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DISTRIBUTION DOCKET FILE AEB?R/F W
WPasedag Plant File Docket No.: 50-424/425 1 3 1934 AD/RP RF MEMORANDUM FOR: Thomas H. Novak, Assistant Director for Licensing Division of Licensing FC0H:
Daniel R. Muller, Assistant Director for Radiation Protection Division of Systems Integration
SUBJECT:
SUPPLEMENT TO THE SAFETY EVALUATION REPORT FOR THE V0GTLE ELECTRIC GENERATING PLANT t
In the safety evaluation report (SER) for the Vogtle Electric Generating Plant dated November,1984, the staff inadvertently did not provide Sections 6.5.2 and 6.5.3; " Containment Spray as a Fission Product cleanup System" and " Fission Product Control Systems and Structures."
AEB finds that the containment spray system inspection and testing plans and proposed limiting conditions of operation for the spray system provide adequate assurance that the iodine scrubbing function of the containment spray system will meet or exceed the effectiveness assumed in the accident evaluation, and that the containment spray system as a fission product cleanup system is acceptable. This conclusion assumes that the CSB determines that the spray additive tank and its downstream valves and piping are able to meet Quality Group B requirements.
Enclosed are SER Sections 6.5.2 and 6.5.3, covering these topics.
i Co.b is! Qned by Dar.lal R. Mu!!cg Daniel R. Muller, Assistant Director for Radiation Protection D.1 vision of Systems Integration
Enclosure:
As stated cc:
R. Bernero R. Houston W. Butler E. Adenum W." T. Hiller b //P/0 Y8D "gg:A gE D
AEB DSI:AD/RP DSI nes)
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T 'T;r x 22P" 6.5.2 Containment Spray as a Fission Product Cleanup System The containment spray system consists of two redundant and independent trains powered from separate sources independent of offsite power. Each of the two containment spray pumps has a design flow rate of 2600 gpm.
All active components of the containment spray system are capable of being tested during plant operation. The two containment spray recirculation intake pipes take suction from two separated containment emergency sumps. The sump intakes are protected by trash guards and fine. mesh screen's from debris that could clog the spray nozzles. The containment spray system is automatically actuated by a coincident two-out-of-four containment (,high-3) pressure signal. Operation of the spray system may also be manually initiated from the control room. The spray system will initially take suction from the refueling water storagetank(RWST)whoseeffluentismixedwithNaOHfromthespray additive tank. When the RWST reaches a low-low level alann, a switchover from injection to recirculation will be initiated manually.
The total amount of borated refueling water injected into the s
containment by the charging, safety injection, residual heat removal, and containment spray pumps will provide a conservative sump pH of 8.5 i
when mixed with ~the contents of the spray additive tanks.
The spray additive tank is sized to ensure that elemental iodine will be mostly l
removed by the injection spray and retained in the sump during the recirculation phase. It has a 4000 gallon capacity which can provide 30 weight percent NaOH concentration, which will be exhausted in about 100 minutes, when a conservative sump pH of 8.5 will exist. However, the airborne iodine concentration within containment will be sufficiently T '- - - * - - ~
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- 3 ~ + - a depleted in accordance with SRP Section 6.5.2, to cieet the spray system design objectives after less than 8 minutes of spray injection operation. The containment spray system has been designed to allow periodic inspection of the components and functional testing to assure the operability and perforce of the system.
SRP Section 6.5.2 states that the containment spray system should be designed to transfer automatically from the injection mode to the recirculation mode to ensure continuous operation until the design objectives of the system have been achieved.
In all cases, the operating period should not be less than two hours. However, tne staff may find manual switchover to be acceptable if, following a potentially severe accident, there is assurance that manual actions are not required in such a short time that the operator (s) may not be able to complete them. The containment spray design for Vogtle is acceptable with respect to this concern, because only two operator actions are required, and there is a sufficient margin of volume in the RWST after level indications alert the operator to initiate the switchovers.
The fo11cwing parameters can be used for a conservative thyroid dose calculation for the postulated LOCA: an overall first-order removal constant of 10 per hour for elemental iodine and 0.45 per hour for particulate iodine, an effective spray volume of 2.75 million ft, and a maximum decontamination factor of 100 for elemental todine.
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O The staff concludes that the containment spray system as a fission product cleanup system is acceptable for consideration as an engineered safety feature in the post LOCA dose calculations outlined in Section 15.6.5, and meets the relevant requirements of GDC 41, " Containment.
Atmosphere Cleanup"; 42, " Inspection of Containment Atmosphere Cleanup Systems"; and 43, " Testing of Containment Atmosphere Cleanup Systems."
This conclusion is based on the following:
The concept on which the proposed system is based has been demonstrated to be effective for iodine absorption and retention under postaccident conditions. The proposed system design is an acceptable application of this concept. The system provides suitable redundancy in components and features or that its safety function can be accomplished assuming a single failure. The staff concludes that the system meets the requirements of GDC 41.
The proposed preoperatianal tests, postoperational testing and surveillance, and proposed limiting conditions of operation for the s
spray system provide adequate assurance that the iodine scrubbing function of the containment spray system will meet or exceed the effectiveness assumed in the accident evaluation. Therefore, it meets the requirements of GDC 42 and 43.
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-1 6.5.3 Fission Product Control Systems and Structures The review for SRP Section 6.5.3 was performed as part of the review for Sections 6.5.1 and 15 of this report. See those sections for a discussion of the review.
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htomas M. A)ovak., Assista Diyaeror for A./e u so's.g Divisin.
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IMPUT FRoM THE ACCIOEAfrts! alt-MTibd B4AMcH
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a-ENCLOSURE 1 DRAFT SER INPUT FROM THE ACCIDENT EVALUATItti BRANCH FOR V0GTLE UNITS 1 AND 2 15.4 Radiological Consequences of Accidents
~.g,h6.Ss&1lv4lt t<cksha Mn. bh" W"?.o. *w tw p.-4 The pe;ttthted detiga ba;b ;; cide c :ns!yzed by the-appitcant-to-determ4ne fuz).'en_
the-offsiteradiological-consequeba= mea the n=== as thans ana.lyz-d fee-
[ ay'N w
p evh siy-tfeensed PWRs.
To eval ate the effectiveness of the ESFs proposed
- 'd %I
- s. b for the Vogtle plant and to ensure -thet-(tWe}radiologica1 consequences _Qthese,_.,
!.CCA d '.( nus,.;c,e, v-
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accident rer* tha
=ng).icable-dose-criteri:,- the.:t:ff hr/.:.;.1"nd :
s'eln+eet,t'o %:ndlin ended a=w. ! de_i n,w? tes...! 4 _.speciVc cit.>-ip {c=koves of sL4.
i pe accidents
';. -h
.t b ed _
- stese ge.:r; tor..
es
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.c. Som erso%ror k.v.s.>e of eckor ~, c
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rs as
- -tu acc d t.,
211 line-f aWres,-and a.controt rod-ejection ac enk..
M to.ha,eout & & A neken e a erra'.s &,ss, %
- tr pet.!tien t-et Ehebrr.+.ds sueci<<kA won.ruri.ssa.a :c.c "u r vs,"
": = := *it.: th:
tandard-Technica l-Spec.14tca t4 oh N '
N= $bt:f
.= :=? =t:f.'he red?:?:;h;! cen.,vences=for-acefdent.
fhav.iDg_ralaases thre 4. ;.h..m.vnJo, ;y:t-, as,suming the-ve he; given in the Westinghous/e Standard Technic'al Specifications 'for primary and se,co'ndary coolant ' activity.concentraliene and p-4eepydo-secondary leakage.except -where
/
etherdse-stated: The/ staff will review,the proposed Voatie Technical Specifi-J
[ca(fons to ensure that these limits are rnet) The calculated doses for these accidents are given in Table 15.1.
15.4.1 Loss-of-Coolant Accident n.1 ih wS.42 3
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':-4 the distances o the ex ion area.bo ndar nd h the 0
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,w t. m o ae-o orth in,10 CFR 100,11(a)$$$::amit:42k The,a.nalysis has included guidtf e
the.S'k h; sources and radioactivity transport at..py:
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.w s *.Ma Leu s( C J.d8slLt, I.4,h@e "nssAtsi.Cu Acs.hs4 do L h.u le. r,% wT. st *.*.l.%.tl 1) from containment,
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" W '. % d' Pd **W2.i. 4. s'/**'* ' * " c 4 V0GTLE SER SEC 15 INPUT
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- t '^^^ leakage from ESF systems outside containment
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(1) the icant'sprovisionsfojr nd design of the con ainment system, and the captability of the au fliary building exhaus system as described '
i Section 6 of this repor
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(
the staff's indepe analysis of the ra ological consequencesyf a hypothetical design-basis LOCA as descri ed below.
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a Containment Leakage t : trit.ti = % O w g 15.4.1.1 s.t My hb dhtk TheVogtleplantinclude/acontainmentdesign,4tognafzet{eleakageof 3
4 g,
'^C?..f The containment consists g y' fission products from a postulated de:fgr. in i:
of a post-tensioned concrete primiry containment vessel with a carbon steel hsh r:-
t liner. Another engineer safe feature ESF) is the containment spray clai.u &
o.e. ns p.i.w,~
ac w e a cua taasic 64 in w w e c.i system with an NaOH addi ve to.r.t.xx th
- tuc, yin micffer in the containment q n.i a'
- 7- -
ollowing a LOCA. The staff's calculation of the consequences of the hypothe-C tical LOCA used the conservative assumptions of Positions C.1.a through C.1.e of Regulatory Guide 1.4, Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressur-t ized Water Reactors." The primary containment was assumed to leak at a rate of 0.2 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 percent per, day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, j in raction jof c5frinvento ' available f release was assumed tyi>9 e
Qper t for iodini and 100 pe
_ nt for nobl asses./~The analysis took into e
account radiological decay during holdup in he cont neent mixin n the caiu rv,-
4 og m
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containment, and iodine decontamination by the ESF spray system st of e,$.w.k A 3
t
. n s-t assumptions used in the calculation of the LOCA doses is given in Table 15.2.
15.4.1.2 Post-LOCA Leakage from ESF Systee Outside Containment s pa'r't the t0CA, the staff has.also evaluated. the consequEd'eTo'f'Teakage]
tainmentsump,wate,r_which.iscirculatefb(theECC er that post 01ated
~
sf c 3 __ ent]'Duringtherecirculationmodeofoperation the sump water is acc 4-hs. GCcs follw& n. f 0CA,
09/18/84 2
V0GTLE SER SEC 15 INPUT i
Q c,4,'nh4 g..'.,{ceFes uk 4z Eus w G-in:,
S No;
.f., p circulatedouts{decontainmenttotheauxiliarybuilding3 u-a leak 5rseuid Q 2-g-:.
N:
uc b s Y p $ al faiYu 7 "$*YS:tf^r. C' th-4 dina
+h:.s te.
n Ngo$1 "airborn EtNe7xNa ui1 fin.N'IfltYN;YN[pN
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For Vogtle, the ECCS area in the. auxiliary building is served by an ESF air e t Tkmwherc.w O'*"2 r%',
wkh,k ahe upeM re coded-M filtration { qu,, tem (the au; ciliary /Is ubuf1 ding exhaust system)6 refore, doses M A~t sys q
e ms4 emu o n* n s a
from passive rat iurepere notYconsidered (as specified in SRP Section 15.6.5, Appendix B).iw our calcul4[/,
Ph In FSAR Table 15.6.5-4, the applicant has identified a value of 50 gal / min as the maximum amount of leakage from ECCS equipment following an accident.
Following the Standard Review Plan, the staff evaluated the potential radiolog-ical consequences from this release pathway assuming a leakage r g ce the applicant's value.
Theresultantradiologicalconsequenc(were71 rems to the thyroid at the exclusion area boundary and 148 rems to the thyroid at the low population zone (LPZ).
The staff considers 50 gal / min to be a larger
- coM than necessary leakage value, which saa be substantially reduced. Th.sppGwh 4 cmvmWee s~.A m rper ~n,l, f, ie,.4M.,w b mb/,qe. 4 44 /eyr.,k menAg Conclusio, In eap
.s 15.4.1.3 ns
<hs
~ Om CTh: :tc.ff's)calculatedthroi and whole-body doses from the hypothetical LOCA we.3 are given in Table 15.2.
The atoy conclude / that the distances to the exclu-sion area and to the LPZ boundaries of the Vogtle site, in conjunction with the ESFs of the Vogtle plant design, are sufficient to provide reasonable assurance that the total radiological consequences of a postulated LOCA will be within the exposure guidelines set forth in 10 CFR 100.11. This conclusion 42.-
is based on the staff review of the applicant's analyses and on m independent analysis performed h the ;tof f-to verify that the total calculated doses are within the guidelines.
15.4.2 Main Steamline Break Outside Containment Both the staff and the applicant have evaluated the radiological consequences of a postulated steamline break accident occurring outside containment and upstream of the main steam isolation valve.
Although the contents of the secondary side of the affected steam generator would be vented initially to 03/17/84 3
V0GTLE SER SEC 15 INPUT
.--,m---
y-m --
theatmosphereasanelevatedrelease,Q::tch=conservativelyassumed that the entire release throughout the course of the accident occurs under ground level conditions. During the course of the accident, the shell side of the affected steam generator was assumed to stay dry since auxiliary feedwater flow to the affected steam generator would be blocked off under the conditions of this accident.
Because of the dryout condition in the affected steam generator, all iodine transported to the secondary side by leakage (1 gpm) was assumed available for release to the atmosphere with no reduction from holdup or attenuation.
M N
M p c t. " P.;... N _ ;tr' nvestigated three scenarios.
For Case 1, the most reactive controi ivu was assumed to be stuck in the fully withdrawn position. The. applicant has indicated, and the staff agrees, that no departure from nucleate boiling is expected to occur and, therefpr.e4, no fuel cladding e
wu ab occ urr failure h 0; he assumed in the calculation. With no fuel failures prese Ma Case 1 becomes identical to Case 2, and no radiological consequences are presented for Case 1.
s le p m i m w hsihn. u de.~ % W M 4.</* 4 l
For Case 2, the staff assumed that (iodinespike)occurredasaresultofthe d;
pnd pressure transient caused by the accident.
power Before the accident,fiv.
was assumed to be operating at the Westinghouse Standard Technical Specification equilibrium primary coolant limit of 1 pCf/gm dose equivalent todine-131 (DEI-131).
The iodine spike generated during the accident is assumed to increase the release rate of fodine from the fuel by a factor of 500. -This increase in the release rate results in an increasing iodine concen-tration'in the primary coolant during the course of the accident.
The radiolog-ical consequences for this case have been calculated using assumptions given in Table 15.3 and the consequence values are given in Table 15.1 of this SER.
For Case 3, the staff assumed that previous reactor operation had resulted in a primary coolant concentration equal to the maximum transient full power Westinghouse Standard Technical Specification Ifmit (60 pC1/gm DEI-131).
As in Case 2, the radiological consequences were calculated using assumptions found in Table 15.3 and the consequence values are given in Table 15.1.
09/17/84 4
V0GTLE SER SEC 15 INPUT
__.a d-
- c.. u ;. 3 -
l-Based on.its findings, the staff concludes that th: di:t:n :; t -th:
xc!u= N
+ hue.:t t
-s..e nd '_P2 bee @-*er 'e-
- Vantio dte ar-e.-suf-f-icient t; pra id f reasonable assurance that the calculated radiological consequences of a postulated mai steamline failure outside the containment of the Vogtle plant dcm lo e amas.t 6f **::*d
- we nta ar &b dec.
er -
A
~%
(1) the empeewe guidelines :; ; t f;+rth in 10 CFR 100.11 for the case that the failure occurs with a primary coolant iodine concentration corresponding to a preaccident iodine spike; and (2) 10 percent of these er;:; re guidelines i
for the case that the failure occurs with a primary coolant aptivity corres-ponding to the m M' equilibrium concentration 'e cc-t4-8E fuP pe"cr
- catica
- :6ebe+-+n the Westinghouse Standard Technical Specifications.
Jne staff tuncluderthatTe roposWdeslgn and 6peraffiin7fJthe Vogtle plan isefjfet eincontrol.lin'gtherelease,of'ffss_fonproducts'followingaposIu-Ia_ tatuain steamUn_e[ break accident.f w
& Mr H. +kL These conclusions,are baseM(1) th :t:" review of the applicant's analysis d on
+ Q ou W::! :: :: don cvr b
of th:
e4eff-using appropriathr:q::n ::, (2) the4 ndependent dose calgulationg, the i
rc.....,.
ad conservative atmospheric ^
g;10t:ry ;;; r;tf:::
y 4
diffusion factors as discussed in Section 2 of this report)and(3)thespecific v
Technical Specifications for the iodine conckntration in the reactor coolant
[which consists of a maximum allowable limit and a limit for the equilibrium concentration for continued plant operation)and the '!:;ti limit on primary-to-x secondary leakage in the steam generators."Th de[f 4WN h Va M p s n tec W ry s f u }dc 4 lo emure fl+l- +l'tJ* of M*SyMSh *
- w,pa w.
15.4.3 Steam Generator Tube Failure The applicant has provided an analysis of the systems response and radiological consequences of a steam generator tube rupture (SGTR) accident. The staff has requested justification for the asserted ability to isolate the affected steam generator within 30 minutes and for the assumed mitigative capability of systems to reduce the radiological consequences of the accident.
In response to the staff request, the applicant states that the Westinghouse Owners' Group is investigating several SGTR licensing concerns and will address the staff's con-cerns through a generic resolution in late 1984.
Upon receipt of this addi-tional information, the staff will complete its review of the SGTR event and the radiological consequences thereof. Tlu..dc[M Mw M *c0*b
(< upcM M A [+Jw - c~ph f h, 4&h.f44.
e 09/17/84 5
V0GTLE SER SEC 15 INPUT j
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.. g 15.4.4 Control Rod Ejection Accident A nonmechanistic rupture of a control rod drive housing is postulated.
Because of the resultant opening in the pressure vessel, primary coolant is lost to the containment with concurrent rapid depressurization of the reactor pressure it oe.tv. sad Io vessel.
Reactor trip, initiated by one of several trip signals, occurf rapidly.
Ejection of a control rod results in rapid reactivity insertion.
The applicant has conservatively assumed that 10 percent of the fuel elements. will experience wre..we s/-4,44.
m nMTe he,ewh im +u Aci-M.t e
cladding failure, releasing P.h:fr gapo :d!:::tivit9 Inaddition,0.25 r
4,
percent of the fuel rods wer: d m=:^rvativ;' celculet;d to^
erfence fuel taals rdem
-mut soienfeased J;dimt!.ity b m h;sas, ~ ge u Ha m A r,o melting.
e re f-df_.:b wi the primary 3
coolant. The s*: " ::: =:d thet release to the environment may occur by get. c, J g [...ava p rlu.c f
either of two pathways.
The first pathway involves a release ofgetivity'to the primary containment, which is then assumed to leak to the atmosphere
.Af the design 4:0E*r( nr. endeusw& (0.2 p*m.4 esr
.'..r
.44. w k. s In t e second athway,an a.ivi trar-f:rr:d frc th^ reuhak, act s %
~m 4 m4mm A
prf::r-- te the second cgo ant via M Mr :d 1 gpm pr mary-to-secondary fs ac a nsol(m t.
~
,ms ul ums A
leak rate With loss of ffsite p wer and s bsequent steam venting, some of g
j the iodine transferred to the shell e is available for leakage to the
(.iisos.wd b accur ea a emu'/,[ m< H/b >M 4N environment.
M In considering the consequences of this, postulated event, t-calculated
.focm M-ava. mL W upar y war; t ta!!y releasgby way _grr-the doses C' activi each of the above pathways.
- Nf ou d expect the actual consequences to be some combination of these pathways.
The assumptions used in calculating the radiolog-ical consequences are presented in Table 15.4, and the resultant doses for each pathway are given in Table 15.1 of this SER.
GW SO 9% Q h; :t ff b reviewed the applicant's analysis of the radiologi al consequences uv c.
followingapostulatejcontrolrode etion accident.
The eieff conclude;/
W :[cr t
Na h.t((h b_.50. e_M. b_._ ;_;_;" 4#N...d.., d,, M.
thatthediet$rT;t A
the LDZ M"ndef ae-br Vogtlep A~h
{nennj,M[.Nn"n 4
y n
~"^" 'a providej reasonable assurance that the ca r"htad radioinnical
- guidelines (lessthan25 percent)h consequences are well with 24 set f;rth in 10 CFR 100.11.
09/17/84 6
V0GTLE SER SEC 15 INPUT
~.. ~,-.
i O
ove~
,h; ;;; "'
conclusion is based on (1) review of the applicant's analyst the radiological consequences, (2) C, h h independent dose calculattogusing the recommendations of_ Appendix 8 of Regulatory Guide 1.77 L
and the atmospheric dispersion factors as discussed in Section 2 of this report, and (3) the Westinghouse Standard Technica: Specifications for the primary-to-secondary leakage in the steam generators.
15.4.5 Fuel-Handling Accident l
In the evaluation of the fuel-handling accident, the methodology used by the staff is based on Positio through C.1.f of Regulatory Guide 1.25 and
(:7 assumed that a single fuel assembly is dropped SRP Section 15.7.4.
l into the fuel pool during refueling operations and that all of the fuel rods f
in the assembly were damaged, releasing radioactive materials in the fuel gaps into the pool.
For the case of a fuel handling accident in the fuel building, the applicant estimates the time for the radioactive materials that escape from the pool to travel from the detector to the isolation damper to be slightly over 0.6 second. The closure time of the isolation dampers in the normal exhaust system following receipt of an isolation signal is estimated at 6 seconds.
Because the travel time is less than the isolation time by over 5 seconds, K M
@ a^M^-"t assgd t tfheentireactivityreleaseescapesfromthe fuel building without r d t N.- ESF filtr;t h ;&r4 In the case of a fuel handling accident occurring inside containment, 25 percent of the containment volume ssumed for mixing with a containment isolation time of 15 seconds.
No filtration credit is assumed for activity that escapes prior to containment isolation.
',2a x Am h The offsite doses for the postulated fuel handling accidents inside containment A
and inside fuel building are shown in Table 15.1.
The list of assumptions and parameters used in the analysis are given in Table 15.5 a/b.
The potential doses for the fuel handling accidents are well within the guideline value given in 10 CFR 100.
Therefore, the staff concludes that the applicant has provided a design that meets GDC 61.
09/18/84 7
V0GTLE SER SEC 15 INPUT
.l
With regard to a potential spent fuel cask drop accident, a Type 1 single failure proof crane designed according to NUREG-0554 and Branch Technical Position APCSB-9-1 is used in handling the cask, which is prevented. ) inte g lockg,Jrombeingmovedoverthefuel.
With these safety measures,D ct@%
Qonclucm3 that the likelihood of a spent fuel cask drop accident is sufficiently c.
small that no radiological consequence analysis is required.
15.4.6 Failure of a Small Line Carrying Primary Coolant Outside Containment Tk<. ns a u.dk of wil % e+yy p;ng a.M w,b. % ce,In%J.
gheapplicanthasprovidedananalysisofanaccidentalbreakintheCVCS containment but dnynst letdown line outs M & M 4, Wsw M. ream of the containment isolation m a swsr-cue valvesf This bre k would release 194 gpm of primary coolant to the auxiliary building before isolation could be expected. pr;. L:nv o....== youn,y.o
+h-
- i r=;;;nb.
The break would cause a low level in the volume control tank, and the operator could diagnose the break and shut the appropriate isolationvalveto(sglatetheleak.9Thestaffhasperformedanindependent assessment of the oseconsequencesofM EN
'W j
wilI sLr h
9Tj etmyammet that 20 min k r:ptc. ke[orefa; ptofthelow-levelsigna1h
~
[
and/s6eraidr action to ism [ break3f g gtotal of 3880 gal of primary Th coolant could be released 4'Qff estimat 39 percent of the hot reactor coolant would lash into steam unon entering the auxiliary butiding M npwfon,eJraction of the*dissolvedcJijd,ln)**er *:m/n, Atud.t f atmosphere, and assum aa aq"d i
sbecome[airbornebgas articulate In the absence of ESFs designed to detect and mitigate the consequences of b release /, W ct:f assumefthatthisairborneiodinecanescapedirectlytotheenvironmen at ground level, without delay or effective filtration.
Other assumptions are given in Table 15.6.
The staff concludes that the dista::: te
- k- =uerfer r:: ad te L M population-20na -outer, heenderlas_,go vggt7: ;p,_ggg,gg,g,,, _,,,,,,u g, ble
- ! r ace th. A s. ko b;..d c;di; h;i;:? consequences of a postulated small line failure outside the containment, assuming the primary coolant equilibrium iodine concentration permitted by the Standard Technical Specifications, in combination with an accident generated iodi spike, do not exceed a small fraction of the exposure guidelines :: cet..rth i 10 CFR 100.11.
The results of the staff's calculations are given in Table 15.1.
09/17/84 8
V0GTLE SER SEC 15 INPUT
+ a. m._
_ _ _ r.z _=.
.r 7
-- - c=- - m - yy-o
-M
+ "'s onclusion is based on (1) review of the applicant's on and identification of small Ifnes in accordance with GOC 55, c
"Rea:: tor Coolant Pressure Boundary Penetrating Containment," and Regulatory Guide 1.11 " Instrument Lines Penetrating Containment"; (2)& cuccM O
review of sS M of the applicant's analysis of radiological consequencesj;73)gindependent dose calculatforpy 3....Tf using Regulatory Position C.1.b. of Regulatory Guide 1.11 and conservative atmospheric dispersion factors as discussed in Section 2 of this report; and (4) the Westinghouse Standard Technical Specifica-tions for the equilibrium iodine concentrations in the primary coolant system.
The staff will review thg,Wgj[le specific Technical Specifications to ens that the 'A 8bNb:t:t:
boYe'arenotexceeded.
~
6.4 Control Room Habitability The requirements for the protection of the control room personnel under accident conditions are specified in Generdi Design Criterion 19.
The applicant proposes to meet these requirements by incorporating shielding, emergency heating, ventilating and air conditioning (HVAC) systems, and self-contained breathing apparatus in the control room habitability design.
The habitability systems also provide storage for food and water, sanitary factif ties, and fire protec-tion that includes a remote shutdown capability.
The design of the control room habitability systems relative to the following areas is discussed in separate SER sections as indicated:
Explosion, fire and toxic gas in vicinity of plant - Sections 2.2.1-2.2.3; a.
b.
Protection from wind and tornado effects - Section 3.3; c.
Flood Design - Section 3.4; d.
Missile protection - Section 3.5;
(
o.
. Protection against dynamic effects associated with postulated ruptures of piping - Section 3.6; 09/18/84 9
V0GTLE SER SEC 15 INPUT l
~
m m.
m.
.z____,
.~
~.
...L.'.
. L.L L L:T::XX~~~W ~~~ "
f.
Environmental qualification of equipment - Section 3.11; g.
Filter efficiencies - Section 6.5.1; h.
Radiation protection aspect of GDC 19 and NUREG-0737 II.B.2; shielding; TSC - Section 12.3; 1.
HVAC systems analysis - Section 9.4.1 (includes seismic review);
j.
Fire protection and remote shutdown capability - Section 9.5.1; and k.
Human engineering, control room environment, and communications - Section 18.
The staff evaluation indicates an inconsistency in the applicalt's estimate of the control room leak rate.
Relative to toxic gas protection, the applicant states that the air leakage is no* greater than 185 ft / min from all pathways 8
based on 1/8 inch W.G. pressure differential. This confifets with the estimated 1500 ft8/ min air intake at the same' pressure differential assumed in the evaluation of radiation doses to control room personnel following design basis accidents.
Until this matter is resolved, habitability of the control room following radiation and toxic gas release accidents is an open item.
- In addition, information is needed from the app 1tcant in two areas:
(1) Response to Question 450.3 on the data used to estimate the control room dose following a LOCA.
(2) Toxic gas evaluation for the chemicals. listed in Table 7.2.3-18.
The evaluation should include data described in Table C-3 of RG 1.78.
A.d dW+ N
/
Based upon the foregoing,3the applicant has not demonstrated that the control room habitability systems will adequatcly protect the control room operators in accordance with the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 19 and, therefore, compliance with NUREG-0737, Item III.D.3.4 cannot be established.
09/18/84 10 V0GTLE SER SEC 15 INPUT
~
i,\\
,3
- t s
\\
C C
Table 15.1 Ratological consequences of design-basis accidents
.c J
Exclusto4:stis Low population
,N..
boundary dose, rem
- zone dose, rems
- PostulatedacNident Thyroid Whole Body Thyroid Whole Body i
Lo'ss of coolant:
Containe.snt leakage 0-2 hr i
98 2.6 0-8 hr
' 'a !
33 0.9 8-24 hr 12 0.2 24-96 hr 10 0.1 s96-720 hr 9
_ 0.1 Total containment leakage-98
- 2. 6 -
64
- 1. 3 ECCS component leakage 71
'0. 2 148 0.2 169
' 2. 8 212
- 1. 5 Total Steamline break outside E*
4; containment:
y Long-term operation case Short-termoperationcas%
(DEI-131 at 1 pCf/ge) l1.7
<1.0
- 1. 6
< 1. 0 e
(DEI-131 at 60 pCf/ge) 3.1
< 1. 0 1.5
<1. 0 Cont'rol rod ejectior.?
h Containment h akige patUsey 26
<1.0 43
<1. 0 Secondary system relssro s'
pathway 9.7
<1.0 1.8
<0.1 Fuel-handling accident in fuel-handling area 53 0.7 8
0.2 Fuel-handling accident inside containment
,0. 3
<0.1 0.1
<0.1
.b Small Ifne break
. i
'.:J 4. 6
<0.1 0.8
<0.1 a,
- The short-term diffusion estimett (X/Q's) used in the analysis are those presented and discussed in SER Sectfon,2.3.4.
The meteorological models described in regulatory guides refoh.nced in these analyses are modified by those presented in Regulatory Guide 1.145.
See Section 2.3.4 for further discussion of thrvateorological models.
" D E1.I5l
- 4L hst. se
%te t- $4.'sa.-I:I enwimi,'gn,.m it.Liul Ib h..[6edw.'e.J p c/h'c4'w 2 Ia 44.
i.
8 6
s Q a, % s%
09/18/84 4%
f V0GTLE SER SEC 15 INPUT o,
7.,a.p.
.y.....
+..
Table 15.2 Assumptions used in the calculation of loss-of-coolant accident doses Parameter and unit of measure quantity
} Power level, MWt Containment leakane N
3,565 Operating time, yr 3
Fraction of core inventory available for containment leakage, %
Iodine 25 Noble gases 100 Initial iodine composition in containment %
Elemental A
91 Organic 4
Particulate 5
Containment leak rate, %/ day 0-24 hr 0.2 After 24 hr 0.1 Containment volume, ft2
' Sprayed volume 2.15 x 108 Unsprayed volume 6.05 x 105 i
Containment mixing rate from cooling fan operation, cfm 174,000 Containment spray system Maximum elemental fodine decontamination factor 100 Spray removal coefficients, hr-1 Elemental iodine 10 Particular iodine 0.45 Organic iodine O
Relative concentration values, sec/m2
- 0-2 hr at the exclusion area boundary (E AB) 1.8 x 10 4 0-8 hr at the low population zone (LPZ) bound &ry 3.1 x 10.s 8-24 hr at the LPZ boundary 2.2 x 10.s 24-96 hr at the LPZ boundary 1.0 x 10.s96-720 hr at the LPZ boundary 3.4 x 10.s ECCS leakaae outside containment Power, MWt 3,565 Sump volume, gal 905,080 F4eek. fraction of fe4%c. es**d Madyd O*'-S'bd 0.1
- es umC 09/18/84 13 V0GTLE SER SEC 15 INPUT
._. -: ymmmy-~
Table 15.2 (continued) i
~
Parameter and unit. of measure Quantity ECCS leakage outside' containment (continued)
Leakrate,gph(tdicethemaximumoperationalleakagedefined in FSAR Table 15.6.5-4) 6000 Leak duration, hr 720 Delay time, hr 0.50
/
s Filter efficiency for iodine, %
Elemental and particulate 99 Organic iodino 99 W M & N c->&~5
( (
l s
I
)
5 i
b 4
h
+
k a
h A
e e
5
?
09/18/84 14 V0GTLE SER SEC 15 INPUT r.
n n - ~ _ _.,-
,.-n.,
~. _
p m,-.-
=,
._c Table 15.3 Assumptions used to evaluate the radiological consequences following a postulated main steamline break accident outside containment Power, MWt 3565 Preaccident dose equivalent I-131 in primary coolant, mci /gm 1.0 (Case 2)*
(Gse.S)
Preaccident dose equivalent I-131 in primary coolant, mci /gm 60.g**
Primary-to-secondary leak rate, as limited by Technical Specifications, gpm 1.0 4'
e'!
- h- - leak [4sses in the affected steam generatorgp>
l.O r-+.a
~
the iodine 'r:r;.. er..d td the shell side of the steam "1
n generator by th: leakage *., ?; r to the environment. !the t I'O
?MU bho of
- .ef pg:odinereleaseratefromfuelScr:re:
f;;ter :f 500dv<!g b
2)ie h ; spilu k rete n e m 4e b m
- .e.;ait of th: ;; cider,t (00;:
- 5. vet Auri.g.r W ds, sec/m( y cl a e m *u n 569 X
v hr 1097 E
\\R 0-0-8 at 8 m ( L, 3.1% 1 s
- Long-term operation case
- Short-term operation case 09/18/84 15 V0GTLE SER SEC 15 INPUT
~. c ~. m._. -
^ - ' ' - -
Table 15.4 Assumptions used for estimating the radiological consequences following a postulated control rod ejection accident Power 6 MWt 3.C,.f.
Primary-to-secondary leak rate 4lEEEED gpm e,1:..it;d t, Techn!::? S; r!'f::tica:. l' O y
L..McWos
..r - of the fuel rods expe,rience cladding failure, re!
'n; ;'l th !r A/
,;;;:--+4.,4+u r.en..a +m de a py 7:r::nt :f 13; ; ;1,;37;;-- : 7: --t; t, af fedfae: 2nd :51: ;::::).
TL. r;le..;d activit, is mixed i. n&te!;
"i+h the
- m. 4.m.,
---s 1
'prAfiek sKhMieMe Invedayin gap e{ Fl<d ed.s v./
Tygeb,ew n ?" ;--c-a+ of the fuel rods experienc/r"sfuel melting
=d t'^
"ah'a r es 0*002 ~
=ad " ;;rcent of th
- edi.- in th, fraction of fu;l f: r:12 ;;d nd #:
ind FretNSi'-;dinICIF Ji IN IEC pTI OIF 000lOGI. @f I# '"4'I"'#"b'M '4I48'd b* dI 8"fi#"N*$ " }
h
'C f'reekon *f.hM emis< hup S'.G.cewulag sdt. rele.* sed ha ess ho u 2 re:r'I w.
aus5 OT Qilssi.c puw. Gad ;^ w QU nt tre= unnting, ln an w--t gj u
Af the -iOd{n. t v.angpopted - ta-Ad4fMed '.Ji th-thG 5 dvndaag Cv6l-dut $5-Ival--
duc.iag4.he-Cour-SO Of the-SCcidefft.
Ntt, oh g'04)JdC.
grimary and secondary system pressures equalize in : tert ??00 ::c, + r-*n:+4ag 32c0 the pr'x c, t: :-
-t y ' 4 Trac 4ow of '.CC. P Ak.ol auf is I
Fe*-4,he containment ;2+h"'y c !:ul:ti:n, 5^ perc c.:. of th
- ud
- n. i ci co.J :o;.-e
- < F th: ;;nt i.c.;nt i. gl.:.ed ::t ' net -+ e t:1y-
"huyw g
plI,per.g A,n,.i=r., cent:in....ni-leak rate WC--percent r
-u y4 containment,1=i49e 0,2.
"r
- u..,. a,.
( M T-trU p C;jg m W[^d. -dine concentration in the secondary coolant /.<:-: ::: =:d to L= G.1 g ^ :/wir-
- .(
Ihe A
X/ va s:
l 0-ra 1097 m i ^ -
<#=J
/m3 ( 4
/, mD.
I*#
0-8 hr t3 m
2.
IJc8 sec/m Lp )
09/18/84 16 V0GTLE SER SEC 15 INPUT
- - - - --=
3y J
Table 15.5a Assumptions used for estimating the radiological consequences following a postulated fuel handling accident in fuel-handling area Parameter and unit of measure Quantity Pos er level, MWt 3,565 Number of fuel rods damaged 264 Total number of fuel rods in core 50,952 Radial peaking factor of damaged rod 1.65 Shutdown time, hour 100 Inventory released from damaged rods (todines and noble gases), %
10 Pool decontamination factors Iodine 100 Noble gases 1
Iodine *- " '
hrms in s%spk ahosc f**I,'1
- '-^^^' "--^- ^^^'
Elemental ".
f
~~ ~ ~ ~ 4 75 p Organic %
25 j
Iodine removal efficiencies for ABGTS (spent fuel pool area), %
Elemental no filters Organic assumed X/Q va s,
c/m3 s
0-2 hr 10 m
xc usion area boun
.8 04 0-8 hr at,21 (low ation zone) 3.
x1 s 09/17/84 17 V0GTLE SER SEC 15 INPUT
.a..
... u.
,[
,?.
3.
r Table 15.5b Assumptions used for estimating the radiological consequences following a postulated fuel handling accident inside containment
- Parameter and unit of measure Quantity Power level MWt 3,565 Number of fuel rods dameged 264 Total number of fuel rods in core 50,952 Radial peaking factor of damaged rod 1.65 Shutdown time, hour 100 Inventory released from damaged rods (todines and noble gases), %
10 Reactor water cover decontamination factors Iodine 100 Noble gases 1
Iodine fractions released from reactor water cover, %
Elemental 75 Organic 25 Iodine removal efficiencies for containment effluent, %
Elemental no filters Organic assumed 3
X val
, se abok) 0-hr a 097 m xclusion
)
8x 0-8 at 3, 8m( w populatio one 3.
x 10-025% containment free volume mixing, 15,000 CFM flow rate for 15 sec. assumed.
9 I
09/18/84 18 V0GTLE SER SEC 15 INPUT
.w..
~...
.=
"n.,
.. -. :: -.. w v:_:
'~ ~ - ~ - - * -
'~
..,,n..
%m+
Table 15.6 Assumptions used in accidents involving small line breaks outside the containment Parameter and unit of measure Quantity Coolant released, lb 23,500 Fraction of coolant released flashed to steam, %
39 Coolant contaminant concentration, pCi/gm
- 1. 0 Spiking factor (iodine release rate multiplier) 500 O
9 o
O
- e 09/18/84 19 V0GTLE SER SEC 15 INPUT
~
._.. -.. = :.:
.:T.
2::#: ~~KTW.?:^^ ~ ^ _.: L.-
.~7L.' 2 ENCLOSURE 2 SALP INPUT FOR V0GTLE DRAFT SER Plant: V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 and 2 A.
Functional Areas:
Licensing Activities 1.
Management involvement in assuring quality.
N/A 2.
Approach to resolution of technical issues from a safety standpoint.
Rating: Category 2 3.
Responsive to NRC initiatives.
See 2 above Rating:
Category 2 4.
Staffing (including Management)
N/A 5.
Reporting and analysis of reportable events.
N/A 6.
Training and qualification effectiveness.
N/A 7.
Overall Rating for Licensing Activity Functional Area:
2 03'37/64 9HL c'!00Il acn sto Ao inrui s
i
. - +
.......e._
_ _. - - - _