ML20141L928
Text
_ _ ... _ _ _ _ _ _ - _ . . _ _ .._. - .. _ _ . _ . _ . . _ - _ _ _ _ _ - _ _ . .
/ \I l .=. I _ .
11 August 16, 1974 /
[
- t.y . (
If i J. Riesland, Project Manager, L C. E. Lear, tranch Chief, L l
OYSTER CREEK REACTOR FRESSURE VESSEL IH-CORE FLUX WONITORING TURE LEAK ,,, , u I Recently leakage was detected in the vicinity of an in-core flux esattering -
! tube on the bottom head of the oyster Creek reactor pressure vessel. Members I of the ACRS would like to have the staff discus 6 the invest 15ation and evaluation of this leak, its safety significance, and the staff's evaluatten of the proposed remedial action.
Please advise the ACRS office of when you will be prepared to discuss this matter with the Committee.
Originni Signed by J. C. lleKinley J. C. McKinley Senior Etaff Assistant cci A. Ciambusso (L)
P. Morris (RO)
B. Crier (RO)
S. Varga (L) becs ACRS Hembers R. F. Fraley ) !, ).
l(
Filed: Oyster Creek l
1 9
omet > ..geg g ... ... . _ .... . . , , , , , , , , , , , , , , , _ , , , , , , , , , , , , , , , , , _ , , , , , , , , , , _ , , , , , , _ _ , , _ _ ,
sumwt> ... jct +:hjw - - - . . . . . . . . . . . . . . . . . . . , . . . . . . , , , , , , , ,
om * . S/16/B . . . . . . . . . . - - --- -- - - - - --
r- uc.u. in . ..n, uca .m ..,,,_,,_,,,,,.,,,,,,,
9203270193 910807 / f PDR FOIA Y My/Oh, j DEKOK91-282 PDR- ~~
- - , - _ __ . _. ___ _ _ _ _ __ __ ___ f/_/
i 1k gammmmes ' 60W 1
KVPPLtHENT NO.179 CATEGORY 8 Oyster Creek August 16, 1974 MS _ Reactor Pressure Vessel fustrtanent Thitable Leak
'4 Memebers Jersey Central Power & Light Company letter dated June 17, 1974 reports the results of the investigation of the small (1 ml/ min) leak that occurred at in core flux monitor location 28 05.
The leak appears to be in the tube to vesset veld, The operator proposes to correct the leakage by rolling the in-core flux tube into the lower head (the technique used to install stesse generator tubes). '
l ACRS review is reconsnended.
The enclosed gaaterial should be retained until the recocanended ACRS review has been completed. 1 i
l l
i 4
I s*
%y
.., 's ,
'*h. s '
'o ALK5 orra > _ . . . . . . . . . . . . . . . . . . _ . . . . .. . . . . . . . . . . . . . . . . . . .
JCH:bjw SURMAME> . 6./.16/.74.. . . . . . . . . .. ... . . . .. . . . . . . . .
DATE > . . . . . . . . . . . . _ . . . . ._. ;.;g. . . .. - - - -
Fores A8C-)l0 (Rev. 9-53) ABCM 0240 *** .4 He-el446-8 +eH's
.., . n -.
_.,.,y ,yv- -
~
l ) I22A
. V.
%p! .
t CAT 100RY 3 1f%,a,$1rFFLEMENTNO.169
september 19, 1973 kggter Creek Was 14ee nf Warrel Power.
b Act8 Directorate of Regulatory Operatione Notification of an insident or Occurrence No. 95 dated September 11. 1973 reports the temporary loss of normal onsite power during }
transfer between transformers. On the loss of pou r both {
diesel generators started and assumed their loads properly.
Normel power was unavailable for about ten seconds. An attempt to reestablish normal electrical power resulted in tripping one of the diesel generators. Normal offsite power was eventually reestablished, no adverse effects wre observed. A move detailed report will be sutaitted by tte licensee.
No ACRS action is recomended.
The snelosed material should be retained as part of the permanent record.
(JCH) .
I I
h t
\
1 k
l I i
\
m U
.4 n,
I l
ACRS omca > _ .. . ...,_, , . . , , , , , . . . . _ , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , ,, , , , , , , , , , , , , , , , , ,
JCM:b
. .. .. . .j w. . . .. . . . . . . . . . ,
8UMME > .. .... , , , , , , , , , , , , , , , , , , , , , , , , _ , , , , ,
i '#
Daft > . . . . 9 /19[ 2 3.. . - ~
. - o c.. . ~ . m n o . . ... ._,._,, - . J 1
g gC f S
14 March 19)5 -
4 4
AC33 Meshers OTSTER CRTIKt CATEGORY B REPORTS RECE1VED 4 TEBRUARY 1973 T1ROUGil 12 MARCH 1975 B02056 he NRC Staf f met with the Licansee and discussed the following:
- 1. Proposed radweste system modifiestion and proposed i
toch spoes for release of groes activity thearugh
}
the atack i 2. Proposed ter.h specs for water quality control, j including a specification on pil
- 3. Use of EXXON reload fuel assee.blies
- 4. Status of staf f review of Licensee's submittal en j high energy line breaks inside and outside of
. c onta ttrae nt . (nts is part of the FTOL review.)
l he Licensee confirmed that armed guards are being hood, that hydraulie anubbers will continue under surveillance (conversion to mechanical snubbers is contorpisted),
aM that a change in tsich specs relating to the use of
}
the rod worth mintuiser is being prepared.
During a tour of the facility, the Stef f noted that
- feedwater ard condensate pump room exhausts go to the l atmosphere without being monitored.
I B02093 The Licensee, reluctantly but in confors.snee with the (WRS,JA) Staf f's position, submitted a proposed change in Technical
}
Specifications to require the use of a rod worth stoimiser.
The requirement applies to operation below 10% power and i with rod insertion densities > 50%. The Appiteant maintains that the present use of two reactor operators and controla rod-pattern templates is edequate at those times. The Staff must find some credibility to that arguiment, since they appear willing to permit one startup per yaar without the 10lM.
t g sw o , , .c .
ACRS .
PTBurne t/mp
..v.
14 Mar 75 f ons ABC H 8 (1.=, p.$ H Ar,CM 0340 Q y,3. ,,yg ,,,,,,,,,p ,,,,,,,, ,,,,,,,,,,,,,,,,,,, *}
.m .
( (- -
s.- 2 D02496 he Liesasse reports rescot espertoase with sembber (WR8 JA, service life and reliability. Installatten of ethyleen @M r . ~
IS , fCPS, propylene (EP) sosts has significantly improved the (hi
- 90) perferusace of Bergen-Patterson hydraulie eaubbers. sp ;
- Movertheteas. the Liseasee believes that a better seletims '
to the saubber problem is the replacement of the hydraulie units with mechanteel saubbers.
802603 Staff aetificatten of poteattal for damage to torua (WRS, JA) structure during relief valve operetten with bet suppression Peel. Siellar letters seat to all itseaseos
! with MARK 1 contalements.
1 302833 he Staff rejected the Applicast's preposed enamented In-Service Inspection Plaa der postulated high energy
- pipe breaks inside of eestainment, and presented an I acceptable prosrac2. Staf f review of Liseasee's evolu-attom of breaks outside of eestatement.has met been completed.
802834 he NRC ' tasued Shange No. 23 to Technical spoe tfiestions permitting the lowering of MSIV low pressure sleeurs set potat.
Abnormal Oecurrerre Reports (All to DO) 801541 on 8 Jan 75 one of a redundant set of terua vacuum breaker (WRS, JA) valves failed to operate. he cause was related to the adjustment of mechanical linkage. Modification of the valve actuator-is being seasidered as well as total 1
replacement of vr.1ve and actuator. More is a history i- of leakage problems with the valves.
i B02494 Inadvertent release of 4,000 gallene of condensate (specific activity 387, of allowable) la se uneestrolled (WRS, JA) manner.
l .
302509 A sentainment epray_ pump fatted to start la automatic mode.
L , 802773 Fallure could met' be repeated. Redundant systems were operable.
l q Paul T. Burnett i
staff Assistant 1
t j ee ric e
- __ _ ___ ._._ _ _ ~ . _ _
i .......* _ _ _ .._ _ _ _ _ _
i fem ABC 518 (Rev. 9 55) AECM 0240 W u, s. sovenmusset eme=reme crescas teva. sos.soe
(QO
{ {
i ,
, .0..
May 27, 197$
h.4 s g d [f ACRS Membors SUMMARIES OF CATE00RY "B" EEPott1 FOR OYSTER GLEIK WUCLEAR GENERATING STATION i
Distribution of the subject reports is todicated by the initists under l
the document control number to the lef t of the sunsnary. Please mention
- that number when requesting copies.
I 503214 Staf f suonary of meeting with Lineesees. Subjects NRS , JHA , D0, S HB) d isc us s ed t Primary coolant water quality, pipe breaka
, outside containment, augmented of f-gast system, in.
service impection and pipe breaks inside containment.
30321$ Stoff susesry of meeting with Licensee. Subjects die.
(WRS, JA) cussedt setpoint drif t of pressure switches, valve motor failure and information required for F1VL review.
103224 Licensee's proposed revision to application for Y10L.
NRS,JHA) Addresses an augmented inservice in1pection program for high energy lines inside containment.
I 803238 Staf f request for water chemistry date. Also, requests
- l. NKS, JHA) information on ultrosonic tests, test procedures, and qualifiestion of NDE personnel. All reiste to SCC l
problem in BWR's.
503625 Staff suusrary of meeting with Licensee to discuss ECCS NRS, JHA) resnalysis. GE blowdown analysis is the pacing item for ECCE analysis of both GE end EKION fuel. Startup
{ af ter refueling may be delayed.
503668 Licensee's emenda. ant 75, Revision 2, to F8AR.
(All)
B03694 Licensee submitted ECCS model for use with reload fuel.
(WRS. JHA) 303780 Amendment 76 to FSAR and requested change in Tech Specal (All) limitina coMitions for operation.
l oeric e
- eununue n . . ~ -
l . , . . . - - . . - ..
i ..... . _ _ .__ _ _ . _ , m i Forto AIC lit (Re'. P 53) AICM 0240 W u. e. oovs nnutur enturene o**icsi sof a.es.-ios O c .
- .- - .. _ - . - .- - - . ._ - . - - . - - _ _ - - . ._ ~ - - ._ -
e .
( (
\
ACR5 Meubers 2 May 27, 1975 B03808 Supplecent 2 to Amendment 768 description of thermela (A11} hydraulic chereeterietice of fuel aesemblies: Propened Tech spec changes to Fuel Clodding Integrity Safety
- Limit Curve and LB58 on peaking factor.
303811 Licensee request for change f 36 to Technical Specifications NRS, J11A) to incorporate results of final ibCA analysis for 7 types of fus!.
B03BAB Staff letter to Licensee requesting program to monitor NR$ JilA, INN) ventilation exhaust not proseotly monitored sad which is +
3 a potential source of airborne conteMnation. ;
i
, 803656 Staf f nottee to Licensee of esandment to POL and Tech
. @RS, JilA) Specs to permit operation with 8x8 assemblies, reduce ina l sequence rod reactivity worth and to require a greater degree of operability of rod worth minimiter.
l B03*n) Licensee submittend proposed revision 1 to Tech Spec MS, J11A, SIIB) change f 28. Revises requirements for examination of pipe v-id and branch line connections.
B04102 Licensee suusnary of safety and technical review of LOCA '
NRS, JHA) analysis in support of changes 36 and 37 to Technical Specif ica tions .
p04724 Stef rejects licensee's proposed changedto Tech Specs.
. FRS, JllA) Change would not require a 11 censed senior operator to j directly supervise fuel handling.
, B04369 Staf f sunenary of venting with Licensee. Subjects dis- i
( NRS , JllA , iMM) cussed relative to Tech Specs: qualified radiation i protection personnel, use of senior operator for refueling operations and sections on " Review and Audit" and " Pro-cedures."
E04598 Staf f requested Licensee submit proposed change to Tech
- NRS,JilA) Specs t specify heat up and cool down rates as degrees en one hour rather than degrees per hour.
B04820 Staf f informed Licenses to increase the number of break
@RS, JilA) aises analysed in ECCS analyses. Hay delay startup following refueling operations then in progress, i
i B04883 Staff requested Licensee provide additional information on NRS, J11A) 26 iten.a on Cycle 5 reload application and on 19 items '
reisted to ECCS performance.
i .m"*
j _ _. _. _ _ . . . . _
j . ._ . _ . . . . _ _
i en> - . - . . . - . . - . - - -
l Pena A!C48 5 (RM. 9.$H MCM 0240 W W. e. 00Vammusuf Pagtene creetse sen.os& tee
., -.. - _ . , - . . _ _ , _ _ . ~ , ._ ,-. ~ - _ - _ . _ . . _ . _ ~ - _ . _ _ _ _ _ _ . _ . . _ _
e .
( ( .
ACRS Maubers May 27. 1975 Revision 3 to AmeMment 75 to FSAR.
B04902 (ALL) ' - 'l a
ATW8 analyses for Oyster Creek. . '
3051LQ (All)
B051D Licensee submitted additional information on thCA (VRS,JilA) analyses of EIION fuel. Clod temperature aM h versus time data provided as a functionoof break staa.
505253 Licensee submits proposed Revision 1 to Tech Specs (WRS, JIM) change D0 to provide for direct supervision of fuel
, hem 11og operations by senior operator.
i
! B05244 Licensee responded to Staf f (see 503888). Subject l
(VRS, JilA, IAH) exhaust duct has been monitored tetermittently. Test results aM analyses do not categorise that area as a "primeipal gaseous af fluent discharge path" as defined in Reg Guide 1.21.
B05253 Licensee submitted nonproprietary version of responses (VRS, JilA) to 72 Staf f questions on the Cycle 5 Ralped aM LOCA.
B05418 Licensee submittal of non proprietary information on (WRS, JIM) LOCA re-evaluation. Includes anevere to questions 64 6 65 by Staff.
B05453 Licensee responded to Staf f fsee 504598) by defining (VRS, JIM) the thermal transient as 100 F in any one hour, which j is the present Tech Spec definition, 505456 Licensse reported changes in radiation monitoring of (WRS , JIM, DbH) personnel. Principally, he has changed from using film
, badges to thermo-luminescent dosimeters, which can be j read on site.
i EO,1 Supplenent 8 to application for a FIDL.
l (All) 1 B05502 Supplement 3 to Amendment 76 to TSAR.
(All)
' I I
f i
j O'hC'* _ _ . .-
- V*"*"'*
f ,._ _ _ . . _ . . _ . . . .
- A*** . _ . _ . . . - .. ..
j Param AIC lit (Stet. 9 53) AILM 024D Q u. a. novenesessert Paswvine otricas sete. ass. tee
~
l.
(
i ACKS Maubers 4 May 27, 1975 i
Oyster Creek Abnormal Occurrences (All to WE8 & 90)
.,.1 B03336 A0-75-4. The apparent failure of a relay to reset (M81,JHA) eaused one rod to insert when an attempt was made se' withdraw it. This action caused a rise in 11esar best '
generstion rates of surrounding fuel assemblies. Fuel damage limits were not escoeded, but the limits on average planar linear beat generatim were anceeded for
~ 14 minutes.
{ 803443 A0-75 5. A breaker trip bar failed to reset leading to
, the f ailure of a containnent spray pump to start in the automatic mode. (The system was being tested.)
B03711 A0-75 6. Stack gas monitor system was inoperable for less than one minute. Causad by circuit design arror.
Redundant punips tripoed by a single thermal overload switch.
B03967 A0 75-7. A dehymidifying hester in the gGTS fatted to (thM) energine reducing the absorption capability of the 4
charcoal filters.
B04133 A0-75-8. Failure to properly monitor the core led to exceeding the limit on a F1.HGR. Calculated = 11kv/ft, allowed = 10.57 kw/ft.
Bot.587 A0-75-9. Cable f ault led to loss of emergency power to one bus.
I
- B04476 AG-75 10. Excessive leakage through 2 torus vacuum breaker volves led to loss of redundancy in primary containment isolation.
t B04669 A0-75 11. Failure of packing on a valve led to ancessive leakage in the main Jine drain and bypass line.
l 305478 A0-75-12 Seppoint drift. Permissies switches for low l
pressure core spray valve tripped below pressure itenit.
B05521 A0-75-13. Failure of time delay relay required to
- ensure initiation of isolation condensor system.
4-
- Cha h p ~
l S cu ' p{An) pu M 4 )) fa %4. hv - '
r paul T. Burnett Staff Assistant j
- . 4ae4c e u r e e I , ._..._._ _ _ _ . . . _ . - - -.
! j A j .u-a.s .s tated I
$4tG UD - .~ --+e -.*-.- -- --
f reem Ascsis (are. p.ss) Atod c240 W u. s. eoveuusanamtme orrec ts ***a **u **
-- - _ - . - . - . - . _ . - . . - . _ - . , . - - , - . = . - . - . - . - - -
( .e, ( -
(qa
' s (30*'f f
,. , - . W Ag6 "g.
BlateofNrtu3Jrrary DEPARTMENT OF CNVIRONMENTAL PROTCCTION "I H E NToN 00025 0, r te t or THE couulpeloNrn October 18, 1976 Dr. Dade W. Moeller Chairman Advisory Council on Reactor Safeguards Kresge Center for Environmental llealth .
School of Public !!calth liarvard University Boston, Massachusetts
Dear Dr. floeller:
As part of our effort to evaluate the need for land use restrictions around nuclear power plant sites in New Jersey (see the enclosed March 10th statement by Commissioner David J. Bardin), we have com-missioned a special analysis of the oyster Creek nuclear power plant.
This analysis, done by Peter R. Davis, basically applies the methodology ernbodied in the U. S. Nuclear Regulatory Commission's Reactor Safety Study to the specific design.and operating experience of the Oyster Creek facility. The tentative interim conclusion in Mr. Davis' preliminary report is that the probability'of a catastrophic accident over the remaining anticipated lifetime of the Oyster Creek facility is approximately one in one hundred. The main determinant of this result, which we recognize is a greater probability than one might expect, is the probability of the failure of the reactor to shut down after an anticipated transient.
We are circulating this preliminary draft for technical rev cw to a small number of individuals and agencies, of which you are one.
We would appreciate receiving any commenta you may have by November 12th.
To speed up this process, I would appreciate it if you would send a copy of your comments to Mr. Davis at 1935 Sabin Drive, Idaho Palls, Idaho 83401.
Sincerely,
=}
flh (ss fh G16nn Paulson, Ph.D.
Assistant Commissioner for Science GP:lew
Attachment:
An Investigation of Ilypothetical Catast rophic Accidents in the Oyster Creek Nuclear Power Plant - Preliminary Draft, September, 1976 J f
, ec: Mr. Peter R. Davis [
f~ .% (
.li.r,,m..,*..a..oeh
{
([,k.i ; \ d W p
M c. , fl[W JERSEY DTPARTMENT Of
- ff**
, N' EfWIRONMENTAL PROTECTION y , --./l M P.O. DOX 13901RT NION. N.J.00025 David J. Hardin, Commissioner '
l' / 009 ?97 2994 1 h Wu Denman, Pubbe inf umation Othcer (51ATCWlDi) gY
-~
tio. 76/231 Ir~nediate release:
BARDlH CALLS FOR LAf40 U$[ CONTROLS March 10,1976 114 VICINITY Of NUCLEAR P0i.'CD PL At4TS TOMS klVER--Cnvironmental Protection Connissioner David J. Bardin today told a Nest Jersey Senate corir.ittee that "the time has come for state and local government to initi.ite practical land-use controls around nuclear power plants."
Testifying here before the Corinittee on Energy and the Environnent, chaired by Senator John T. Russo, Bardin said nuclear power plants will reduce New Jersey's reliance on fouil-fueled plants and help mininire the air pollution in the state from sulfur-containing oil and coal.
"This tradcof t is accompanied by other new risks related to rediation associated with nuclear pp.nr," Bardin said. But he added that in contrast to rany other recent technologhal developments, " nuclear power is probably the most carefully scrutinized regulated tx hnnlogy in the United States today "
According to, hardin, "the bechive off ers a rNgh analogy: It's useful and pretty safe. LWt Lees can sting. A stinging swarm can hill.
50 the beekeeper takes trecautions and the rest of us keep our distances.
We don't put the kids' sandbox or the swimming pool up against the hive.
That's common sense."
Bardin noted that the federal Nuclear Regulatory Commission (NRC) has a policy of limiting the human population near nuclear facilities, but said that while the policy has been in effect since the earliest days of the power program it has yet to he implemented effectively af ter a nuclear power plant has been licensed by the federal agency.
(oore) e
. ( ( ,
e Nuclear power plants Add 1
.' "We can and should mnimize the humn exposure to the risk of hmn from a major nuclear accident," !!ardin v,id, "by enforcing population density goals and other ressures." He called 4ir conpatible lend use near present and future nuclear plants. Certain industrial operations, for instance, might prove compatible, Bardin noted.
The Department of Invirontertal Protection is conducting an in-depth study of potential ha u rds and land use reautrements, Bardini said.
The study headed by Dr. Glen, Paulson, assistant comissioner for science, will onlist the help of municioni and county officials and lay the groundwork for a full review later b..' all interested parties, Bardin added.
Pending co9pletinn of this repyt, said Bardi;., LLF will not approve the construction of residential duelopment or uther heavily occupied facilities in the proximity of eitter toe Oyster Creek nuclear plant in Ocean County or thoc,e at Artificial Island in 5slem County. The action is talen under the state's (castal Ares locilities Kevuw Act (CAFRA).
Derdin sa id OU also is stu,1ying '"e'hRC's '>reposal on the clustering of nuclear plants its a me: ins c f reN ing lar.d un imua ts. But he said DEP has advised the f,'RC that Occan County seems a not.r plate for suct clusters.
Fle said DEP would deterriine at a future date snett er th? concept has a place elsewhere in flew Jer" ey.
(Text of ter.ti mny attach.'d.\
l O
WW
, . , le.it Imony .'
utvid 1 h mfin. e..1 41.ne:,
New It'l .i lMp s- af .! (LV i r6 Hif .esil ..) 8't *t 8 ( t i oin 1 ~c
%i 4 t. ,, - . .j .m . : i one "
' h . t.
1 s .M G.v.i. '.a
. h?wy M.1rch 10, 19 7'e Senator Po sso, r er,hers of t he Cormnit tt e
- Nucicar power ir. ne rt rene.ac t o New .les sey . The Oyster Creek e.tation, located abeut 30 irfim south u: Tens River, ha produced elect ricit y F ince December of 1969, wit h a pet ,t ial mniw. sutput of 620 negawatts. The first of two unitn of the Salen 4t.9Joc.. locat"d e- .', r i . f i c l a l !s)and in Salem County, will be t,t a r t ed up lat er t hi! . t. 1his unit c pro hire 1100 mep,awat t s of electricity. Its identical twia, scheeulea to rnrte on line later this decade, will have t he sare ; et ontill aut pot . Uther triclear plantr, both in New Jersey and in adjacent states, are at va r i um r t a r,i a. in local, blate,and federal regu-latory proceedin p. .iite kw Jer**v ih pa r treta t of F.oviron"mtal Protection (DEP) f ollows the licensinr. ,$nd topslaun y tiroc.mes for facilitics in nearby stater, in additinn to en.rcisinf. our niet ntory respont,1bilit ies over nuclear power plants within the berders of t hi 'tatt of h.e.lerney.
The operation of both exist in; and f uturc nuclear power plants will reduce New Jersey's reliance on fossil-toeled power plants and thereby help minimite the air pollution in the c t at e f t a. nul f ur- centain tne, oil und coal. This tradeoff is acconpanied hv other nev risks relatcJ to the radiatten associated with nuclear power. In contrast t o n iny ot he: recent t ecio.ological developments ranging f r om acrosol cans ard pestI<1 der to thc nupernenie transport. naclear power was sub-jected to extensive satety anu bn. ironneatal evaluat ion before its commercial introduction, and it. .' r n*,,b l y t h e 'ust c.dref.11v scrutinized regulated technology in the Unit ed States t o. n.
The bechive offerr .i reugh cai!ef.y: It useful, and pretty safe. 1)u t beer can sting. A s t ingin,. rwa rta eaa kill. So the beckeeper takes precautions, and the rest of us keep our dim ance . %'r don't p"s the kids' sandbox or the swim-ming pool up against t he i is e. 'ih a t ' s c u.ta. . n s e n s e . That 's been our national objective reg.u ding land eae near no.: ar po.er plants, but this country has not f ully inpleme.nt ed t his corren eenv e acal.
The S t at e r,overm-co t imt wke t h" s ea l i st i c .in,I respons ible protectien of the public health and s.;i;s krset w e lo all os it s act iv i t ie>.' rer,a rd ing nuclear power.
Review of Available 1.i t erat o re j There are many sources for scientific, medical and engineering information on nuclear power generally as well ar. on specific facilill> Some significant l
ones are: ~ .
l l
l
.. . ( .
(
\ .
posed f acilit y, t he a.ldit lenal .som ane e xpo'u o r to the ent i t e populat ion wit hin "3 0 ti le s is 0.003 (ibrec rino- t hou . andi hs) rat a / n : Ihis, too, should he com-
. pared t o t he l'10 mren/y t r om h e l nt o.ent r adi a' ion. Vittoally all of t h i s cornes from the used ("Irradfatc1"I 'oc! 'od, whiih 1: m nt ont ol New .lessev Ior storage and eventual se l i i e. i n i si .o! !i t ii i a : . e r .' ri .o ir ce u. l'iot ope s for use in ftedical tre1 tent anJ r m atch, .'nd f oi n h"i purposes.
Ultittate Vaste Disp y l
- 1 bete are no f acilit ies now propo .ed f or const ruct ion fu New Jer+.cy to either process used loci roJ, or to store 01 dispoue on the radioact Ive wastes altet such pr oce sing Tht closest pat ent ial sites known at t h i s t i rne a r e in opstate Ne9 York and in Barnwell, ' ruth Carelina. We agree with the Congres-nienal Of f t re of lerhvilegy Annes .~ent , which c.tattJ last October, "Satinfactoi, handling of nuclear fission wast er, appeats. to be technologically feasible, though it han yet to be demnuttated". This area of weit has not been funded adequately at the fedeial le'e1. We <opport m.ditional funds for retcarch and especially denonstratico project: in this at a.
DEp doe. ent belies < t hat N~e .itr~es tr il: :t - w!1, receive ans aJditional exposure fron this part of the tnel cyrle, c :s c e p t for tne tianspottation of the used fuel redb alteady discu ued.
Major oij a t a s timable fu . ! . ' - n t .
During the operati<' e' "*nnd<at pow t plant, nrw radioact Ive mat erial is produced by ficsion wd hs ot ot ron at tiva: Ica reactions irom the metals and other elenentr. in th( ieactoi m, s t ei n Thete 1 no question that the inventory of radioactive f i s r. i e: proJarts rhl.h would .a., %1 ate 1:. t he reactor's nuclear '
core during egeration woul', it relea.cJ In aigallicant yn nit it y due t o sorae catar. trophic accident or Jeliberste a. t , pm. aa a v e riske to life and health for several tilles dovmtind o f the acc uent. 1 01 this r ea non, SitC sc r ut iny o f any proposed reactor de.jp p!att -
heavy cepha is on a ieview of baf et y-relat ed featurs8 of the reactot te and a w riated hardvale (i nc l uding o'ne r genc y cooling cyttetM , t he siiu.tatal iot egri t v of the building in which the tcactor is to be houseJ and f act ors c..t e i a il to the la,111ty liself related to its .
proposed location (e.g., population .h n <, i t y ) .
There aie a serles of 13, pes of ace iderit > I h it can be envisioned, ranging wjdelv in degree of severitv. The s t and ird Al i: /N!W r antine, tanges over Classen 1-9. A Class 8 accident, the de I o h iis for react or saf et y analvsj s, is t erraed liy t he N14C t he "nax icunn c t edi ble acc ident "t WC appaient ly considers Class 9 at eident s incredihte. A t.1 :e. . S accident is one ' hat does not involve a tua, lor ca t aut t oph ic t allui e of bot h t hc reat t or core and t he hullding around it. II )t he r it involves a serles of f ailores' t hat inercase tbe normal, dav '
to-day seleanes substantiallv, but noi necewis ily t o ra p l il l y lethal levels.
D Cl" 9 continnous radiation suiveillance at nuclear plant sites provides an ext ra level of salety assurance to thd publIe for Clao. S and less seriou.,
accidents. Thin et f-sit e sorveillance nyoten, while not able to detect in a t imely mamwr t he onset o t' an ext remely unilleiv hut rapidly orenre ing cat a8t rophie Clan 9 aceident , nonetheless has the capabiiitv to detcet Ihe development of malfunctions wit hin the facilitv which, over the course of days or weeks, m hht reonit in significant teleases of sadialion into the local environment. Such releanes can be detected even at 'l e ve l s well below those
.- h .
. t currently allo.ced by ! W laufsiot a. Nell .. .cr.ls .p,o vathing t hei.e st a'ndard s. ,
'11 tis Ot.ite rmnitotitw e is,ett,o a mn 1 l.. tally toh rendent of any on-
- going monitorinr,and rt oi".i.. o i _iis . a ., . . ! in ,
aa. ot f lit y or by any federal agency.
A Claim 9 m r ident n . .. i t an Mtn.- :be seattor core and the building housing it. Th i s coul d r m.ul' .n tapid relo m of larce ar.ounts of radioact ivit y that would drift dounw h 1 'o.ient pahlle t eut r .o wy Iw usen on t his rist.
Thin cont rover r.y hat; been fneied, ii ! cast in latge patL, by the iallure of the NRC's predere aor, t he At ed e Ene r t :s Commj ',1;ik , t o carry out in a linely man-net long-planned sarc a studitx; to the finel n port of the Energy Policy Pt oject of the Tord roundat ion (197!.! not' 'an adequate experimental basis for nuclear saf et y sptem is IsrLira:". We rondear. the federal budgetary and policy deciniont that delayed thir, important safet y resear ch and testing work when the AFC was in charge of i t. . We looh t u the Energy Research and Development Adrinist ation to elle-Inate these attas of uncertainty.
Probabill tv of a RQor Am i lent The NHC 1:ecc tor Sa f et .- S tudy est imates (in the ab:,coce of experimental data) t he probabili ty er f *..*quency of let hal ace j dor.t r for nuclear reactors as well as f or other technnionical f ailures and f or natural disasters. Final professional reviews of these (.stimtes and tht uncertainties. in them are not completed, and may lead to sigrdficant changee in the W Study.
By way of perspective, the Study bcs a cenparison betecen the frequency of .n airplane crash killing, up to tuveral h. , icd people on the ground (for example, a large airplane c i 161 r:, into a residential area t.uch au near Newark airporO and the probability ca a neejear reacter acciJent killing the same number of people. In reoeral, t he serious renet er ce nideal, according to the NRC Study. is about. ene t hencind t ines less likely t han the air planc accident, which is in turn projecteJ to armur at t he rate of about one every f ew cen-turies. However, sone profe olonalt4 have challenged that t h i ti underestimates the uncertaintier, that r.... y le.nl t.o a maj or rc c t o r accident.
The r.ciences of stat ist ical r ink .inalysis .mJ radiation d%e estirnat. ion have evolved grajually and relatively recently. In th(tr prenent formative l
stages, they do not deal, i tequ ately with the probabilit y of a series of simultaneous f a tinres or with t he pot ential long-l art itu, (i f ei t s of t e8 ult ing exposures. Apparent!v ih, ser:v- of even t s t hat led y to the fire at TVA's lirowns ferry reartN , for eunple, was net incluacJ as a possible set 01 event s in t he NHC St udy.
l To t ry t o r educe scrce ef t hese uncer t aint ice , New Jersey petitioned
! the NRC in 1974 to conduct anal yses of the 11Lellhead and i onsequences of a Class 9 accident before 11ceanin>; any new t rac tor con f ir.urat. ions (NRC llacket PRM-50-10). Regret t ably t he NI:C has y et to act on our pet i t ion.
Consequences of a Major Acc ident If there is disagreement as to t he Iikelibood of nuch eventa, there in no disagreement that the consequences of a maior release of radiation would i be potentially ~
catast rophic. The Ni:C Study as well as previous studies show t hat a major radiation release passing over any human populat ion could kill l people either frem radiation slekness or cances and coulti rontaminate land l
and water t o such a degree as to render them unsafe for a subst ant ial pet iod j of t ime. '
i.- 4 3
k (
', 1his led the MC to long ano define the concepts of " exclusion rone",
" low population rene", and " population center distance" (10 CFR part 100 en. bodies t hese cr it er ia and def ini t ions) . The " low population rune" is an
'a re5 '*vbi c h con t a i ns re s iden t >. , t he t ot al numbes and densit ) of which are such that there it a reasonable probability that appr opt late prot ective mea-sures (ould be talen in their behall in t he event of a *erloos accident", The NRC's (at alog; oi "pi otet t ive ne.o.ut es" includen evaruat ion and f allout shelters.
The NRC has ref used t o act u ill e Jeline t his tone eit hen in terms of populat ion density or total population, tuntending that "the situ.ition tuy vary from case to we". In t he case of Dy*.t er Creel , the "c).clusion rone" is land around the station ovned by the utilit yl the " low population rone" in the State's view is a 3-mile radius a round t he plant ; and the "populat ion center dist ant e" in a radius 4 milen from the plant.
Erner gency plan s These sarne concerns rerarding the potential consequences of a triajor reactor accident have led to the development of emergenev plan requirernents and crit et la by the NRC and other federal agencies. These plans involve preparation of de-tailed response procedures by the utility itself and by state and local of ficials.
These offic;als include, at the state level, the Goveinor's office, DEp. the State Pollcc, the civil def ense agency, the llealt h Department and it.deul altnobt all other cabinet apocles. At the county ar.d local level, local officials as well as similat agencies (police and fire departments, civil defense agencien, etc.) are also involved.,
These plans, developed in view of the specific characteristics of each nuclear power plant, include.
- 1. nethods to asscu the severity of an accident
- 2. potential corrective actionn (fire-fichting, danage centtol)
- 3. methods of shelterinc, propie o: evacuating them irem the path of the radioactive cloud
- 4. techniqucs to keep people frorn entering the affected area during periodr of rist
- 5. methods to prevent exposure (protective clothing or breathing equipment)
(; .
e s t ab li chnen t of chainn of communication and responsibilit y araong t he va r i o us public and private agencieu involved.
- 7. procedures by which accurate i nf orreat ion is made available to the public during such an emergency l
- 8. criteria f or when to allow re-ent ry into affected arean.
To be effective, these planu require close cooperation between the many other agencies of state, county and local government and DEp's Bureau of Radiation protectlon. The basic component s of these plans are publicly avail-i able. The only confidential sections are t hose relat ed t o detail s of corwuni-cation links, state and local police enobilization procedures, and the like.
es e.
l
~
rw- - - - - , .
( (
1he plan is flexible, in t hat it allowe. t he nature of the overall response t o
' b t-t a 11 tired t o t hi extent of t he ener r, enc y Md ( ond i t i on't atAt he t ime I' lass 8 (e.g. ,
ext rene weat her and it s etiett on available I re nspo r t at 100) .
reactor accident it piirex a far note substantial respo%e t htuianearies, reh vant a Class and I, accident. The plan is perioditall. retico d 'tby h ci tae, e tests to date have stopped i t; tested at l oin t in patt once per ytar.
t.hort of t he at t ual evat uation of pecple, hot have intluded the nobilization and deployment 01 agency perne mel at all levele.
I 1he maintenance of thi> merp ncy respon e ccpabilit y is impar t ant.
Luvironmental quality to convene annually a have directed DLp Div.nion of meetinr. of all appropriate a:ancies in the vicinity of each New Jersey rmclear power plan t . The pur pose sill be t o review all r.pecific plans and proceduren, update t hem where nstessary, to nate certain each agency understands its respenhihilities, and to ensure that neans of im;ilenent at ion ar e in good order.
location of Nuclear power Plants: St at e and Local Henulation of Sutrounding I and l's e s "the pntential reator-accident threat to the puh11e htalth and safety pronptrd the NPC cbjtttives of liniting the populat "there ionisdensit y near nuclear a reasonable probcbility f acilit ies and cont rolling L.nu uses so thatt aken. . . in t he event of a serious accident" that ptotective measures could b:
(10 Crn part 10:0. The NRC has not f r pl er ent ed its objectives, apparently hoping to pass t his recronsibilit y on to state and loccl governments.
Local decisions could plr. a Ley land-use role. Moreover, bot h of the New Jersey nuclear power plant sites are in the coast al area as def ined in the Coartal Area lacilit. Revies. Act (CAPPa), alt hough t he Oybt cwestern r Creekboundaty plant is only 1 1/ 4 n i l e *. i ror the Carden Stat e Parkway, the prose nt of the CAIRA coastal area in this region. A particularly denanding responsi-bility rests on DLP to evaluate land uses within the coastal area near nuclear power plants, including appteval or dimppr oval of all residential developments of 25 unitc or moie.
1 h e t irne ha s t ene for state and local governt.ent to initiate practical land-use centrols around nuclear power plants.
Exinting state law provides some of t he regulat ory, tools for realistic and responsible f oolement at ion of a major nat ional polley: t he ALC/NRC policy of limit'this ing the hu un peculat ion policy has been in within reasonnble diniances of nucleat facilities.
effect since t he earlient davs of the national civillan nucletr power program, but it has yet to 5e implement ed ef fect ively af ter a nuclear power plant h.w been licensed by the AEC or NHL. We can and should ntnimize the human expo-sure to t h,* risk of harm f rom a major nuclear accident by actually enforcing low population density goals and other neasures.
The approiriate land ure controls need not necessarily foreciose develop-For exarrple. we recognire the dist inct pos lbilit y t hat certain indu -
t rial act ivit les could be most compatihle wiih nuclear power plants. Non-labor ment.
intensive industry is one possibility. Industries using hy-pr oduct heat are another.
As a f l r r.1 blep to this end, I have directed the Annistant Cornml 8 s t oner for Scienci, Dr. Glenn Paulsen, to analyze the following issues as t hey re-late to the existing oynter Creek plant atul t he near ly-l inished Salern plant
-h-1
-____ . ~ ___ Z _
l (
- * - - population density 4
- meteorological characteristics
- engineering characteristics of the facilities
- the probability of and the potential radiation releases from various types of accidents
- the potential human consequencen of various types of accidents under characteristic weather conditions
- he status of the emergency plans for these facilities
- app l cable federal, state and local laws and regulations
- compatible land uses Dr. l'aulson will be aided by other appropriate DEP agencies, including the Division of Marine Services, directed by Donald T. Graham, the Bureau of Radiation Protection, headed by John J. Russo, cad the Office of Coastal Zone Management, headed by David N. Kinse; . They will seek the help of municipal and sunty officials. Dr. Paulson's report will provido the basis for a full review by all interested parties of '.hese issues and proposed lan; 'se requirements.
Pending the submission of Dr. Paulson's report and appropriate followup procedures, DEP's Division of Marine Services, directed by Donald T. Graham, will not approve the construction of resident ial developments or other heavily occupied facili'ics in the State constal arca in the proximity of either the Oyster Creek nuclear facility or those at Artificial Island.
Nuclear Enercy Centers and Mini-clunt ers The clustering of nuclear power plants has been proposed, amonE other c reasons, as a way to reduce the overall land use inpact of nuclear power plants.
Fur example, clustering veuld reduce the square miles of low population zones as compared with dispyrsed single nuclcar power plants. 5 mall clusters- per-haps half a dozen or sr plants, appear more practical to consider than large aggregates, in any event, Ocean County seems a poor place for either a mini-cluster or a large aggregate, and DEP so advised the NRC. The Pine Barrens wilderness, the highly developed harrier heach, the intermediate back bays and the adjacent wetlands do not provide the proper bet ting f or a nuelcar cluster. We :are evaluating the NRC study of the clust ering concept, which was required by the Energy Reorganization Act of 1974, to det ermine whet her the concept has a place elsewhere in New Jersey.
e+
(
.. ( .
Prel)minary Draft All INVESTICATIOff OF Tile PROBABILITY OF llYPOTilCTICA1.
CATA51HOPillC ACCIDL!JTS IN Tile OYSTER CREEK NLICLEAR POWER PLANT Prepared for the New Jersey Department of Environmental Protection By P. R. Davis October, 1976 g.a 9
9-9
Prelimin.y y Draft TAB 12 OP CONTENTS Page I. INTRODUCTION 1
-II. POWER REACTORS AND SATETY l III. RESULTS 4 IV; ANALYSIS 4
- d. ,
'* i' A. Determination of the type of accidents with the potential for catat. trophic off-site consequences 4
,ffy D. Determination of the significant accident sequences in WASH-1400 leading to core nelt and loss of above-f " A(
ground containment integrity 5 C. Determination of site and plant factors specific for Oyster Creek with the potential for changing the BWR core melt accident probability as assessed in UASH-1400 6 D. Calculation of core nelt probability for Oyster Creek 11 V. PIFERENCES 18 APPENDIX A: An Evaluation of the Relative Significance of BWR Core Melt Sequences as Addressed in WASH-1400 20 APPENDIX B: Sensitivity Studies . 23 APPENDIX C: Assessraent of Oyster Creek Scram Failure Probability 25
( ( .
Preliminayy I aft LIST OF TABLES Tch l e !!n . Pag I Frequency of Anticipated Transient-Induced Scrams for Gyster Creek 6 .
11 Differences between Peach Bottom II and Oyster Creek 10 III Assessment of Probabilities Associated with
- Anticipated Transient without Core Power Shutdown 13 IV Assessed Tellure Probability for Oyster Creek Systems 15' V Core Melt Probability for the Oyster Creek Reactor 16 VI' Core Melt Sensitivity Study 16 VII Survey of Core Power Shutdown Failure Probability 17 A-1 Dominant Hypothetical Core Malt Accident Sequences for Each Assumed Initiating Event 21
-A-2 Definitions of Accident Sequence Symbols 22 A- 3 Summation of Accident Probabilities 22
(
i-c
. ..?
we 9
l
( '( .
. Preifminary Draft LIST OF TIGURES Figure No.
1 Computation of Probability of Core Melt in the Oyster Creek Plant from an Anticipated -Transient with railure to Reduce Core Power 2 Computation of Probability-of Tailure to Remove-Decay Heat within 1 3/4 llours 3 Computation of Probability of railure to Remove Decay 1! eat after 1 3/4 Hours B-1 Influence of Anticipat.' Transient Frequency B- 2 Influence of Scram railure Probability B-3 Influence of Recirculation Pump Trip Failure Probability B Influence of Manual Poison Injection Failure Probability C-1 Simplified Fault Tree of Scram Failure Probability for Oyster Creek -
4 a
g 9
4 6
4
-dh 6 9
9
. {
f .
Preliminary Draft An Investigation of the Probability of Hypothetical ,
Catastrophic Accidents in the Oyster Creek Nuclear Power plant By P. R. Davis September, 1976 >
I.- INTRODUCTION This report describes the results of an effort to determine the probability o an' hypothetical catastrophic accident at theOysterCreeknuclearpowerplant{y) .
This study was undertaken for, and under the sponsorship of, the New Jersey Department of Environmental Protection and was directed by Dr. Glenn Paulson,
-Assistant Commissioner for Science. This effort is part of a larger effort to determine-the necessity for,'and extent of, appropriate land use requirements in
'the proximity of nucicar power plants located--in the State of New Jersey.
- The approach taken in the effort was as follows. First, a determination was made of the type of accidents which could occur in the Oyster Creek facility that had the potential for catastrophic consequences. (" Catastrophic" in this context means those accidents which night be expected to produco human fatal-
- iLies from. radiation-damage beyond the boundary of the Oyster Creek plant within 30-60 days from such an accident.) Second, a deter'mination was made of the specific applicability of the Reactor Safety Study (hereinafter referred to a WASH-1400)recentlycompletedbytheU.S.NuclearRegulatoryCommission(NRC)gi';
WASH-1400 cssessed the overall probability of catastrophic accidents for the two general types of nuclear power reactors currently operating in the U.S. Third, using appropriate adjustments to WASH-1400, a numerical probability, with the range of- uncertofnties defined as vell as possibic,.-was established for the catastrophic acci d ent potential in the Oyster Creek reactor.
II. NUCLEAR POWER REACTORS AND SAFETY For background and' perspective, this section represents a very brief descrip-tion of nuclear power reactors and their safety features. A nuclear power plant consists of a nuclear reactor and those associated components (such as generators, turbines, and electrical generators) necessary for the production of electricity.
A nuclear power plant is similar to-n conventional power plant in most respects except for the means-employed to produce heat. In a conventional power plant. a fossil-fut ! (such as oil or-coal)-is burned to produce hent for steam production; this steam is then used to rotate turbinegenerator units which produce electricitf.
In a nuclear power plant, a nuclear reactor in used for the heat generation; the rest of the plant is, in essence, identical to a fossil fueled plant.
w
Prelininary Draft .
For* fuel *, the teattor uses a ceapound of uranium which contains a small percentage of a special uranius atom (or isotope) known as uranium-235 or U-235. The U-235 atoms are " fissionable"; that is, they can be split apart when struck by a small atomic particle, the neutron. When a U-235 atom splits apart, or fissions, after being hit by a neutron, energy, in the form of heat, is released. The process of fissioning also produces " fission products", which are other radioactive elements and additional neutrons. The neutrons produced by the fissioning of one U-235 atom can be made to cause further fissioning of additional U-235 atoms and, if the U-235 atoms are properly spaced in a ratrix of appropriate materials, a controlled, continuous " chain reaction" occurs which is accompanied by the release of large amounts of heat.
The part of the reactor which contains the uranium fuel is called the reactor core. The core is located inside a large vessel (the reactor vessel) which in connected by large pipes to either a steam generator or a turbine. Under normal conditions, pumps continuously circulate water through the reactor vessel when a chain reaction is occuring; this water removes the hcat produced in the core by the .
fission process.
There are two types of power reactors in general use in the United States today.
'The esst cc =on is the pressurized water reactor (PWR). In this type, the water in the reactor vessel is, kept under high pressure and is circulated to a steam generator. The stern' generator is a large vessel designed to both remove heat from the water which circulates through the core and to transfer this heat to a lower pressure water reservoir which then boils to produce steam. The steam leaves the secan generator via pipes and is directed into a turbine-electrical generator unit which turns to preduce electricity.
The second type of power reactor is the boiling water reactor (BUR) . In this type, water circulating through the core is held at a lower pressure than in a PWR (though still higher than atcospheric) and ellowed to boil. The steam produced frca tbc boiling is piped directly to the turbine-generator units, causing then to rotate and produce electricity. The Oyster Creek reactor, the
- subject of this study, is of the BWR type.
The fission products produced in the fissioning process are highly radioactive ators of various chemical' ele =ents. Their intrinsic radioactivity causes additional heat to be generated in the core above and beyond the heat asso-ciated with the actual fission of the uranium atoms; this additional heat is called " decay heat". Thus, even af ter the fissioning of uranium atoms in the core has stopped (e.g., by causing the neutrons to become unavailable for further
. fissioning), a substantial amount of heat continues to be generated f rom the decay heat process. 'There is no way to stop this decay heat from being generated. Even
.-though it represents only a f raction of the full-power heat (perhaps 6-10% of the heat generated by the fissioning or chain reaction process when t' he core is pro-ducing maximum power), continuous cooling must be supplied to the core af ter the
. fissioning process is stopped to prevent overheating which coul,d Icad to a catastrophic accident. This is because, if the decay heat is not removed from the core, overheating from the decay heat can lead to fuel melting. If the fuel melts, the radioactive fission products are likely to be released from the core and could
. ( (.
4
.* .* Preliminary Draft find their way into the outside world. Because they are highly radioactive, the fission products represent a serious hazard to human health if the Icvel of radio-activity is high enough. Needless to say, these materials can threaten other living things as well as man.
In order to minicize the chances that radioactivity cauld be released from a melted core, several systems are designed for and built into reactors to maintain core cooling even in the event of unlikely accidents which could interrupt norchi cooling water flows; these systems are generally termed " emergency core cooling systems". Also, a large, airtight containment building is always crected atound a reactor core so that, even in the unlikely event of a radioactive release from the core and primary teactor system, the material might be confined inside the con-tainment building rather than escaping to the outside world.
There are two general types of accidents which have the remote potential for causing a release of radioactivity from the core of a reactor. These two types are: (a) an accident initiated by a loss of (normal) coolant, and (b) accidents initisted as a result of brief (or " transient") conditions in the reactor.
In a loss of coolant accident (LOCA), a cooling water pipe that connects with the reactor vessel is postulated to rupture. Because the water contained within the pipe is under pressure (even in a DUR), it is expelled rapidly from the pipe, and ev entually , if nothing else hcppened, the reactor vessel would become empty; this vould stop the chain reaction. However, if emergency core cooling systems were not provided to refill the reactor vessel with water, the core would eventually melt just f rom the decay heat , releasing radioactivity through the pipe rupture into the c ont ai nment . If the containment *s integrity (leak tightness) were, for some reason, also lost during the accident, the radioactivity could be released to the atnosphere. Depending on weather conditions existing at the t ime of the accident, the location of people near the site, and the ability of them to either leave the area or take shelter, the radioactivity could be carried by the wind to persons living the site in quantities sufficient to cause their death. Under very unlikely conditions, many persons could be killed from such an accident (UASH-1400, Main Report, ref. 2).
The second type of accident is initiated by a so-called " anticipated transjent",
an event which briefly but drastically disrupts the orderly transfer of heat energy from the reactor core. This can occur, for example, due to an abrupt increase in the amount of heat produced in the core, or by an abrupt decrease f
in the ability of the circulating water to remove the heat produced in the core.
If the normal heat transfer balcnce is not restored quickly by the various systems designed to do so, or if the reactor is not quickly shut down'g"th',F E6Ic' iould overheat and possibly melt. The overheatir.3 process could ca'une certain pressure relief valves to be opened WdiE"J, and, if a melting of the core had occurred, radioactivity could exit through these valves from the core into the containment building. Under some conditions, the containment building could fail to retain the radioactivity, and a release to the outside atmosphere would occur. This type of accident has been found to be more likley for a BWR than is a LOCA and is discussed in more de. tail below.
( (
Preliminnry Draft r
It should be emphasized that, to date, neither of the above accident sequences has yet occurred in a nuclear power reactor in tht United States. (At present, some $8 power reactors are ope rat ing in the U.S. ; the pet iod during which they have been in se rvice tot als some 250 reactor-years of operat ion) . In order for a radioactive release to the atmosphere to occur, several events must occur in sequence. These include (1) an initiating event, (2) the failure of more than one of the redundant systems designed to cope with the accident-initiat ing event, and (3) a f ailure of the containnent building. If any one of the three events does not occu{y)therevillbenosignificant radioactivity release. Calculations in UASil-1400 indicate that, for both general reactor types, the probability of a significant radioactive release is exceedingly remote; while useful, this con-clusion does not provide guidance on sppgific facilities. The probability of such a release from the Oyster Creek reactor specifically is the subject of this report.
111. RESULTS Based on the prelininary scoping effort described in this report, the proba-bility of a core celt accompanj'edbycontainment failure in Oyster Creek is assessed to be about 3.5 x 10 (or about 3 1/2 chances out of 10,000) per year. This would nean that, over the remaining 30-year designed li f et ime of the Oyster Creekgreactor, the chance of a core pelt with containment failure is about 1 x 10 ~# (1 chance in 100), although relatively c.inor riodif ications to the plant appear availabic which could signi ficantly reduce this value. This result is based on several simpl.ifying asaunptions which are stated belov- further, the relatively limited resources available fr this effort did not allow .n analysis fully comparabic to that in WASli-1400 (which ccst about $5 million) . The result should thus be considered no more than a pr ;ninary estimate, subject to revision on the basis of comments and review.
Also, the result is strongly dependent on the probability assumed for failure of " scram" to occur in a transient-induced accident. There does not appear to be a sound basis tendily available at this tire for selecting a " scram" failure probability. The value used in this study is based on a sinplified analysis which -
relics on assumptions f rom other sources which review may show are not appropriate to use. The failure probability value thus calculated is at the high end of a rather wide spectrum of values derived by other investigators. Substantial additional effort, probchly core properly a federal rather than a state respon-sibility, appears needed in this area in order to establish in the future a value which could be used with greater confidence. But, given the basic goal of the larger state effort to protect human I!fe in the event of an acccident, this value provides useful guidance.
IV. AN ALY SIS This section describes the analytical evaluations which were undertaken in pursuit of the overall effort. These evaluations proceeded in four discrete parts, as follows:
A. Determination of the type of accident s with the pot ent !al for catastrophic off-site consequences (one or mare short-term radiation deaths to the human population around t he Oyn t e r C rt _e L_ n i t e ) .
In order for rndiation death to oteur to any individual out side the reactor site boundary (" exclusion area") of a reactor site, radioactive mate 'al from
( l
(
1 Prelfrinary Draft the reactor rust be carried to the vicinity of the individual in suf ficient ,
quant it ies such that the result ing cadiat ion-induced cell damage is extensive (
enough to cause r.ortality. The only source of radioact ivity within the plant that has any substar.tial potential for releasing such quantities of radio-active naterial is the reacter core. The spent fuel storage pool could conceivably contain, vithin the stored fuel, a quantity of radioactive materini approaching that contcined in the core. However, as assessed in WASH-1400, the probability of a significant release fran the spent fuel stored at the plant is negligible, on the order of one in onc aillion er less per reactor per year; 1 concur with this assess:ent. It is also generally accepted that the only conceivable pay f or the radioactive caterial in the core to be transported of f-site would be for a meltdown
' of the core to occur in conjunction with an above-ground failure of the integrity of the contcinment structure which houses the reactor system. The rationale for this conclusion is given in VASH-1400. It was thus assumed, for the purposes of _
this effort, that only accidents in Oyster Creek which are postulated tu result in loss of tbove-ground containment integrity along with core meltdown will have the potential for causin; inredicte human fatalities outside the site exclusion area.
B. Det+rnination of the sienificant accident sequences in WASH '1400 leading to core telt and lors of above-tround containment intenrity.
After extensive investigation over a period of four years, the authors of WASH-1400 concluded that cnly about 60 accident sequences were of any substantial signifi-cance in assessing the general probability of a core meltdown in a BWR such as Oyster Creci.. These sequences are shown in Appendix V, pg. V-27 of WASH-1400.
As part of this effert, a further analysis was made of the relative signifi-cance of these 60 accident sequences. The result of this analysis is that two accident sequenecs drainate the probability of BUP, core meltdown to the extent that all the others (scre 58) vould increase the core melt probability by only 4% over that contributcd by the two dominant sequences.
For this recson, the aralysis below focuses only on the two dominant sequences. These
- tuo sequences consist of accidents initiated by a severe an anticipated transient (centioned above in Section 111), which would comentarily but drastically disturb the heat production and transfer processes in the reactor. Following the occurrence of certain kinds of anticipated transients, certain safety systems (including ones that vould totally shut down the reactor) must successfully operate in order to prevent a core teltdown, a release s f radioactivity into the containment tailding and an eventual loss of containment integrity. The failure of the reactor shutdown systems following an anticipated transient constitutes one of the two major accident sequences leading to core celt. The second sequence involves the f ailure of the decay heat renoval systems to prevent core melt after the transient oc c u r,s . Appendix A to this report contains a detailed numerical assessment of all of the dominant accident sequences considered in WASH-1400 and illustrates the dominant role of the two sequences discussed above. Thus, if there arc no factors specific for the Oyster Creek plant which would drastically inercase the significance of the other 58 accident sequences or decrease the signif2,cance of the two dominant sequences, then the probabilty of a core melt in Oyster Creek will also be dominated by the probability of these two accident sequences. _The next subsection considers this area in niore detail.
(
Prelininary Draft C. Determination of site end plant factors specific f or Oyst er Creek wi t h
. _the potential for chanc.ing the BWR core melt accident probability as ansesned in WASil-1400.
The Oyster Creek plant's design and characteristics were investigated to deter-mine: (a) if the 58 insignificant BWR core melt accident sequences in WASil-1400 could become significant f or Oyster Creek, and (b) whether, and to what extent , the probabilities of individual occurrences in the two dominant core nelt sequences in WASH-1400 might need to be altered for specific application to Oyster Creek.
The WASH-1400 assessment of E'p'j) core melt eccident probability is based on the design of the Peach Ecttom 11 reactor. Although the General Electric Company designed and supplied both the Peach Bottom II and Oyster Creek reactors, several design and site-related differences were found which have the potential of naking the direct application of the WASH-1400 results to Oyster Creek inaptopriate.
These differences and an assesument of their significance are:
1, frequency of Anticiosted Transients. According to UASH-1400, the average ftequency of anticipated transients in a BUR is about 1C per year per reactor. This value was based on reactor operating experience for 1972 (UASH-1400, pg. V-36). Table I shows the actual number of anticipated transients that have occurred ct Oyster Creck; it is based on scal-annual operating reporte for Oyster Creek (References 4 through 16). The first column gives the time period covered, and the second coluan gives the number of reactor shutdowns (" scrams") which were precipitated by the trgypjents. Since no scram failures occurred during the period frem 1970 through 1975 , then the number of scrams which were initiated by anticipated transients is also the number of anticipated trancients. (All anti-cipated transients, by definition, result in a reactor condition requiring scram.)
TABLE 1 Frequency of Anticipated Transient-Induced Scrats for the Oyster Creek Power Plant Time Period Anticipated Transient-Induced Scrams 7/1/75 to 12/31/75 3 1/1/75 to 6/30/75 1 7/1/74 to 12/31/74 1 1/1/74 to 6/30/74 1 7/1/73 to 12/31/73 1 1/1/73 to 6/30/73 6 7/1/72 to 12/31/72 4 1/1/72 to 6/30/72 4 7/1/72 to 12/31/71 1 1/1/71 to 6/30/71 0 7/1/70 to 12/31/70 6 1/1/70 to 6/30/70 _8 Total for the Period 1/1/70 to 12/31/75 36
~6-P'
- . .- -. - - - . = -.~. - .- - -. .
Preliminary Draft As can be seen from Table I, 36 scrams occurred as a result of anticipated tran-sients over the 6 year operating history covered in the table. This results in an average of 6 anticipated transients per year, as compared with the 10 assumed in WASH-1400. However, for the first year covered in the table (1970), an inordi-natelygjpgnumberofanticipatedtransientsoccurredwhich,accordingtothe utility , were the result of problems associated with the f act that the plant was just starting to operate. If 1970 is ignored, the tabic shows an average of 4.4 anticipated transients per year. Experience stree the last half of 1973 indicates that, sdth the- notable exception of the last half of 1975, an even lower tate eight be expected in the future. For the purposes of this study, an average frequency of 4 anticipated transients per year was selected as representative of the Oyster Creek plant. (The effect of assuming different anticipated transient frequencies for Oyster Creek, to covet the full range of experience, is given in Appendix B, along with other sensitivity studies. These results are discussed in a later section of this report.)
- 2. Differences in Ercrgency Core Cooling Systems ECCS). *ihe Oyster Creek plant does not have a so-called high pressure coolant injection system (HPCIS) as part of its ECCS, Peach Bottom II, the reactor used as the model for BWKs in UA5H-1400, has such a system. The system is provided (along with other redundan t systems) to'pr'tect o the reactor core from overneating in the event that a small pipe rupture cecurs in the primary cooling system. This dif-ference would indicate that the probability of core melt in Oyster Creek would be greater from s all breaks than that assessed for Peach Bottom II in WASH-1400. However, as an upper limit, it can be shown that if the HPCIS is assumed to always fail in Peach Bottom (this is equivalent to not having the system), the overall core celt probability would not significantly change; this can be verified by the information contained in Table A in Appendix A. The HPCIS failure proba-bility is included in the small pipe break accident sequences (details are included .
in Appendix A). Since the failure probability of the HPCIS was assessed to be about one in ten in WASH-1400 (Appendix II, pg.11-394 et se q -. ) , increasing the
, total contribution of these specific core melt accident probabilities by a factor of 10 would be equivalent to assuming the HPCIS always f ailed (or did not exist) .
As can be seen from Table A-1 in Appendix A, this increase vould not result in an accident involving a failure of the HPCIS dominating the a ;idents initiated by
- anticipated transients. In addition, the Oyster Creek-reactor has a system not l included on Peach Bottom II (called the-isolation condenser s designed to provide reliable core protection for small breaks {ggymThis .
or ICS) systemwhich is will be described in detail later.
l L A second diffe'rence in emergency core cooling systems between Peach Bottom 11
( and Oyster Creek exists relative to the low pressure coolant inj ec. tion systems.
l The Peach Bottem Il reactor has systems which sdll both spray and flood the core, i cach of which is designed to provide emergency coolant to the reactor core in the i
-event of a loss of coolant accident. The core spray systen sprays water down on j the core from above, and the core flooding system delivers coolant to the reactor vessel from the bottom to reflood the core. In the Oyster Creek reactor, there is l
- no core flooding system. Instead, two redundant core spray systems are provided.
l (The reason for these_ dif ferences is expla.ined in Item 5 following.) As shown in e
WW
Preliminary Draft VASH-1400 (Appendix 11), either the core spray system or the core. flooding system is assumed to provide adequate core cooling for Peach bottom 11 in the event of a loss-of-coolant accident. in addition, depending on the type of accident as well as assumptions made regarding the operation of the safety systems, various combi-nations of particl operation of the two systems can provide adequate core heat removal. An equivalent analysis for the Oysster Creek plant has apparently not been done either by the utility or by the U.S. Nuclear Regulatory Conmission or its predecessor, the U.S. Atonic Energy Coemission; it is also beyond the scope of this effort. Such an analysis is required in order to more accurately estimate the differences in core ec1t probability as a result of the detoiled differences in these emergency core cooling syste=s for Peach Bottom as compared to Oyster Crcch. However, in view of the fact that the systems on both reactors appcar to be generally si:ailar in terns of components, emergency activating signals, general arrangcaent, etc., there is no obvious reason to expect that the failure -
probabil; ties are substantially dif f erent. . Al s o , the contribution to core nelt fro: accidente thich thace systenc are designed to mitigate is very small (as illustrated in Appendix A).
However, it should be noted that the Oyster Creek dual core spray system could '
be core susceptiele to so-called " common mode failures" than the Peach Bottom IT systems, since the two Oyster Creek systems are essentially identical. (An exa:ple of a cotron code failure for the Oyster Creek system would be the plugging of all core spray nozzles in both systens by sone mechanism.) The complete assess-cent of possibic conron node contributions is beyond the scope of this effort; however, due'to testing and caintenence procedures prescribed for the Oyster Creek plant, significant common code contributions would not be expected.
A further concideration relative to the differences between Peach Botto: 11 andOysterCreekistheeffectivenes[27{spraysas core, pp sed to flooding in cooling a it has been recently asserted , based on the results of foreign core spray tests, that the effectiveness of core spray cooling may be less than pre-viously assumed. Since Oyster Creek depends exclusively on spray cooling from its -
two low pressure injection systems in the event of a loss of coolant accident, this could nean that the probability of core melt could be greater if these systems are
. less ef f ec tive in fact than they have been assumed to be. However, recent tests by the Cencral Electric Company at their San Jose, California facility have shown that core spray cooling is roughly gg79ffective as has been assumed; additional confir-uatory tests, uccording to CPU , are planned by Exxon Nuclear at their Richland, Washington test facility, it was, therefore, assumed that no increase in core melt probability for Oyster Creek would occur due to the possible reduced cooling effectiveness of core sprays. The validity of this assumption should be reviewed when the Exxon Nuclear test results become available.
- 3. Reactor Protecti_ System. The Oyster Creek reactor core contains 137 control rods,pps(Shutdown)
all rf which are designed to be :.aserted quickly into the core when scram (rapid reactor shutdown) is called for; as noted above, scramming is always required following an anticipated transient. The Peach Bottom 11 core contains 185 control rods, primarily due to the fact that the Peach Bottom core is larger-due to its higher design power output. (See item 6 below).
l 1
l
k
(
_ Preliminary Draft _
The ef f ect of this dif ference in number of control rods velative to scram failure was not directly evaluated. Instead a preliminary, simplified, independent analysis was performed of scram f ailure probabilitics, since the analysis provided inWASHdtDyg""gbeenfoundtobeofquestionablevaliditybasedonpublished h
reviews . This analysis is presented in Appendix C; the results indicate that the scram failure ptobability for Oyster Creek is 1.4 x 10'f' (about 14 failures,ger hundred thousand) per demand, compared to the VASH-1400 assessment of 1.3 x 10 (about 13 failures per million) per demand for Peach Bottom. A sensitivity study which relates the effect of scram failure probability to core ec1t probability is given in Appendix h and is discussed further in a later section,
- 4. Recirculation Punp Chutdown. In the Peach Bottom II reactor, according to WASH-1400, the reactor recirculation pumps ire automatically shut off (" tripped")
when scram is called for following an enticipated transient. This pump trip (in conjunction with the manual injection of a liquid reactor puloon into the primary reactor coding system) is considered in WASH-1400 to be an ef f ec tive backup in the event of scram f ailure Oyster Crock has no automatic recirculationpumptrip((ppendixV,pg.V-41).
- thus, manual activation of thic systca vould have to be relied en in the event of scram failure. Based on information provided in Appendix III of WASH-1400, a probability of 0.5 (one out of two) was assumed for manual recirculation pump trip f ailure in the event of scram failure. This f ailure probability value is based on human reliability, and includes the fact that the pumps cust be tripped immediat ely (e.g. , wit hin a f ew seconds for the most serious transient and within a f ew minuts f or less serious ones) after the scram ' tlure in order to be effective. The effect of changes in this failure probability are discussed in a later section; details are provided in Append 1x B.
- 5. Location of Pricary System Recirculation Loop P,iping. In Peach botton II, all recirculation loop pipes are located at an elevation above the core. In Oyster Creek, the pump discharge portion of the recirculation piping is located below the level of the core. This could mean that, if a pipe rupture ocurred in that portion of a recirculation loop (five such loops exint in Oyster Creek) Selow the core IcVel, it would be difficult, perhaps imponsible, to maintain encosa water in the reactor vessel to reflood the core. For this reason, Oyster Creek depends on spray cooling (see Item 3 above). This analysis assumes that this difference is not significant overall, since core cooling f rom reflooding the reactor vessel is not relied upon in Oyster Creek as it is in Pcach Bottom II,
- 6. Plant Sinc. As noted earlier, at maximum operation the Peach Bottom YI reactor produces 1065 megavatts of cicetrical power, while Oyster Creek produces 620 megawatts. Because of this size dif ference, there are other plant design differences, includ ing :. (a) fewer main steam lines in Oyster Creek (2 vs. 4 for Peach Bottom II), (b) fever recirculation loops, (c) reduced pumping requirements for emergeccy cooling, and (d) smaller sizes of the core, reactor vessel, contain-ment vessel, etc, in Oyster Creek. None of these size-related differences signi-ficantly affect the accident prrbabilit ,ies .
Table II summarizes all of the dif ferences described above and indicates their relative significance.
9-e
- WM
( .
(,
,' Preliminary Draft
. . TAhLE 11 Differences Between Peach Bottom 11 and Oyster Creek Nuc1 car Reactors Significance in Terms of Change Differenec __,in Core Melt Probability
- 1. Trequency of Anticipated 4
~
pcr year assumed for Oyster Transients Creek, 10 per year assumed in WASH-1400
- a. No Righ Pressure rio t significant
~
Inj ection System in Oyster Creek
- b. Oyster Creek Plant Significant in providing residual contains an 1 solation heat removal capability following Condense'r System an anticipated scram (see Section D following)
- c. Oyster Creek plant contains Not significant two core spray systems rather than one spray and one inj ection systen
- 3. Reactor Protection System Failure probability assessed to be 1.4 x 10"" per dgmand for Oyster Creek vs.1.3 x 10 in WASH-1400
- 4. Recirculation Pump Trip RPT failure assessed at 0.5 for i (RPT) not automatic with Oycter Creek; it is considered 1 scram in Oyster Creek negligible in UASH-1400
- 5. Some recirculation loop piping Not significant loc _ced below core in Dyster Creek
- 6. Dyster Creek smaller than Not signficant WASH-1400 reactor 6
. k Prelininary Draft D. Calculation of Core Melt Probability for Oyster Creek.
This section presents the method and results of the eniculation of core ncit probability for the Oyster Creek ractor. Due to the limited scope of this effort, the results need to be extensively qualified; future more detailed analysis might change them. The limitations of the result s are described below:
- 1. Qualifications and Limitations of the Result s. A fully comprehensive evaluation of the Oyster Creek core celt probability would entail substantially more effort than described herein. Ex tensive " event trees" would have to be constructed; as-built drawings of all the reactor sub-systems vould be required; all active components (pumps, valves, etc.) would have to be identified and available reliability data for each of these components would need to be gathered and co piled; extensive fault trees for these systems would have to be prepared; and substantial effort would be required in evaluating any common mode failures, design adequacy, human errors, test and maintenance factors, etc, specifically pertinent to Oyster Creek. In short, an effort similar in magnitude to that expended in U.SH-1400 to establish the core melt probability for Peach Bottom 11 would be required if one wished to ensure that the results approached the thoroughness of UASH-1400. The expense for such an effort would probably approach
$1 cillion (WASH-1400 cost $4 million), orders of magnitude above the resources available for the effort described herein. In order to accomplish this analysis within the available time and resources available, it was necessary to make several simplifying assumptions. The validity of these assumptions should be reviewed and evaluated by appropriate individuals and organizations before the results can be generally accepted. The caj or simplifying assumptions are:
- a. Except as noted elsewhere in this report, the failure probabilities for reacter systces involved in the core melt accident sequeaces are assumed to be identical f or Oyster Creek to those determined in UASH-1400. This assumpt ion isbasedonanecessarilybriefreviewofthefullOysterCreepjysignandon discussions with personnel f rom Jersey Central Power and Light , the owner of 7 Oyster Creek, as well as with appropriate of ficials of the State of New Jersey,
- b. There are no features or activities related to the Oyster Creek site (such as major nearby airports, LNG f acilities, etc.), which would, over the remaining lifetite of the plant, be significant in initiating an off-site accident (e.g., direct plane crash) leading to core melt. This assumption is based on a review of the site characteristic and environs as well as discussions with New Jersey state officials.
- 2. Description of Doninant Core Melt Accident Sequences. As discussed earlier in Section IIB, and as numerically assessed in Appendix A, which in turn is based on WASH-1400, there are two accident sequences which dominate the core melt probability in a BWR as analyzed in WASH-1400. These two sequences are also assumed likely to dominate the core melt probability for Oyster Creek (Section IIC). The tko accident sequences are: (a) the occurrence of an anti-cipated transient followed by the failure of systees designed to rapidly reduce core power , and 'h) the occurrence of an anticipated transient followed by an
~
Probability of 3.4x10 ~4 per year
~
core melt from anticipated tran-
[(4) (E .4x10-5)) . -
sient and failure reduce core power and 8 v'
Anticipated 8.4x10 -5 7, y Transients [(1.4x10-4) (0. 6) ] reduce core m e.:n -
4 per year and Failure of 0.6 per demand #II" * -
" scram" '*"*r"*
s_-
(0.5+0.1) shutdoan system .
~
1.4x10 ' por dc=and Or
./"%
I Manuni re-O.5 per demand cire. purp ,
).iquidpel-trip fall. 0.1 per demand son injec-rion fetitre FIC. 1 Computation of Prohnbility of Core Melt in the Oyoter Crcck Plant from an Anticipated _ Transient with Fcilure
preliminary Draft eventual failur- to remove cor e decay heat VASH-1400 assumes that. both of these accidents are followed by a rupture in the containment structure and thus lead to a
-release of radioactivity to t ie atmosphere. In order to understand the assessment of the likelihood of these tsn accident sequences in the Oyster Creek reactor, it is first necessary to understand the basic features of these hypothetical sequences.
-a.- Anticipated transient followed by influre to reduce core power. As noted above, an aniticipated transient is an event shich Icads to an imbalance in the core power production and core heat removal processes so that the core temperature begins to rise. Unicss core power production is reduced, such an occurrence could lead to core melt and a catastrophic accident. A discussion of which events can constitute an anticipated transient is discussed in Appendix V (Section 4.3, page V-36) of WASH-1400. As shova in Section III-C.1. of this report , the average frequency of anticipated transients at the Oyster Creek plant since 1970 is about 4 per year.
Following the occurrence of an anticipated transient , the required core power reduction can be accomplished in one of two ways. The first way (end the one which has always successfully occurred in the Oyster Creek plant and all other BRRs) is for the' automatic reactor protection system (the " scram" system) to rapidly insert control rods into the core.
'If the automatic scram systcm fails, immediate power reduction can be accom~
plished by the manual tripping (shutdown) of the reactor recirculation loop pumps. This step must then be followed within about ten minutes by the injection of a neutron-absarbing caterial (a reactor " poison"); this injection, which can only be accomplished as a result of manual activity by the operator, further reduces the probability of a core ocit.
In order to assess the probability of a core melt accident from an anticipated transient without core power shutdown in Oyster Creck, it is necessary to know the probabilities associated with all of the events described in the preceding discussion. Table 111 describes each event, summarizes the assessed probabili-ties both for Oyster Creek and the EWR analyzed in WASH-1400, and indicates where the assessed probability is found in this report and in WASH-1400 respec-tively.
With the information contained in Tahic 11, it is possibic to compute a probability for a core celt accident in Oyster Creek caused by an anticipated transient followed by failure to reduce core power. One way to perform such a computat, ion is to construct an " event tree" as was done in. WASH-1400. The event tree describes the inter-relationships i.etween the various events in Table III and, if enough probability inf ormation is known, leads directly to the computation of the proba-bility of the event of interest (in this case, core melt from an anticipated transient followed by f ailure to reduce core power) . Figure I shows such a com-putation associated with the event tree. As shown in Figure 1, the computed probability of a core melt accident from an anticipated transient event followed by f ailure to rgduce core power based on the assupptjons and analyses presented herein is 3.4 x 10 per year, about 3.4 times per hbad<ed thousand. Over the remain-ing 30-year lifetime of the plant, this means that the lihelihood of this accident is about one chance in 100. ,
l, Preliminary D7 aft TABLE III
- Assess:ent of Probabilities Associated with Anticipated Transients without Core Power Shutdown Probability F;v en t WASH-1400 Oyster Creek
- 1. Asticipated Transients: 10 (Appendix Y, 4 (Section 11-Trcquency per year Sec. 4.3) C.1.)
~ ~
- 2. "Scrau" systen failure 1.3x10 (Appen- 1.4x10 (Ap-probability, per demand dix II, Sec. 6.2) pendix A of this report)
- 3. Pecirculcting loop pump Ucgligible 0.5 (Sec. 111-C.4.)
failure probability, per demand -
- 4. Failure to nanually 0.1 (Appendix V. 0.1 (Same as WASH-introduce liquid poison Sec. 4.3) 1400 value)
- b. Anticipated transient followed by failure to remove deccy hent. The second accident sequence vnich was determined in WASH-1400 to contribute signi-ficantly (see Appendix A) to the BUR core nelt accident probability was a trar. stent-initiated tecident followed by f ailure to remove core decay heat (described on page V-52 of VA5H-1400) . Af ter reviewing the accident sequence described in VASE-1400 for Peach Bottom II and investigating decay heat removal systens in Oyster Creek, it was determined that the UASH-1400 analysis does not B. apply to Oyster Creek. Several reasons exist for this, but the main one is the existe:ce of the isolation condensor system (ICS) in Oyster Orcok. There is no co: parable system in Fench Eattom II. In order to determine the core melt probability in Oyster Creek from a transient accident followed by failure to remove decay heat, an independent asse.ssment was performed. This is descrlhed next.
Following an anticipated trcusient with successful core power reduction (scram), .
the ICS vould be used in Oyster Creek to remove decay heat. This system consists of a closed loop heat receval system which is connected to the primary cooling system. The ICS is normally isolated from the primary system by two valves in parallel. When the ICS is needed (such as for removal of decay heat), the valves are opened and the primary system cooling fluid f rom the core (mostly steam) flows through the ICS to a heat exchanger, is condensed, and the resulting water flows back into the bottom of the reactor vessel via the main recirculation lines. The ICS is arranged so that the cooling cycle occurs by natural convection; special punping power' is not required. In addition, the ICS valves are operated by DC power available from within the plant so that the loss of all AC power (e.g., from off-site) does not disabic the system. The valves are automatically activated f ollowing an anticipated transient. Heat is removed from the ICS heat exchanger by steam production; the steca is released to the outside atmosphere. The water supply for this heat removal is provided by a supply tank which can supply water for 1 3/4 hours following initial operation of the ICS. If the ICS fails (by failure of the valves to receive an opening signal when required, by f ailure of the DC power supply, or by_ mechanical f ailure of the valves), two other mechanisms are availabic W
( (.
Preliminary Draft to'recove beat. One is removal via the power conversion system (PCS); this requires that the mai" steam isolation valves (MSIV) be manually opened, the turbine bypass valves opened, and the feedvater pumps started. The HSIVs automatically close for cost anticipated transients, and the feedvater pumps trip off, that is, close. If both the ICS and the PCS fail, an automatic depressu-tization systen (ADS), designed to automatically activate, should come into play, reducing priaary system pressure, kten the pressure drops to a pre-set Icvel, the core spray injection systems activate automatically to maintain water in the reactor vessel for core cooling. The steam produced from the core is vented through the ADS relief valves to the water in the " torus", a water-holding chamber beneath the reactor. Heat is remos ed form this vater by the torus heat amoval systems (two redundant sys t cas) . 11e a t is transferred by these systems into a heat exchanger fed by vnter from the emergency service water system.
Since the ICS self-contained water supply lasts only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes, it is necessary, after this time has elapsed, to add additional water to the ICS. _
1 bis can be ace (eplished by either pumping water f rom a storage tanh or from the fire water system. The storage tank capacity is sufficient to supply adequate water to the ICS heat exchargtr for at Itast one day. The fire water supply is essentially unlimited, coming frem sources external to the plant.
Given that the ICS se.lf-contained water supply is adequate for 1 3/4 hours, it is convenient to consider the decay heat systems f or two time periods during the cccident, these decay heat renoval systems available at times les_s than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes and those systens available after I hours and 45 minutes. Fault trees were constructc d for both time periods and are shoum in Figures 2 and 3; the f ailure probabilities f or each system are also showm. These probabilitics were derived by various means as shown in Tchle IV. It should be recognized that the failure probability of some of the systems, notably the ICS, have not been deter-mined with the f ull rigor used in VASil-1400 since such quantification is beyond the scope of this effort. Instead, in these cases, failure probabilitics were assessed based on values for sicilar systcns in UASil-1400. The explanation for each systen failure probability is given in'the right hand column of Table IV.
Figure 2 shows the probability of f ailure to remove core decay heat within 13/4 -
in the Oyster Creek reactor. As can be hours following seen, this an probability, failure anticipated transien{y 1 x 10 (or about ene in one hundred million) is negligibic. Figure 3 abovs the probability of f ailure to remove decay af ter 1 3/4 hours (when the self-containedwatersupplyfortheICSisexhausted).
The result is 1.4 x 10 (about 1.4 times in one million) . Obviously, this second time period dominates the decay heat removal failure following anticipated _
trsnsient.
Assuming, as discussed previously, that the frequency of anticipated transi-ents is 4 per year, the probability of a core melt in the Oyster Creek reactor from an anticipated transient fogleredbyfailugetoremovedecayheat is assessed to be 4 multiplied by (1.4 x 10 ) or 5.6 x 10 per year (about 5.6 chances in a million) . Thus, over the remaining 30-year life of the reactor, the likelihood of a_ gore melt by this echanica is thirty times larger this value per year or, 1.68 x 10 (about one chance in 6,000).
. o Failure to re-nove decay heat ~
~
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 10
- 45 minutca -
.g-and -
Y Failure of ~
PCS fail-10 urc Failure of -2 isolation . low pressurc 1.3x10 condenser sy 3.
7 10 -3 heat renova:
or or
/~% .
ry
.(u Failure of ~3 Failure of DC power to 10 valves to 10
~
~
- "* h" i valves ADS failure
- open I*" V81 failure
-3 -2 5x10 1.3x10 and I
Failure of Failure of firac second .
valve valve
~
10 10' FIG. 2 - Computation of Probnbility of Failure to Recove Decay I! cat within'I hour 45 cinntet-
u--mu-
. i Failure to Re- ..
. cove Decay IIcat 1,4x10-6
, after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, . .
45 minutco .
and *
} }
Isolation -3 Failure to Y Condenser 10
-2 PCS failurc 7x10 remove de- 2x10
-2 Fniture pm?ay heat at vrevur .
1 Or r%,
k*
l Vciv2s Failure of
-2 es fa nure Failure to Torus heat fail water to 10 deliver va- removal 2]csed heat exchant .
ter to core, failure
-3 4,3x; 74 (negligibic) 5x10 1.3x10 -2 and -
and s . I l .
Failure of Failure to CSIS fail- Fire water cokdna ! f om fite "$3hj{ CSIS fail- injection )
storage 1 eater cynterit
-2 ure (2nd frflurn 0.1 0.1 4.8x10 sy:te-) l 0.1 0.1 FId.'3- Ccciputation of Frobability of Failure to Remove Decay Heat after I hour 45 minutec
_ _ _ _ _ - ' .I '
) *&
. Preliminary Analysis
- 3.
- 1:umerical Result of Oynter Creek Core Melt Probability. Based on the analyses presented in previous sections of this report, the Oyster Creek core melt accident probability is dominated by an anticipated transient followed by failure to shut t'oun core power generat ion. 1he total core melt probability pe r year f ro, the tug dominant. accidents considered, as illustrated in Tabic V, is 3.46 x 10 , about 3 1/2 chances in ten thousand per year.
TABLE IV Assessed Tailure Probability f or Oyster Creek Systems System or Component Failure Probability Rationale for Failure Probability
~
DC power failure 10 pg. 11-88*
Tirst valve failure 10~ Appendix III*
Second valve failure 10~ Appendix 'III*
(cor.aon mode considerations)
~
PCS f ailure 7 x 10 pg. V-41*
~
ADS failure 5 x 10 pg. 11-404*
~
Torus heat recoval failure 1.3 x 10 pg.' I1-177 (similar to PWR LPRS)*
Valve fails closed 1:cgligible After being opened, valve failing closed is very unlihely, especially conpared to failure of water to ICS heat exchanger
~
CSIS failure (first system) 4.8 x 10 pg. 11-387*
~
CSIS failure (second system) 10 Appendix II*
(conmon mode consideration)
- from WASil-1400
_______e_.___._
. .. ( ( '
.' Preliminnvy tiraf t
. . . TAbhE Y Core Melt Probability for the Oyster Creek Reactor Core Melt Accident Probability (per_ year)_
~0 Anticipated transient without scram . 3.4 x 10
-6 Anticipated transient with failure to 5.6 x 10
- remove core decay heat after 1 3/4 hours - ,
TOTAL. 3.4 x 10 "
Taking into account both of these dominant sequences,.this probability means that, over the remaining 30-year design lif etime of the Oyster Creek reactor, the overall probabilityj of a core melt accident is 30~ times larger than the annual value, 1 x 10 , or about I chance in 100.
It would be of considerabic value to establish the uncertainty liraits for the Oyster Creek core nelt probability as assessed here. However, any computation of uncertainty linits vould involve substantial further effort requiring detailed hnowledge of the failure uncertainty associated with each component of each systen, the test and naintenance schedules, possible common mode contributions and their related uncertainties, etc. This information would have to be combined in a complex statistical manner in order to establish uncertainty limits and confidence
-intervals. This full quantification of these uncertainties is beyond the scope of the present effort. As an alternative, a sensitivity study was undertalen to determine the relative significance of factors involved in the computation of the Oyster Creek core nelt probability. The details of this study are presented in Appendix B. Table VI is a enrposite table showing the results of the study. The first colunn identifies the contributer used in the core melt probability compu-tation, and the second column gives the decrease factor in the core melt proba-bility for a factor of two decrease in the value used above for the contributor listed in the first column. Only those factors associated with the anticipated transient wi thout scram accident are considered since no other sequences are significant by comparison.
TABhE VI Core Melt Sensttivity Study l
Factor of Change in Core Melt ,Probabilit y' Core Melt Probability for Factor of 2 Change in the Core l
Contributor Melt Probability Contributor
- 1. Anticipated Transient Frequency 2
- 2. Scram Failure Probabilities 2
- 3. Recirculation l ump i
Trip Failure 1.7
- 4. Ibnual Poison injection Failure 1.1 Probability l
I I.
(. i 1
Prelininary- Draft i As shown in the table, the core nelt probability is most sensitive to the frequency of ant icipated t ransient s and the scran f ailure probability. In both cases, a reduction of a f actor of 2 would result in a sinilar reduction in core melt proba-bility. 1hc resul ts in Appendix h shows that if the probability of recirculation pump trips were to be made negligible (as was assumed in WASH-14g0 f or Peach Bottom 11), the core celt probability vould be reduced to about 2 x 10 or 1 chance in
$00, rather than 1 chance in 100, over the reeaining des:lgn operating life of the Oyster Crcok reactor.
The largest uncertainty, by f ar. is in the foregong analysis of scram failure probability. She simplified analysis given in Appendix A f or scram f ailure at Oyster Creek is undoubtedly inadequate in view of the importance of this f actor.
Ilowever, it is not clear that techniques readily exist at this time which would al' low the derivation of a value that could be applied with greater confidence. In view of these uncer tainties, a survey was undertaken to determine what scram failure probabilities have been derived by various investigators. The results of r.his sugvey are susnagized in Table VII. The table shows that values ranging f rom 4 x 10 to 1.5 x 10 The value used in this studyhave (1.4been x 10qug)ted forthe is at scram highf end ailure of probability.
this range, and therefore tends to produce a high value for the core melt probability. Table VII also includes a core power shutdown failure probability which includes the reliability of the cerbination of recirculation pump trip and manual poison injection as a backup to scram failure (evaluated only for WASH-1400 and this study).
It should be noted that, as a result of the U.S. Atonf E2pn rgy Comnission's investigation of anticipated transients without scram , General Electric Co. has preposed to incorporate automatic recircula matic poison injection as a backup to scram f ailure{jgp pump , it is tripnotandknown auto-what the current NRC requirements are f or the incorporation or retrofitting of such a system in EWRs. If such a sysgem were employed on Oyster Creek, and assuming a failure probability of 10 * (which should be casily attainable),
the core celt probability, based on the results of this study, could be reduced to one chance in over 3000 for the remaining lif e of the Oyster Creek reactor.
TABLE VII Survey of Core Power Shutdown Failure Probability Source Reference Scran Failure Probability Core Power Shutdown Failure Probability AEC 22 lx10
-4 and 4x10
-4 (1) --
~
WASH-1400 2 1.3x10 1.3x10~
~ -6 Fullwood 23 1.9x10 8. --
1.5x10gxlg2),and This Study _
1.4x10~4 8.4x10-5 (1) These values are based on data from a variety of reactors.
(2) Based on different statistical treatments of data from light water reactors.
. . .s '
, _ , . , Jfroliminary Draft V. REFERENCES ,
- 1. Oyster Creek Unit No. 1 Pacility Deneription and Safety Analysin Report, Jersey Central Power and Light Co.
- 2. ~ Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plant s, UASil-1400 (!!UREG-75/014) .
- 3. .Pinal Safety Analynin Report, Peach Botton Atomic Power Station Unit s No. 2 E. 3, Philadelphia Electric Co. (October 1970).
t
- 4. Soci-Annu:t Report No. 2, Provisional Operatinn License No. DPR-16, Dyster Creek Nuclear Generating Station Unit $1, Plant Operations
'(1/1/70 to 6/31/70).
$. Semi-Annual Report No. 3, Provisional 0,p_cfatinp, License No. DPR-16, Report of Operations (7/11/70 to 12/31/70), Jersey Central Power end '
Light Co., Oyster Creek, Nuclear Generatina Station.
-6. Ibid, Report No. 4, 1/1/71 to 6/30/71.
- 7. . Ibid, Report No. 5, 7/1/71 to 12/31/71.
- 8. Ibid, Report Nc'. 6, 1/1/72 to 6/30/72.
- 9. - Ibid, Report No. 7,'7/1/72.to 12/31/72. -
- 10. Ibid, Reporc No. 8,1/1/73 to 6/30/73. ,
.11. Ibid, Report No. 9, 7/1/73 to 12/31/73.
- 12. Ibid, Report No. 10. 1/1/74 to 6/30/74,
- 13. Ibid, Report No. 11, 7/1/74 to 12/31/74.
'14. Ibid, Report No. 12,.1/1/75 to 6/30/75.
- 15. Ibid,-Report No. 13, 7/1/75 to 12/31/75.
- 16. Operatinn History of Oyster Creek (through) 1974. Dept. of Environmental Protection (New Jersey), Bureau of Radiation Protection.
- 17. - Heeting at.!!orristown, N.J., June 7, 1976.
Attendees: Dr. Glenn Paulson, State of New Jersey, DCP l- Dr. Peter Preuss, State of Ne Jersey, DEP l'
Donald A. Ross, Jersey Central Power and Lipht L Thomas M. Crimmins, Jr ., CPU Service Corp.
Nick C. Trikouras, CPU Service Corp.
. P.R. Davis, Consultant, State of New Jersey ,
4
a
( (
- **
- Preliminary Draft i
- 18. Reactor Safety Study (WAsil-1400) : A Review of the Draft Report,
" EPA-$20/3-75-012 (August 1975).
- 19. Renetor Safety Study (WASil-1400) : A Review of the Final Report , U.S.
F.nvironmental Protection Agency (June 1976).
- 20. Oyster Creek 1;ucient Cencrating St ation, Docket 50-219, S h ic railure Analynis - ECCS System (June 24, 1975) (including At tacturent 11 ECCS Modi f icati on) .
- 21. Testic.ony of Birdenbaugh, llubbard, and Minor, Joint Committee on At omic Energy (11ay 13,1976) .
- 22. Technical Roport on Anticipated Transients Ulthout Scran for Unter-Cooled Power Reactors, WASil-1270 (Sept. 1973).
- 23. Estimation of Scrco Tallure Probcbility in a P_n;'esian Frerounrk, R.R. Fullwood,1976 Armuni llecting of the Ancrican Nuclear Society, Toronto, Canada (June 1976) (TASSA023 1-637).
- 24. Studies of EUR Designs for ititigation of Anticipated Transients Uithout Scrnms Z. V. Baysinger, et al, General Electric Co., t;LDO-20626 (Oct. 1964) (also Amendment 2, July 1975).
- 25. The flucient Re r;uln t ory Cor. mission' n Reac t or Sa f et y Study, J. Yellin, ltell Journal of Economics 2, 3171 f , 1976.
5 4
we .-
u Prelininary Draft APPEND 1X A AN EVALUATION OF THE RELATIVE SIGNIFICANCE
}
OF BWR CORE MELT ACCIDEST SEQUEPCES AS ASSESSED IN WASH-1400 According to the VASH-1400 analysis, there exist some 60 BWR hypothetical accident sequences which are " dominant" contributors to significant radioact1~vity relcoces (pg. V-27, Appendix V of WASH-1400). All of the sequencen result in core melt. In order to establish the relative significance of the sequences to determine which ones, if any, can be ignored, the sequences were organized in groups according to the initiating event. Repetitious sequences with di:ferent assumed modes of containment failure vere clicinated.
The five initiating events assumed in WASH-1400 are: (1) a large loss of coolant accident (LOCA), which is initiated by a pipe break equivalent to 6" in diameter or greater in the cooling systen; (2) a small LOCA (S y ), which is initiated by pipe breaks of equivalent diaceter in the 2" to 6" range; (3) a small LOCA (S,), which is initiated by pipe breaks 1/2" to 2" in equivalent diameter; (4) anticipated ,
transient-initiated accident sequences; and (5) accidents initiated by rupture of the reactor pressure vestel. The docinant core ec1t accident sequences associated with each initiating event with their probabilities as assessed in WASH-1400 are shown in Tabl e A-1. Symbels are used for each assumed event in the sequences.
These symbols are defined in Table A-2 and are identical to those used in VASH-1'.00.
Table A-3 shows the composite result of the core melt probabilitics as com-puted 1n Table A-1. As can be n +n, of the total accident probability $of 3.0 5
x 10 , the centribution fro: t sient-initiated accidents (2.9 x 10 )
dominates to the extent that t .ata1 probability from all other accidents contributes less than 4% to the uotal.
w.
f
( ,
,., f relistinary tiraf t, TAlsl.E A-1 .
1)orainant flypotlictical Cot e 11elt Accident Sequences for 1:ach Assumed Initiottog 1; vent Accident Sequence l'robabil i t y
~
n AE 1x10
~0 g8AJ 1x10 g 5 Alli 1x10,3 p S A1 1x10 L*S AGJ A!;G
- AGill '
~I Total l'robabilit y 13x10 ,
- SEI 2.1x1g)~
he SJ 3r10 ~g 4x10
^d8Sf1 g
f., S 111 2x10,9
$SSfC 8, S CJ 3x10,,g 3x10
" I 3.6x10
~b S GI 8, S El
- 3
_g -
S3 Gili 2x10 Total Trobability 3.83x10
~I
~b SJ 8x10 ~
2 h- S12 9x10 ~ n dE S 1. 9x10,,3
^J SE 4x10 .
${SC
-n 2
S cc
- Y" S,Citi 2 M S EG *
$ GJ
- S 01 *
~
Total Probability 3x10
& u f, W
- .
- 1.6x10
~b '
5 5' S TC 1.3x19 ~
OES TQUV 4x10 total Probability 2.9x10' hh k% 0xidizin- ctm' sphere 1x10~0
$8$3 Kon-oxidi w . Ateosphere 1x10" .
Total Probability 1.1x10 ~I
- Negilgible (leau than 1x10~0)
~21-
e g.g * * *
. l
. i
.' . 1 N.
- f$ 8~3 i
.U. .g 8 .
g hg 4
. ' ' .. g l ., . -
- p. . , , . .g
. . . . . , . , , . N (1
O l 8 e
,:. . .._ I.1 p . o.. r.t,l . :. 4 ' u
.: . l .. " .J ' ' l .: .' .*.. *. . . *t. *
. ....i1,s
. g t. ., .. . .,.. ..y.. .n ..
~ a .. a u,i ..s.: t . , ::s
'**. .ln.:. .
, .. t 4* 8 O. , ,n,, .9: .-. p . -1.. o. . ' .. . ,1. ?. ! i t* 1-].?;43,,. go y
- . 3 ,1.. . ' , . t. .'L* t.,.'.
y . * -
U
.. .i i
- 3 [, . , . .
g n H .
'.. . : "s :. ., .. . *..* r.l'. . . ..;- . (l . : . ". I ? : 4., . i . 8 , .,, .
. i.' 4; ;;..' r4 er e g
1
- :.5 ,
'I'... ..e .
. ; IJ g
. . oN ' . ! . ....
. p :. *f..*.
. . .g .. .. r. 1:, -..
- ;;.*. t.......i"fl....;. t H "3 *4 * . ;
- ' . . 3. I . . /
ouh a e' j.t. l. f. - sn
',...t... ..
{
. ..{ . . .... ;. . . .. jf3g ._..:.,.... .l;; ,~., q : . ._'. ml .- ; q 7 2 p., a 8
!. . 3
. L 3 ; s t . w .. ....,7.. j:. .m. ;;,;q ,. g. r:, . j.
- o *d '
.).} .e . .
. . h ,4 g . . t. , .l , . .3. , . g y
. ,.,.r,* i. n. .e. ,6 .. La e4 gg
- e
- 8
.. .). ., *p4 ffa ,l ,m _ _p. r I : . ,..p....,....,
. , .;,,g f .,.sy .g n ,r3
.l.
.....t..,...g,<..'
. . . .. .J.....g..
- .e.. . ;,, . g y9y
. ~*
4
. . ' " e-4 qe4 N *
. .. -. ..u* :g i .*.l:.. f.
.?;t J .. . * .O
. ;, . i.,,:3. e 1. . .
.w ~ . . . <a l..'.. ....*
. . .. . . . . . , . . .-3. . .n ,t,, ,. ;, ...; r . . v . . ., :. .n... ,. p. . .. i. .:
.....1 * . , . . n .a o u
- s. : J T .* . " *.'_ .ut :, .:.'.y *:k.1. ;.. . a. . ..noa . a n.
. ' , . .n, , u ! .
Y'% . .m,L.f. /s < o o,
. . ... , n.:.1:.:.
.i.._.::. . .s....x.-
- , , r .. : .
l.: . .:..n . g. L 1: auoo,n, oua
. .. ,., ,. .. .i
.+. , 7, . . ,
u,u .
. m. .... . .
e .; .* . . r . . . l . i.". s . sp l. ' .:. * ... r.j.:J.:.s . . r.t p.a ;p . _..*a.ei!.. ..8 .. .g :>t;.
.:. l. : .n.
m u
- 4. .. la;. w
- ,. i.:...... . .
es u nou -*
.p;. ;d. .t
. s .i . ."-
i... .
.t .p... '...-.
... p.
. ; n....
. F. .
. >:. ., s.o,: 'o u o
!< ...;: ,W. . .w,....
.r
. . e. I .: . l. . :..
. . .a,. .. . ., ..::yuaw a o
,s,. . a <, . . . .* .. . .. .. . .. e.. .
1 0
'. u-[ = : aay (**>
4 ?.... . i.P4 . e4 11 =.- * .
- a
~
- - 2.-,
-- .-r,.::. O. *n e o...;..
1 .-r n . 9.rv. .id ,. . .. yT.: ": e - - H N .a 'I.:,k
- .y .. i'.s ".W r ; !Mi rr.'s. 1 ..,
.. :; ..- :1 .. . . .. 7 yl,.
- 3. . . . . v. v,.,,: ,, i wn
. . .f, ..
.?
..~'r.. . .
.. . ,:. ..,.. .... ;.<:.r. . . . . . .,:. . : .
e u3 u
.-. -.. .- ; ..p.
u s ,... . . . . .. .i . . . . . . . .
,. .- . ..... .. .. . . . , . ..:.s.... t. .,. .3.ua .,s . ,. 3
. :_ . . ; _, - .. 3 .: . x aa o
. b c, .
.s t .: ; y. p er,. u , :..: e o. p* u
.. .y. .! . ..: . ., .. , . .. ... " .
. up. . .:. .... : m<
v, ,
.er ...r.i;. .... . .
-:-r,,: ..o .. . . .. ..m- . , = . . . . .: . <.:.. .e . . s, i.
. . ..... . . :: . . m .
. . . . m. .
,. + . : , , ;.r....t.n.",_._.._.tm.._.._..,,._.-
l ,. ... .
. 1..,
. _ _ j'. _- . .. m:, :.- :::. ,, ; w . o, t ...g. . t .
. .. .. . gg , .. . . . .. .
..,.j, :
- L ; .J. . .- . t i .*. ,.f.l .: . r 1&,. .::
8 .. . . . . . , ' . .
Lt 4* :I' ,-
- ,i .- '
r2
- rn J!..- .
-.3,. ....8e
) ,,."
. .. 67 g
u
. ;
- 6... ....,,,,--
. j . t;. ;.g ::..: -- . . ..
g.: L
- . ~ ; l. ' ;*! t , : 4': .;' .
4 . '( 3 ;, L. ;.- N . c
,.3* * . g ;
E 1
- [. * .t . * ,
.( -.1 .I y .,* . r s ." -.j-;.:J . . . b:
. r .*,q,.... i.: t.- O. p
,.,. ::;. ..11..
4 .
1 .: ? . .v.. . 4r . "gt .. :*+ l . . , J.L;. . br. :. .; ,[2
.. .. r. *; v :.: s eg:
. I. . .
t o, .
3 .. ..1. .
.g . . l. ... .
n m
m 1.
..- .i.. :.
J *. . .. .,
O N" e (
- w. . . . .. .. . .. ..
. r v. t.a. a.s ' .< ..
.-< en e4 '
Ix) at'M/X1 TIT 9v903d man DJo3 . Q 01q m A/ N IT M oJd 3i g DJ M e
............,-1..-.,..~..,.n..1 -
4.. s . , . . . . . . . -
. l. . . q.
o.
.....:... . . < v. r " - i. ...i n:. r. . .;.i u
(
.. i.
,. e4
- t. .o r.. :.. ..; 'l . a : ,.c.:. . . j .:v .
a vs " *
- ' .I;?< .! 1. . N *C 'r,J. :.'..:*.T.:!:i ;b q.; .i'- u N ,A .o i * ... *. : i, .. ...*
l' , ' .. .. +.
4.; - ~ .v,;.
. . ..: :: . :. . :. -?.
- o4 oo . . . . -
.l l..: c'* O.
.m .l.
vc
. , t g
-: . :u :r r... t:. . * ;:.
= .- r.:. - i .q s,s:-):..
r~.. . , r ::.: t:.. ~r- *
. .t e :. . .d. : .....r. . .!
- s n
. 1.
.t. .. . .
. I;. : .
N
-p- .
- p :. ._.
- *. ;:- a..: :e im. .q: .
r-
. :a.:. j; :au _ .r; j. f..
e, e wa y, , w . .
- -h; l'
.n
. u a
.. A . . L. r.e F . :;-.3 ....;; .
- ;-.. to.
dh OM
- s. .t,' t.i.:.;d ..: g 8.,.ou
. : }i ,7. 1 . .. .,
- : _.: r. 'I .'t.:.3."
. . . . . . : *., q . .a:
- r .: .:,ttf. at
- 2:
- r. .- .
1 . .* . . : ..
g ,' .. . , - . * ; ,8 :.-
- f. . .ay { g g d, p p
- J'. . a j' .?. :e :. tra. . *
.- . . . ., *i U 4 2... l . .: .:L ..=. t :];.: 1,
...I***"I.
.. ' - ' * *' . *CO.h ' -t a ' . . v.
.e.. .'.
G .
- t<- rA e4 p
~~.t . :', l /.*.".;* . : . l. ' v: . :g . + r:
~r
- . . : '. .* .e. :.. P? ;;3.l. ' * :: 1. ; . ~ , . .
- . x * *}. - - * ;;
- w ". . . . t *~i '. ' ' *. :
y g 4*.s. . %. .* ."'a ' t~. -;=" . .4 p r3
"..:.J t ' *: *; **: 'I ' b * 'I *O ON
. . . .... . .. b - .
- W . .T
. I'? : . . .j I'* t .: 7 pi' 8
- 1.
- o***I .
+ = -*
j .
U k O
- . ".:f.*.;.*J ". .P : ' j 6. .*. * .?,7. U*?. 4. *. *.. ';. ,4 3 t ***m.
I *.,.. . "-* /T. .;t- , .- ..
- H Q .U r( ug .
, ? .* . .
O ,a 34 W
. p, y y .
'n - - - - ; - .- - . ooc
- r. .: l /, ). ... : l .:. . . 9 . .,r :.-. ...3. l. .)i-f.
- .; s. m . b l .. .~ ,.. .
, ep
<* u=
. -n r. . ! ;..a..*.- . v ,
..e
- t - .
- 9 y a u .;.i - .. r-. t$ u u
. V.: ..3.:. . . p.. n: .: r : .- .:: .m a,; r. . g. : . ~r s,) uuu
,:.1 ~ :.
tv.1 ::- v :.. o e
m a,d w :s
. . . .. ' . 3'* i. .
. :. . --- : . . .,,-. >:. r. .. ....f. . c. , - - .- 5.
o ON !
. . I*u . . . . . . gr . ' . O '*4
- - . ... . .,.. a.",... .,:
- * -~ 4 . ; A.
kJ **d s .
- 4.-i- .v g;;s.uo ',,, p,:,a y ,:. ;;;;.;.): .,," *
'* ari U . * .. c) :s
! .. ta .i' * *
% :: -p.. mu ... +
' m.4 o
.! : l. . '. * . t. u. ; ..: .L. ;.., . .w.
- - 8
. ..: .r: .- .:. .. :. .L ~ ,. ., . f: + % =1* . , *. .. .. . . J.
' a% : .. 8 I*, N
- ,r. ' J ~- .' I* ,0
. I, g *
- L .
- ! . + *.
- si-*' n4 () .
.L. U O
. ;. *: .I **/ 4 - .. ..
5 ,
- . . :.' g '*g.?. .
!'. 54 p *4
- i. g : .1t8r.' 3 * *
.g.
k..*,
p v.8' 5*' '.. 1-#' 7 t ;!! A.":"*. ];* : .5.l: d.**
(4 e4
- * ' * ! * .' ..i
- f4 wH ps u
, :." .J. . .:. % o.. " '::rc.: .a.r.r.
. . . . I,.*d; ': . T .:. . 4 * :l-"" r "
- w
..ap.
s s . . t. . N ' I. ..* .-.. '4: +'# 'l gN . . '" P. '= - 4j
. 4. ..:. .. ... or att (.
..-...c. . . p a . . - ' ' . ' *t'.'el. .- 4.,
. . + . .. . . - ?.G4 ' g '.4;., 74 C 0.
- "'l e :.t*, ". . . 6 * . - .;
- s ;,r 4 * *
- es s p
' ] . .' ... *' . "* .* -* '* '.. .g' *'.
- H I L
- j*. '?: 1 g ". *. '.* f- . " . ~.
- ;*.: t
- *...*ll. *.*a.*
f
. .I . ..
- .i.
' * *-..."*;; , #1.: .y'h
- L l ' . . ' g* . .. ...-
- ..;"I ..a.
- ' .. b.** g l 4-.
s'*>
- 4 ;*
. .'..*~3 e... 3.**. .f=
.l**
g I . .'. ' . .- j. *.
. : : .,.9. * ..'.; t ': t ;. * * * ; 17 '
. . .V:".".. -- - - --g . L. . * .1'. :.(.n : - grg ,C% g , e a . _, t ..' *
- d 1 ; ' a 9.. d2 :~ I . 4 . .. N O
- :V.
- m ,. : : f . ,
%. . ' - O 1 ..*..g
- 2. * .;. * $ ,.1.*. . . . . .. -. ., .. .., T..: ..: . V.. J.t ,'..*.l.....t.., * .:. .
1 .,4 ..< .e g.l)
' ... art SQ .. .e . .
~ . *.. 4 f; SJ ..: . I '. t b^. $. . l.".
J ' ' . g . . .*.'.:**[ 7.l * * . . . . . . ., . t * * . .* "]. J. b : ;- ? , **t *i~.' ..k f(-* . S o, ;, . g '! Y
- 71 ' L : . . . . .*;.I.!. *
- J :..""1.. *1 d O
- .
- e.-
l * * *i .
A- *~4
. ! . * { *. p .T* . *. -* -.
W
. * *
- r':g N #* 4...a -a-.
-r.s*e. * ',
Z. *"W.:;".~ .. :; *... *** * ~., *' :... .. ., . g , t . . ..l. * ... . - - . . ..
- p;v.: .:.... . . . . ..v.. . : v ' -
.:...t.**.'.' .f 4*
$ e., 3 . . . . *g- .
- ..
- M
. . + . e ' . ,p
~ .
. e k* "' "J ! e .
- J:.. :.. .::4 :.:;:. .t .s:.d .
. :....s. . .-
-, t ,* .t,4 ..
. % ; ;+ ... -<
.: n, 1
t
\ =
" r: . r' p& .
N
- . -:.o . . .:: 4.r.a. v.*:
,n * ' g . :- 5. *
- q
. g g
.g
- g. .
p.3 e .-
s y-c, s .. ,
- l1I n ~ ~
a -
s e
n r 6 eeu -
d .l 0 re a 1 e
c
- r. i f 9 7
1 y
- t t
. c i ns l d : e i nnr aeu
~
0 b
a tl 1 d t ni x e sir 6 r
," eaf F
~ t r 2 e
0 r 1
x u 4 n 1 e r
. =ke cc rr .
cC .
s r
re f ot o f s r -. y e "n m o -
eC rae e
urt rr l cs To i cy f a o t F" l u
s .
a s
m r i d n e a 1 h 1 -
c 1 e 1 t p
- m
'. h ., d i s
cmy c
es,
'01
~ e r .
x l d iy n 7 o 1 oc, 3 r -
r r t c 1 n
- e - .
c c
. - i ht r i
.. w
- . d
.s .- e t
a i
c o -
' , s c e s e
A er 8 .
l u -
t i
bl 0 ui 1 o .
r df oa x n 2 .
3 e .
n<
l ,I ll
- l \1 l.II \ 1Ill lll 1 l f (
(
- .- Er211minary Drnft TAllLE A-2 Definitions of Accli,ut Sequence Symboln Svnbol Definitton A hupture of reactor coolont boundary with en equivalent dianeter of greater than 6 inches E Failure of energency core cooling injection C Failure of contaircent isolation to Ifmit leakage to less than 100 volume percent per day 11 Failure of core spray recirculation nystem 1 Failure of low pressure recirculation system J Failure of high pressute service water systen 3
Sna11 pipe break with an equivalent diameter of about 2 - 6 inchec C Failure of reactor protection (shutdown) syrtem 5
2 S::all pipe break with an equivalent :tameter of about 1/2 - 2 inches T Trancient event k' Failure to remove residual core heat Q Failure of nornal f eedvater system to provide core nakeup vater U Failure of IIPCI or RCIC to provide core makeup water V Failure of low-pressure ECCS to provide core makeup water TAELE A-3 Su:tmation of Accident Probab11 tics Accident _ Probability
~
Larpe LOCA 1.3x10
~
Sy LOCA 3.8x10 S 1.0C A 3x10 3
Transients 2.9x10
~
Prersure Vessel Rupture 1.1x10 Total Probability 3.0x10'
- 1
_22-I.__.___.____..___.__._._._..______
.- ( ( . 1 1
Prel f t.11 nary Draf t l 4 4
. .o l APPgNDIX H SESSITIVITY 57UD10S in lieu of a definitive assessment of the unceitainty associated with the Oyt.ter Crt-ek core celt probcbility as computed in this study, a sensitivity study van '
under takt n. A definitive uncertainty assessment would involve, as explained in Section IV D.3, e substantial ef fort beyond the scope of this study. ;
t This sensitivity study establishes the relationship between variatiers f n f actors i involved fu the enticipated accident without core power shutdown accident end the '
i core celt probability for Oyster Creek. The factors involved in the accident sequence arci (1) frequency of anticipated transients; (2) failure of the scram systca; (3) f ailure of canual recirculation pump trip; and (4) lailure of manual poison injection.- Figures B-1 through b-4 show these relationships. The curvea on the figures show the reintionships between the variation in one of the four accident f actore (as indicated below the abselssa) and tue core nelt probability (ordinate',
r.s calculated over the remaining life of the Oyster Creek reactor. The arrow on the abseissa of the figures indicates the cocinal value selected in establishing the core esit probsbility.
Tigure B-1 shows the drainating influence of the antiefpated trancient frequency. ,
For a giver change in the anticipated transient frequency, there in 6 corresponding and equal-change in the core nelt probability.
Figure L-2 shows the equally dominating influence of the fai' lure of the scram syste . Accin, a change in scran failure probability produces and equal change in core celt probchility. .
Tigure B-3 shevs the influence of recirculation pump trip failure on tie core ceit probability. As can be seen, the influence is quite strong, though not
- as decinating as scram failure or enticipated transient frequency.
The manual leison injection failure probability influence io shown in Tigure B-4. The influence is not great, since the recirculation pump trip muct alco occur to shut dovm core paver, and the ' recirculation purp trip failure proba-bility is larger and dooinates the failure probability of the combination. ,
The general formula used to compute the relationships was dortved as follows: -
. 1. yearly core colt probability (CMP) =' yearly f requency of anticipated tran-
. 'sients -( AT) cultiplied by the f ailure probability of core power shutdown l (CPSD), or
+
CMP = AT x CPSD l
I vhere,
- W* 6 e
e
.. t_. _ . _ _ _ -_ _ _ _ .-_ _ - _ . _ . _ . . . _ _ . _ . . _ . _ - , _ . . _ _ . . .
k .
Preliminnry Droft
. 2. CPSD = scram f ailure probability (SF) multiplied by falluto of alternate
- 3. A.M PCS = failure of rectreulation purap trip (RPT) plus f a!!ure of snanual poison injectico (itP1), or MIPCS = RPT + HP1
It should be noted that the raitionship expressed fri c:quation 4 is valid only if the core nelt probability fron anticipated transients followed by failure to rennvc decay heat is cuch less than the failure of core power shutdown following an anticipated transient.
4 e
- 4
- e
~24-gilmenos>
%., ,_, .- _- , _ . - , 4 - w - ~ '-"
(
('
Preliminany _ Draft AppC!mlX C ASSESSMENT Op OYSTER CREER SCRAM Pall.UME pg0HLh11.11Y As vau noted in Section 111-C.3, the core desig;n f or Oybter Creek containn 137 control rods US11e Peach bottom 11 contains 185. During t.crum. all of the rods in both cores are rapidly inserted. 1.. t he WAS11-1400 analysir. of the probability of f ailure to scram resulting f rom the f ailure of rod insertion in the l'each bottoci 11 renet or, it un achuned that the failure of any three adjacent control rods to y int.et t vauld conttitute sc ram f ailure. This assumpt ig'g described nr. " extremely g c on se rva t iv e" in VASU-1400, was criticited in reviews of the report. ~
Al so , the se t,ame reviews quest loned various other.aupects of the WASil-1400 neram failure analysen from rod failures for Peach bottom 11. In view of thene design differences, questionsble assumptions, nnd uncertainties, it van determined that on independent es.nen nent of the Oycter Cr eek scran f ailure probability should be undertchen.
- lhe f ailure probability foy' the int.crt ion of a single cont rol red was assent,cd
~
in VASH-1400 to be 1 x 10 (one in ten thousand) per demand. Thfu assens-incnt was based on fa' lure rate data conpilt-d from several ractors. To deter-rnine the applicability of this, rate to the Oyster Cre(k reactor, operating histories were reviewed. It van determined f rom study of the semi-annual operating reporte (keferences 4 through 16) that about 70 r,cramn have occurred in the plant since 1970.* About half of these (36) were the ter. ult of anticipated transients as indicated in Section Ill-C.1; the rest vere associated with tents, spurious bignals, planned shutdovan, etc. No reports could be found of any rod failures during these scrams., discussions with cpu Servigerporation and Jersey Contral Power and hight .
(owners of Oyster Creek) perconnel confitned that no rod feilures hnve occurred. Thus, since there are 137 control rods in the Oynter Creek core, there have'bcen 70 x 137 or 9590 rod insertions without one failure. (Ilowever , t here -
have been minor instances of slow rod insertions and it,ilure to insert completely (16) ,
These proalens did not result in scram failute et even single rod f ailure.)
This would appear to confirn thd the WA3}l-1400 f ailure probability for single control rods (one f ailure in ten thousand) la at 1 cast that low f or Oyst er Creek.
This neans that the probability of any sing 1c4 ontrol rod in y rtion failing during scram in Oyster ::rcek is (137) (1 x 10 ) - 1.37 x 10 (about 1.4 Lines in a hundred).
Sinco severni rods must fall (at 1 cast three adjacent rods, according to WASH-1400), it would appear reasonabic to assume that a common mo3e failure would be the only signifleant nechanism which could cause enough rods in the right configuration to fail such that t he core power was not reduced, if only cost.pletely independent f ailures are considered, the_gjrghability of ang3 rods failing to insert would be approxiciately (1.37 x 10 ) , or 2.57 x 10 (about 2.6 times in a million) . *11:e probability of these three rods being
- Hote - Table 11-A of the Semi-Annual reports indicates the tot al number of scrams experienced through 1973. For nome unknown reason, thin inf ormation goinitted f rom Table 11-A in the repot t for the last hall of 1973 and doen not appear in any nuhuequent reports, l
L 25- !
( (..
o .
preliminary I)rnit -
.p-,. '
. adjncent (the minimum required to cause scram failure, based on the " extremely connervative" WASil-1400 asuumption) would be exceedingly remote.
On the other hand, according to WASil-1400, the failure of any single control rod vill be accompanied by at 1 cast one other rod being affccted by the name influte cause 10% of the time. Thuu, the probability of a single control rod f ailure thich is accgpanied by at leant one other god experiencing a r>lmildt problem in 1.37 x 10 (about 1.4 times in a thousund). Tabl e 111 3-5 (pr..
111-37/38 of UASit-1400) does shew that redundant conponento experience, on'the average, common mode f ailure about 10% of the times that single failure occur
-(the percentacco range f ro:n 3.4% to 33*;). WASil-1400 also rnaintaint (pg It-162) t h, for control rod f ailures, the connon mode r..echanism retault s in corplete f ai4ure of other rods only 10% of the tine. Thun, the probability et a co'amon n,.de failg'o caucing tro or more rods to inil in the Oycter Crcok reactor vould he 1.37
>. 10 (about 1.4 times in ten thourar.d).
It should be noted that two rod influres vill not cause scram failure; depending on the location of the rods and the charm teristico of the core at the time of the influres, it would take at least three adjacent rodo and perhaps, several hnre for most taps, llovever, if the common modc f ailure rtechanitz x i t. t s , it is likely thnt it vill af fect rpny todo (.i.e., more than just t wo) s.. % the rod deninn, tnaintenance, tor. ting, etc. is similar for all rodt,. (A similar argument is pre-cented in UASH-1400. Appendix XI, pages 5-2 and 5-3.) Thuc, based on thin o l '"-
p1111ed analysic, the scram failure pyhcbility from failure of control god memanicmn is assumed to be 1.37 x 10 f or Oyster Creek; a chance of nhenit 1.4 ,
times in ten thoucand).
There are addj t f onal malfunctions which c<.n cause ceram f ailure other than f a t ture of the individual control rod insertion mechanicm considered in the foicnatug.
These vare considered in WASil-1400 and appear to be applicabic to the Oynter creek design. These malfunctions, which have to do with sensing and actuatton nyntems, are f hown in Figure C-1. 'ihe numerieni probabilitles are also gown, and .u) ove{all probability for scram failure has been conputed to be 1.4 x 10 (1.4 times in ten thousand).
e 0 e
+4+ 4 4
9
-"M 9