPTIONS lllo ;.f those given in AEC Safety Guide No. 3 or 4.
ACCIDENT - 1.0 TRIVI \\ L INCIDENTS (d) Consequences shuuld be caleu!ated by weghting thc effects m different du etmns by the f:equency the These incidents shall be meluded and evaluated under wmd blows in each direenon routme releases in accordance with proposed Appenda 1.1 F # '"
ACCIDENT-2.0 SMALL RELEASE OUTSIDE COL avadaNe at the CommoxonN PWh Doment Ro. m. 1717 11
. FAIN M ENT s,,eg s w, w onyton, o c,,7 3 cn,,yn,,o n,e nys to,,
Dnmon R% tur StandarJt U.S Naslear Regulatory Coin-These releases shall include such thines as releaws m:sw,n. w ash.nton, D.C.
a5 5 <. (I he w tu gu.Jc s hn e he n through stean.hne rehef valves and sme l s'pdis and leaks
"" "" 3 '*"" 3 " 5 R '."l e L ", W '*> C' d " Il '"3 l
Reymon 2 Regalatory Gub 14 both dated Jane 1974.
Cer:cs of ti.ese gWn m ay be @ tamed h nyuest from the U.S. Sa.: car Regulatory Com m m w n. % viurgian, D (. 205 5 5.
M I R 111 t L June S.1971.
Auento n Ihret tur of Utin e ef stan 1.cds Deu tepment.)
f] *g (j 7.1-A-2
February 1979 ACCIDENT-4.0 FISSION PRODUCIS TO PRI\\l ARY (c) Secondary sy stem equihhnum radaucinity poor SYS T E.\\l ( BW R )
to the transient ' hall be bned on a 20 caliday sicam generator leak and a 10 epm blowdown rate 4.1 Fuct cladJmg defects.
(d) All noble gases and o l' of the ha s:ns m the Release from the se esents shall be mcluded and 3 cam reashing the condenser shall be awumed to N evaluated under routme releaws in accordanse with released by the condenser air eiector.
propoWd Appendl\\ I.
(C) \\Teletirtphigy auumptions' \\.Q s J!aes sb(M!d be 4.2 Ofj-Jcsign transients that induce fitel fadures lll0 of those gnen in Al C Salety Guide No. 4.
abore those erreded (such as flow blockage and flus maldistobutions).
< fi Consequences should be calcul.eed by w eightmg the ettects in dif ferent ducctions by the ta quen ) the (a) 0 029 of the core msentory of <
'.,te gases a nd umJ Nows m each duection 0 02 of the core insentory of halogens shah be assumed to be released mto the ieactor cool.mt.
5 3 Stcam cencrator tube ruptsrc.
(b) l' of the halogens m the reactor wolant shah be (a) 15 of the cerag unentory of noHe me, and assumcJ ta be released !nto the steamlux halogens in the pinnary wo! ant shall be ass J is be n
released mta the secondary coi lant.
(c) The meshamcal vacuum pump sh dl be a,unned to be automatically isolated by a high radution signal on The nem prnnary coobnt utnny shad bebaed the steam!me-on UT Uiled f uel.
(d) Radouttnity shall be assumed to carry user to (b) ! qmhbrium radmactnity pnm to ruptme shah th: con den se r where 107 of the 't m shad be L N.ca.m a 20 cillon per day stean mencrat< r leak a:. !
assumed 'o be am!aNe f or leakag from the wndenser a 10 gm P..dow n rate to tk ensironment at 0.5 Jay for the course :f the acsident (24 houa h
't) AH noHe gases and U.I' of tb lu!c ens m t' 1 steam renh.ng the condenser du!! he wn:.d to b (e) \\le!corolog) a Wtm[ Gon s \\ Q sables sha!! he idened by the conden3er,nr ejf tt;,r.
I!!O of those yivan in Al C Safer) Guide No. i datrJ Nmember 2.1970.
(d) \\1eteorology assan.ptmns \\Q valws s!un be
!!!O of these pnen in Al C %fety Gmd> No. 4 af) Conwquences should be calmu!ated by weightmp the effects m different directions by the frequeacy the ic) Comequences should be talcubted h3 weghtme wind blou s in exh d:rection.
the effects m dif f erent directior s b the tr.42:n ) th; s
wind bh >w s in cash diresinm.
ACCIDENT -5.0 FISSION PRODUCTS TO PRINI ARY ACCIDENI--6.0 REFUI. LING ACCIDI NIS AND SECONDARY SYSTE\\1F (PRESSURIZED WATER REACTOR.
h l Tuc ' bro:J/c drop.
51 E<cl t /f!Jm: Jctats and ucam generator IcA.
(a) II.c ;eap aan n t m M sesand E-!cena n,m Rc! case from these esents shall be induded a'.d euhr row of ta'l pms 'had be av mned to N re' eat ! n t i ik a!ed under routine rekases :n nterJante with pr, p oed w ater. ((;ap actnity a l' ' of total. s t wit) m a [m )
App ndix 1.
(b) ()ne wet k decay tm hefore !! a;odent '> vs 5: Offdesir; trannen s that induce fuel fadrire fallbe a m med.
abMr those crpet led Jr'J s! cam ger:crJ:or !cE (such as tiow blockar ana Hux nu:distob atmns).
(c) Imha Jctontanun A m ! Ator m %Jtef sha'l h 500.
(a) 0 0: ' ef the core inventory of rn ble gases and 0 01' of the o,rc insentory of hal,u n, shah be id) ( har i.i I;!:er e f fister cy rio i, !nics shan he anumed to be releiscJ into the reactor coolmt.
W:
(b) Aserage msentery in the primary sy stem prior to (c) A reshstic tricti m of in sont.nnmm< ' volamt the t ran sten t sha:1 be based on operation with 0 5 shall be arameJ io leak to the atmosphere pnor to faded fuel.
- n. Lt ny th mnta.n n ent.
7.1-A-3 00 Pl4
/
YaU
February 1979 (f) Meteorology assumptions: t/Q salues shall be 7.2 /kary ob/cc t drop onto fuci rack.
1/10 of those given in Al C Safety Guide No. 3 or 4.
(a) The gap actnity (nob!c yases and halogern) m one (g) Conwquences shou 1J be calculated by weightmg awrag fuel assembly sLdf be assumed to be re! cased the ef fects m dif ferent ducctions by the f requency the into the water. (Gap actaity s 1" of total actmt> m a i
wmd blows in each duection.
pin.)
(b) 30 day s decay tims bef ore the accident oums 0.2 #cJrf obget drop onto /ur/ /n core, shall be assumed.
(a) The gap actiuty (noble gases and halogens)in one average fuel assembly shall be assumed to be released (c) Iodme decontanunation factor m water sh:dl be into the water. (Gap actisity shall be IS of total actnity 500.
in a pml (J) Charsoal litter ef ficiene for iodmes shall be (b) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of decay tune before object is dropped W 1.
shall be assumed.
(e) Meteorclogy assumptions xfQ va!ues shall be (e) lodine decontamination factor in water shall be 1/10 of those gnen in Al C Safety Guide No. 3 er 4.
500.
(f) Conwquences should be cakulated by weightmg (d) Charctal filter efficiency for ioJmes shall be the effects in J,tferent ducctions by the freque.,cy the 997 wand blow 3 in each Jirection.
(e) A reahstie fraction of the containment solume 7.3 Fucicask drop.
shall be assumed to leak to th( atmosphere prior to isolatmg the containment.
(a) Nob!c gas gap activity from one fully loaJed fuel eask (120. day coolmg) shall be assumed to be released.
(f) Meteorologwal assumptions: xlQ values shall be (Gap actiuty sha!! he l'1 of total actmty m the p;nsl ill0 of those gnen in Al C Safety Guide No. 3 or 4.
(b) Mein
.ogy assumptio is \\f0 salues shall be (g) Conwq iences should be calculated by weighting 1/10 of those gnen in AFC Safety Guide No. 3 or 4.
the effuts in ddferet ducctior.s by the frequency rSe wind blows ;n each diredion.
(c) Conseg ier.ces should be calculated by we:ghting the effech in d.fferent ducetions by the frequency the ACCIDENT-7.0 SPENT FUEL lixNDLING w:nd blows in each direction.
ACCIDENT ACCIDENT-8.0 ACCIDENT INITIATION EVENTS 7.1 Furlasscenb/r drop in /hci storage pool.
CONSIDERED IN DESIGN IMSIS EV A LUATION IN TiiE sal ~ETY (a) The gap activity (noble gnes and halogens)in one ANALYSIS REPORT row of fuel ;4 ins shall be assurned to be released into the water. (Gap actiuty shal! be l'~ of tota! aetnity in a pin.)
8.l Loss-of coolant accidents w/l Ape Brcal /6 irt or less)
(b) One week decay time before accident occurs shall s
be assarned.
(a) Souce terni the aserace rad:oadmt> msentery (c) lodine decontamination factor m water shall be in the primary coutant shall be assumed. (This nnentory 500.
shall be ba,ed on operatun with 0.50 faded fuelh (d) Charcoal filter efficiency for tod nes shall be (h) I:her ef ficiencies shall be 950 for internal ideis 99q and 99 ; for exterr.al li!ters (e) Meteorology assumptierts: 6Q valaes shall be (c) 50 buldmg mixir4 for boding water reastors
!!10 of those gnen in Al C Safety Guide No. 3 er 4 shall be assumed (f) Censequences shall be calculated by weighting the (d) For the effects of Plaicout. Sprays. Decentami-effects in different di:ections by the frequency the wmd nation Factor m Pooi, and Core Spray s. the fo!iowmg blows in each duection.
reduction fattors sha!! be assumed
-7 k
7.1-A-4
February 1979 Eir picssuri;.
../cr reactor, J 05 with hemical (a) The pnmary coolant insentory of noble gases anJ additnes in spre 0 2 f or no chenneal additives.
halogens shall be based on operation with 0.5 faded fuel For b J'ng war
. tors a 2.
(b) Release rate through failed hne shall be assumed (c) A realie:t buildmg Icak rate as a function of time wnstant for the four-hour duration of the acudent shall be assumed.
(c) Charcoal tilter effiaensy shall be E (f) Meteorology assmnptions: \\ O sal es shall be
/
1/10 of those given in AEC Safety Guide N,3 or 4.
(d) Reduction factor from combmed platenut anJ budd:ng nudng shall be U:1.
(g) Consequences should be calculated by weighting iie effects in different directions by the frequency the (e) Meteorology assumptions 1.Q values shall be wind blows in each direction.
1/10 of those given in Al C Safety Guide No 3.
O) Consequences shall be calculated by weighung tht effects m ddierent duettions vy the f requency the wmd Large Are Brcak blows in eat h direction.
(a) Soune ternr The aseraec radioactivity insentory RM Rod c/cction accihnt (piessun/eJ w ater reac in tne pnmary coolant shall be assumeJ. (T his unentory
,g,3 shall be based on operatioa with 0.5S f.nled fuel), plus release into the coolant of.
(a ) 0.27 of the core insentory of noble gases and halogens shall be assumed to be relea sed mio t ! n.
E>r pressuri:cd water reactors 27 of the core inventory of halogens and nob!e gases.
primary toolan t plus the aserap ms en tory in the rnmary coolant based m oper ition with U.Y faile d F >r boiling water reactors 0.2C of the core mven.
I"'L tery of halogens and noble gases.
(b) loss-of-coolant accident occurs with break, /e (b) Fiher efficiencies shall be 95? for internal filters equna!cnt to diameter of rod houwg (see assun:ptions and 99'1 for external filters.
for Accident RI).
(c) 509 building mamg for boihng water reactors O b) Rod drop ac cident (boilmx watcr reactor /
shall be assumed.
RJdMactive mJtcilJlic/ cased (d) For the effects of Plateout, Contamment Spray s, f a) 0.025~ of the con mventory of noble gas and Core Sprays (va!ues based on 0.5'T of halogens in organie 0 025% of the core inventory of halogens Nil be form), the following reduction factors shall be assumed:
assumed to be released into the coolant.
For Pressuri:cd water reactors 0.05 with chemical (b) !? of the halogens in the reactor toolant lall be additises in sprays,0.2 for no chemical additnes.
assumed to be reiesed into the condenser.
For boiling water reactor 0.2.
(c) The methanical vacuum pump shall be assumed to be automatically isolated by high radiation signal on (e) A realistic building leak rate as a function of time the steamline.
and includmg design leakage of steamline valves in BWRs dall be assumed.
fd) Radioactnity shall be assumed to carry over to the condenser where 101 of the halogens shall be (f) Meteorology assumptions: SQ values shall be assumed to be available for leakage from the condenser 1/10 of those given in AEC Safety Guide No. 3 or 4.
to the enuronment at 0.53/ day for the coarse of the accident (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).
,g) Consequence, should be calculated by weighting the effects in different directions by the frequency the (e) Me teor ology assumptions: \\iQ ulues shall wind blows in each direction.
be 1/10 of those gnea in Alf Safety GuiJe No. 3.
8.!(a) Brcak in instrument linc #om primary system (f) Consequences should be calculated by weightmg that penetrates the containment (lines aot provided with the etfects in different ir tyns by the frequency the isolation capabihty inside containment).
wmd blows in each direc a V O'O
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bJV 7.1-A-5
February 1979 0
8.3(a > Steam!me breaks (pressuri:cd uater reactors (b) Blow duw n h) 10ypm outsik contaimnent) Breal si:e equal to area of'safi ty ralrc throat.
(d) Volume of one steam generator shall be awumed to be released to the atmosphere with an iodine partitum factor of 10.
Small break (e) Me teorology assumpi.ons x 'Q salues sha'l be (a) Pnmary coolant actnity shall be based on opera.
1/10 of those ynen in Al C Safety Guide No 4.
tion with 0.57 fai!ed fuel. The prinury sy stem conhibu-
- ion dunny the cours of the acudent slull be based on a (f) Consequenses shall be calctdated by weightmg the 20 gal / day tube leak, effects in d2ften
- nt ducctions by the hequensy the wmd blows in each Jaes tion.
(b) During the course of the accident, a halogen reduction factor of 0.1 shall be apphed to the pumary S.9 b) Sicam/mc brcaAs ibmimg water reactor!
coolant source when the steam cenerator tubes are cosered, a factor of 0.5 shall be use[1 when the tut,es are SmaJ pipe break (of 1/4 f t - /
uncosered.
(a) Pnnury cool.ut actnity shall be based on opera-(c) Secondary coolant system radioa tnity poor to tion with 0.5 : failed fuel the auident shall be based on (p The nu steamhne shan be assumed to fad.
(a) 20 gallons per day ponury-to-secenJay leak releasing on' ant until 5 seconds af ter isolation s:enal is rec *. is ed.
(b) Blow Jown of 10 ppm.
k) llalogens in the fluid releavJ to the atmmphere (d) Volume of one steam cenerator shall be reicased to the atmosphere w1th an iodme partition fator of 10 (d) Meteorology assumptions x!Q salaes shall be
!/10 of those in Al C Safety Guide No.1 (e) Meteoroloey assumptions. 1/Q ulues shall be
~
I/10 of those gnen in AEC Safety Guth No.4.
(e) Consequences shall be calculated by weyhting the ef fects m d:fferent directions by the frequeno tfu
~
(f) Conwquenus shall be ca!culated by weightmg the wmd blows m eas h duection.
effects in dif ferent directions by the fr;quency the wind h!ows in each directwn.
/
c bred (a) Prinury coolant
'y sha!! be based en opera-large breal tion with 0.5 l faded fta.
(a) Pnmary coolant actaity shall be based on opera-(b) Main steamhne shall be auumed to fail. re!casin >
tion with 0.57 faded fuel The pomary system tontnbu-that amount of coolant correspondme m a 5 seconJ3 tion dunny the course of the accident shall be based on a rolation tune 20 gal' day tube leak.
(c) 50 of the halogens m the flu:d eutmg the (h) A halogen reduction fastor ef 0.5 sha!! be apphed break shall be assumed to be re: cased to the atmosp'me.
to the romary coolant source during the course of the accident.
(di Meteorology assumpnons x Q sahes shall be 1:10 of thew m Al.C Safety Guide No. 3 (c) Secondary coolant sy stem r dicactnity prior to the accident shall be based on:
(e; Consequences shall be calculated by weyhnng the ef fects in dif ferent duections by the frequency the (a) 20 gallons per day prinury-to-secondary leak.
wmd blows m eash ducction b,
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7.1-A-6