ML20070R248
| ML20070R248 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 01/25/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| MFN-011-83, MFN-11-83, NUDOCS 8301270360 | |
| Download: ML20070R248 (26) | |
Text
1
. 4 GENER AL $ ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CAUFORNIA 95125 MC 682, (408) 925-5040 January 25, 1983 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. D. G. Eisenhut, Director Division of Licensing Gentlemen:
SUBJECT:
IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT DOCKET NO. 50-447 REVISION OF CHAPTER 15 RADIOLOGICAL DOSE CALCULATIONS Attached please find a revision of the Chapter 15 radiological dose calculations for the Inside Containment Lost of Coolant Accident. Also included are the affected pages of Chapter 6.
These calulations have been revised to account for the capability of the suppression pool to retain particulate fission products. As demonstrated by General Electric's suppression pool scrubbing test program, and documented j in Section ISD. 2, the suppression pool provides an extremely effective fission product retention mechanism. General Electric personnel have been
, working with your staff during 1982 to facilitate the review of these test l results and their application to GESSAR's Chapter 15 analysis.
l This revision demonstrates the capability of the suppression pool to l significantly reduce offsite radiological doses for Design Basis Accident
! conditions. Following your preliminary review we will schedule the submittal j of an amendment to implement this revision.
! Very truly yours, l Glenn G. erwo , a a er Nuclear Safety and Licensing Operation Attachments cc: F.J. Miraglia, w/o att.
! D.C. Scaletti L.G. Hulman C.0. Thomas, w/c/ att.
L.S. Gifford, w/c/att.
8301270360 830125 PDR ADOCK 05000447 .
, A PDR
__ _ J
e CESSAR II 22A7007 I
, 238 NUCLEAR ISLAND Rov. 0 l
(
i 6.4.2.4 Interaction With other zones and Pressure-Containing Equipment (Continued)
'l outdoor air are mixed and drawn through filters, a cooling coil and sone electric reheating coils.
There a're two intakes, a normal intake located on the roof of the Control Building, and an alternate intake on the opposite side of the Nuclear Island at the end of the Auxiliary Building.
Radiation monitoring sensors located in each duct warn the operat-ing personnel (by means of readouts and alarms in the main control room) of the presence of airborne contamination. Also, the signal automatically closes down the normal air intake dampers and starts
, up the reduced flow (2000 cfm) air intake. This alternate ser-
[ vice air, which is classified as makeup air, is routed through the EEPA and charcoal filtering system for cleanup before be2.ng used for pressurization.
(- The control room must remain habitable during emergency condi-
) tions. In order to make this possible, potential sources of I
danger such as steam lines, pressure vecsels, CO fire fighting 2
containers, etc. are located outside of the control room and '
rekoved from the compartments containing Control Building life support systems.
l l
A tabulation of moving components in the Control Building HVAC System, along with the respective failure mode and effects, is shown in Table 6.4-1.
- i All dampers except the mixing dampers in the air condition'ing units are of the two position (open or closed) type.
l l
6.4-11
._,a- , , , , - , ~.n - - - . , , - - - . , , - - - , . , . , , - . , . ,
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. O o
s 6.4.2.5 Shie1dina Design j
6.4.2.5.1 Design Basis )
l l
The control Building shi~elding design is based upon adequately 1 protecting against the radiation resulting from ;; incid;;t ::
incidente i;; ding t; m3N e dose rates received under nor-mal operating conditions of the reactor are not determining factors in any of the walls sized in this specification. Under normal operating conditions, a dose rate of less than 1 mrem /hr is i
enticipated in the area surrounding the Control Building. Assum-ing an average gamma energy of 1.25 MeV and 2-ft shielding walls, this will yield a dose rate of less than 0.01 mrem /hr within the Control Building, which is well-within the acceptable limit.
Radioactivity released by an inadequate response to LOCA can
+
result in four different activity distributions, or sources, that can affect control Building personnel whole body doses.
)
(1) The fission products held in the containment " shine"
, an the Control Building. (Those remaining in the reactor vessel, however, contribute negligibly to this effect.)
(2) The fission products which are released from the SGTS stack form a cloud in which the Control Building is enveloped (Figure 6.4-5).
(3) Some of the fission products released from the contain-I ment will be taken into the Control Building via the building ventilation system air intake's. The majority of the iodine taken in will be absorbed on a charcoal bed, which will then become a concentrated source within the building. Also, solid daughters of noble gases s
)
6.4-12 J_ -_ _- -
GESSAR II 22A7007 233 NUCLEAR ISLAND Rev. 0 6.4.2.5.1 Design Basis (Continued)
[
collect on the filtars. Personnel on the control room level, as well as the equipment room and HVAC room levels, will be shielded from this source.
(4) Fission products that pass through and evolve from the filters become a source of radiation exposure to control building personnel. This source determines a portion of the whole body dose, as well as the entirety of the thyroid and beta skin doses. See Subsection 15.6.5 for these dose analyses.
The DBA analysis is structured on the conservative NRC assumptions.
Theff$Nbeg;; cf 105% ret;'. cr:r fission products;quilibri=
_2 . ... . u_ . _ _z__
... .. _ 3 f!*nK.E.REA M R VESSE'_'.TD DW CC'M.TA_'4..!1[ L..
. . released ___.. ._ _______ .___
r:102:e frer the centair ent are gifen helc.= ls PRESENTED IM
(. '
SCB scc.n oN 15.6.S E:1:0:2d frc=
n.....m_..,
nele::: frer Ficcic: Product Et::ter 'J::::1 Ocntain= nt F da m e- 100 4444
- alsgen: -4% -iH H C li': M gligible The containment leak rate assumed for the design analyses is 1.0%
of the containment volume per day. Radioactive decay during transport through the containment is taken into account. The leaked radioact' vity goes into the s..ctJDfS(u;M. seq ;e.m . C ;6WPtEA)T'lm and r ng annu then to the SGTS, from which it is vented to the atmosphere.
Mixing is assumed to occur in half of the shield building annulus fr'ee volume.
The SGTS charcoal filter is assum0d to be 99%
owc wb rm:ncuArE fcans c efficient for filtering 3eee64eiodine$, and none of the vented gas
~
. is assumed to bypass the filter.
6.4-13
.GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 6.4.i.5.2 Source Terms ;
wtTwELL AHO POTENGALLY AVhilAOLE containment3 sources " shining" on the Control Buildingf,are listed foR RELEASE in Table 15.6. b . Source terms for the cloud and filter are Eg R ME'MT consistent with the activity releases of Table 15.6 . Con-centration of each isotope is calculated as the product of the release rate (Ci/sec) times the appropriate relative concentra-tion,orX/Q(sec/sqtencemeteM). Th.. . 1.ec , den .ed fica.
":aclacz " x:: "lant C .. trol ".:: 'Jentil:ti n Cy:tc= 00 ign for :::: ting Con;::.1 02:ign Crit:ri:n 10", by ;;p; and ."urphy,
- prc;ented 5:lcu:
(X/C (;;;/- )
T1... "cesed LOCA in C :: nij .
mm t__ m om,,
o m. m mm,,
. ,t__
1 0 day; 0.0000 -
0 100 day; 0.0005 For the cloud source term, no credit was taken for decay between the release point and the cloud location. Buildup and decay of radiohalogens and solids on the filter was appropriately accounted for.
6.4.2.5.3 Results
- C 1culated ah:1 hedy d ::= t: centr:1 ::: personnel ::: gicen b low for n ;;;umed six month post--LOCA period. One hundred p;rc:nt :: up:ncy f :ter in ::: =:d.
Wf4LE BoDV GArttin DOSES CALCUt.ATED BY CMSERVArtVELY A S SJrfHUG l00% OctoPAnc.V FOR A WA - rt 0MTH PoGT- LOCA PERIOD ARC GivEN RELou FoR THE coarbrNr1ENTs CLoab AND l=tt rEk GH/NE CONTKtBurt on s, -
6.4-14 1
~"
c1usnsu . utwnwns
- 4 238 NUCLEAR ISLAND R3v. 0 6.4.2.5.3 Results (Continued)
.k.
Dose Iocation (Dose in Rem) s Component Control Room Equipment Room Containment Shine 0.0054 0.0063 Cloud 0.19 0.20 l
I' Charcoal Filter 0.0012 0.078 l
0.20 0.28 Ou VDY,Twkorb, Ave sgra DOGES RESQLTIAG f6btf hl AltBORNE A cTIwry wimIN ME CDN$O*
Roort ARE EVALUA7CO IN S&CTlCA l , $, AND ARC SM TD LrN tr$
qIQ C[Gbo DLcelt&MEo howokc_ ON__pt GEQM f,%,6C WELL InllTHIN YttEry,aof!py_M_?_C _
sencarisme. 6E:in si;T 7Fi'soAETWs T2kea kWop5dm GlGJrW (GW,C~~~
i..pli;; th; ;;;;ptati tp e. ..1 ejcupancy of these areas for the ocAario, 4
OFTNE post-LOCA period,fNcupancy will, however, have to be restricted somewhat in the chiller rooms at El (+)28'-6" during the first day post-LOCAc due to somewhat high dose rates (up to 0.5 Rem /hr attributable to radioactive gases in the intake duct. However, safety-related equipment redundancy obviates the need for full occupancy. After the first day, full occupancy in these areas would result in less than 5 Rem.
Concrete shielding thickness effecting the above doses are seen in plan and elevation in Figures 6.4-1 through 6.4-4. Penetrations and resultant streaming through the adjacent walls of the Auxiliary and Control Buildings are not considr.tred to be significant for the following reasons. The penetrations are all relatively small (cabletrayslessthan24in9pipeslessthan6in.)[ndnot radially aligned with the containment. Also, the overall con-tribution of the containment source component is quite small; thus, an increase due to straaming would not be a significant overall increase. Figure 6.4-6 shows a cross section of the Division 1 and 2 HVAC air intake. This is the only significant penetration of the Control Building external shielding. Examina-2 tion of this figure shows no credible streaming for the external cloud source. Finally, arrangement of the filter cubicle precludes t- .
6.4-15 P
v e
- - - ,. m. ,---. ,---.,.g.--m . - . , , - . -
-GESSAR II ^22A7007 238 NUCLEAR ISLAND R3v. 0 4
6.4.2.5.3 Results (Continued) )
4 door streaming. There are no floor penetrations of any consequence, cnd wall penetrations are kept well away from the source region. -
T 6.4.3 System Operational Procedures During normal operation, the control room group operates with mixed recirculated and fresh air, which pressurizes the subject spaces. Emergency conditions such as LOCA or high radiation l
cause automatic change over to reduced outside air and charcoal
[ filtering of all outside air to effectively isolate operating personnel from the environment and from airborne contamination.
Protection from direct radiation is discussed in Subsec-tion 6.4.2.5, isolation can be complete, even to food and water (Subsection 6.4.4).
Detection of radioactivity is instrumented, and changeover to reduced circulation and charcoal filtering is automatic. Redun-
)
( dancy of instrumentation and air handling systems ensures against cystem failure due to single component failure.
The above operational description is brief. For a more detailed description of normal and emergency operation of the control room habitability systems, see subsections 9.4.1, 9.5.1, 9.5.3, l 12.3.4, 6.5.1, 7.3.1.1.17, and Chapter 8.
6.4.4 Design Evaluations 6.4.4.1 Radiological Protection
.t Assumptions used in the generation of post-LOCA radiation source terms are described fully in Subsections 0.2.4.5 and 15.6.5.
6.4-16 l
I I
CESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 6 i I
l
('i 15.6.5.3.3 Results Results of this event are given in detail in Section 6.3. The temperature and pressure transients resulting as a coasequence of this accident are insufficient to cause perforation of the fuel cladding.
Therefore, no fuel damage results from this accident.
Post-accident tracking instrumentation and control is assured.
Continued long-term core cooling is demonstrated. Radiological input is minimized and within limits. Continued operator control cnd surveillance is examined and guaranteed.
,. 15.6.5.4 Barrier Performance The design basis for the containment is to maintain its integrity endano
- 1;;;;_
ExCUD L.E
- ELnC /kCLVTANCE CRorER A r:1 ricarre dter the instantaneous rupture of the largest single primary system 1.iping within the structure, while also accommodating the dynamic effects of the pipe break at the same time an SSE is also occurring. Therefore, any postulated LOCA does not reruit in exceeding the containment design limit (cee Sections 3.8.2.3, 3.6, and 6.2 for details and results of the analyses).
15.6.5.5 Radiological Consequences Two separate radiological analyses are provided for this cccident:
(1) The first is based on conservative assumptions con-i sidered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet 15.6-14
. GESSAR II.
22A7007 238 NUCLEAR ISLAND Rev. 0
(. 15.6.5.5 Radiological Consequences (Continued) 10CFR100 guidelines. This analysis is referred to as the " design basis analysis".
.(2 ) The second is based on assumptions considered to provide esoRE a3 realistic estimate of radiological consequences.
This analysis is referred to as the " realistic analysis",
At.THowf twat CONS $VATivE ASSOf1EDOMS S T H t- Edd%M.
A schematic of the transport pathway is shown in Figure 15.6-2.
Additional parameters and information for specific design basis -
j accidents are provided in Subsection 19.3.15.1. I a
15.6.5.5.1 Decign Basis Analysis 1 INSERT h ~Y :
D The n thed;, ;;;" ;ti n; =0 seeditie;; e ed te ;Tel st; thi;
- id;;t
- in
- ::: d=;; with th;;; ;;id:lin:: ;;t f;rth in i gul:t;;i 0;id;; 1.2- = d 1.7. "'h; p;;ifi; =f:1;, :::_ ;ti;;
(.
nd e r;;t;r ;;d; ;;;f t: ;;;l;;t thi; ;vnt b;;;0 ;; th; i:v;-
- it;;i; C
- p
- ::;tre in M f;;;;;; 2. Op;;ifi; 7;I;;; ;f per;;.
- t;;; =;d ir. thi; ;; .12; tin T.r; p;;;;;.ted ir. T;hl; 10. ', 7.
15 Fission Product Release from Fuel Insears .6.5.5.1.1 ,
6 t is assumed that 100% of the noble gases and 50% of the iodine are released from an equilibrium core operating at a power level of 3651 MWt for ,1.000 days prior to the accident. While not specifically stateEl in Regulatory Guide 1.3, the assumed release of 100% of the core noble gas activity and 50% of the iodine ,
activity implies fuel damage approaching melt conditions. Even though this condition is inconsistent with operation of the ECCS system (Section 6.3), it is assumed applicable for the evaluation l of this accident. Of this release, 100% of the noble gases and l 50% of the iodine.become airborne. The remaining 50% of the iodine is removed by plate-out and condensation, therefore, it is i
< not available for airborne release to the environment. .The
(
)
activity airborne in the containment is presented in Table 15.6-8 any _
15.6-15 1
GESSAR II 22A7007 238 NUCLEAR ISLPND Rev.0-6.5.5.1.1 Fission Product Transport to the Environment .
I The transport pathway consists of leakage from the containment to N the secondary containment-like structures by several different '
' mechanisms and discharge to the environment through the Standby
- Gas Treatment System (SGTS)
(1) Containment leakage - The design basis leak rate of the primary containment and its penetrations (excluding the ,
main steamlines) is 1.04/ day for the duration of the accident. All of this leakage is to the secondary con-tainment and from there to the environment via a 994 SGTS. Credit is taken for mixing and holdup within the secondary containment. The Shield Building exhaust rate, leakage rate, and mixing ratio are given on Tables 15.6-9 and 15.6-10.
4 (2) Leakage from engineered safety feature (ESP) components ,
outside primary containment.
() Hydrogen purge, - In the event of failure of the Hydrogen Recombiner System, purging of the containment may be necessary to control hydrogen concentration inside the primary containment. The earliest this purge may be utilizedisonehouraftertheaccidenthteof100scfm minimum. The purge would be processed by SGTS prior to release _ to the environment. SincE ThE HvoRoGE4 CONTRot EQuaPt1ENT ss ESF Eouufr1EMT, oT Is teor ASGur1EO TO FAIL 19) EVAt.OA rtAJG 771E PoTENTthL RhosOLOGoc.At E.tPOSURES ASSOCIATED WITH TWIS ACCIDENr, l
Fission product release to the environment based on the above assuniptiorsis given in Table 15.6-11.
15.6.5.5.1.3 Results
.45 WW6WM&Wb WM WM&WW 4W& M 5W FW w g eu ---ww ww ww m eaw wew aw ,
A 2 ad m er e e ~$ NA N wa h.7 I $ [d G h a^sh d b 16ePR&66.
NSER Q C .
15.6-16
i
.i -
} /usar A (p3e is.6-IO The methods, assumptions and conditions used to evaluate the potential radiological exposure to onsite and offsite personnel are consistent with the guidelines set forth in Regulatory Guide 1.3 except as noted below .
The
,from: guidance of Regulatory Guide 1.3 has been supplemented using in
- 1) NUREG-0772 to account for the expected chemical forms of the fission products; and 2) Section 15 D,2 to account for the retention of particulate fhsion products in the suppression pool. The specific models, assumptions and computer code used to evaluate this event based on the above criteria are presented in Reference 2. Specific values of parameters used in this evaluation are presented in Table 15.6-7.
1 l
l l
e -
- w. , ,_- * --- , ._ ..--, - - . . . . - - - - _ , , - _ - -
It is assumed that 100% of the noble gases and 50% of the iodine are j
releasedtothedrywellfromanequilibriumcoreoperaten$atapower level of 3651 MWt for 1000 days prior to the accident. These assumed releases imply that severely degraded ECCS performance has result,ed in fuel damage approaching melt conditions. Even though this condition is inconsistent with operation of the ECCS system (Section 6.3), it is assumed applicable for this evaluation of containment system effectiveness.
Any iodinc which is released from the reactor vessel would exist predominantly in a particulate chemical form. This analysis assumes that chemical forms of iodine released to the drywell are distributed in accordance with NUREG-0772, with'99.97% being in the particulate form and 0.03%
being organic iodine. Of the particulate iodine released, 50% is assumeo to be removed by plateout and condensation and therefore is not available for potential release to the environment.
\S,G.S.S.l,2 Ftssem PMooc.r TMMSPORT To THE &NmoWEtTT '
f The Mark III containment is designed with the drywell and suppression pool totally enclosed within the containment wetwell. In this configuration, any fission products released to the drywell, or discharged through the safety / relief valves, must pass through the suppression pool before they can reach the containment wetwell airspace. Once in the primary containment airspace, the transport pathway consists of limited leakage from the primary containment to the secondary containment by several different mechanisms and discharge to the environment through the Standby Gas Treatment System (SGTS). Consistent with the SGTS design capability and in accordance with R.G. 1.52 and the BWR/6 Standard Technical Specifications, ,
it is assumed that the SGTS has an iodine removal efficiency of 99%. +
l TheseTransportpathwaysandanyassociatedretentionmechanismsare identified below.
e 9
._ . ~
I 7 .
\
NSERT D COMTwOEb
- 1) Suppression Pool - All fission products discharged from ge drywell or safety / relief valves enter the suppression pool 7 to344 feet under the pool surface. As described in Section 15 D.2, General Electric has performed tests which demonstrate that the suppression pool will act as an extremely effective retention device for particulate fission products disc'harged into the pool under these conditions.
Based upon the testing and modeling documented in Section 15 D.2, a suppression pool scrubbing decontamination factor (DF) equal to 10,000 was applied to reduce the activity of particulate iodine reaching the primary containment airspace. All noble gases and organic foms of iodine were assumed to pass through the pool without retention (i.e., DF = 1). The resulting activity airborne in the containment wetwell airspace due to noble gases and iodine is presentedinTabia15.6-8.
- 2) Primary Containment Leakage - The design basis leak rate of the primary containment and its penetrations is 1.0%/ day for the duration of the accident. All of this leakage is to the secondary containment and from there to the environment via the SGTS. Parameters applicable Tt:A45PcRT Fret 1 to,the primary and secondary containments are given on Tablet 15.6-9 and 15.6-10, =d th: ::thit; cid;7.. in th: prinrj :st:i..;.ut h pin n ted in T:M e 10.0 +.
l l
- 3) Engineered Safety Feature Leakage - Engineered safety feature (ESF) components located outside the primary containment contribute insignificant leakage to the secondary containment.
e e
-ys - . - - -_ ,__.
= .
NSERTb( ;
)
The results of the' design basis analysis are presented in Table 15.6-12.
The calculated exposures at.the Exclusion Area and Low Population Zone are well within the guidelines of 10CFR100.
These _ calculated results are believed to be significantly greater than the maximum potential dose due to the conservative assumption $used in the design basis analysis. For example, the calculated exposuras would be significantly reduced if a more realistic time dependent release of fission products from the fuel were assumed. Further, other fission product retention mechanisms are present,in addition to the suppression pool,which would act to limit the release of fission products to the.
et.vironment. These additional retention mechanisms include agglomeration and settling of particulate forms plus surface deposition and absorption in the water vapor which would exist in the containment-air space
.following a LOCA. An additional retention mechanism exists as a result of the containment sprays.
1 A sensitivity study has been performed to determine to what extent these results.would change if the fuel release source term model recommended by NUREG-0772 were used instead of the Regulatory Guide 1.3 source term. If the NUREG-0772 model for a fully melted core were applied, the fuel I
release would change from 100% noble gas and 50% iodine to that specified in Table F.3-1 of Section 150.3. Since the noble gas release remains at a maximum value of 100% and the suppression pool and SGTS would reduce any increase in he articulate release by a factor of 10s, the effect on i
theoffsite[dNewouldbenegligible. Only the thyroid dose would be ,
calculated to change, with an increase from 0.05 rem to 01 rem in the Exclusion Area and 0.1. rem to 02. ree in the low population zone.
l.
. Thus, the results would still remain well within the guidelines of 10CFR100.
l l
i.
,- m
. . . \
GESSAR'II 22A7007 23P NUCLEAR ISLAND Rav. 0
, (, 15.6.5.5.2 Realistic Analysis The realistic analysis is based on a realistic but still conserva-lNSWtive assessment of this accident.
T.'.; :p;;ifi:
- f:12, 2 2 : c '- !
.iens.;;d thc. pre;;;= ::d f;r-: 1,;t;; ;cci;;ti;n e 0;;;;it;d D in f::: :: 2. f;;;ific V:1s:: cf ;;;- : :r; :::d in the-
....,...2_.
.n
_a. __e,-___
i 15.6.5.5.2.1 Fission Product Release from Fuel Since this accident does not result in any fuel damage, the only activity released to the drywell is that activity contained irs the reactor coolant plus any additional activity which may be released as a consequence of reactor scram and vessel depressurization.
r While there are vgfous activation and corrosion products contained
( in the reactor coolant, the products of primary importance are the iodine isotopes I-131 to I-li5. The coolant concentration for these isotopes is:
I-131 2. 0; ; 2.02. Ci/; . O.02. A C@m I-132 -2.0^ 2 1.%yCi/; .O.26 F.Ci/Sm
! I-133 1.:: : 1.0 A y ,= 0.15 M.Ci/Sm .
I-134 4.;; ;
1." Sci /;' O.4g n.C;/gr
^""
l I-135 . .
.. . __, ... O.lif a Ci /3 m l C nri'ering th t pprerirst:1y tOS cf th; ::12;::S liquid f12 hes t: :te =, it i: ::necrvatively ::::: d th:t 40t of the rele:::S Mbe6in ;;tivity 10 ri:5 rn; initi lly. " U:ver, :
re: ult Of plet ::t 2nd ::nd::::tien eff::te, 2:17 SOS cf the : tivity-
[
i initi:11y cirb rn rc=:in: :::ilchl: f:r : 1:::: te the
- nvir:n ::t.
3 15.6-17 -
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
q 15.6.5.5.2.1 Fission Product Release from Fuel (Continued) I
- f 1
As a :onsequence of reactor scram and depressurization, additional.
iodine. activity is released from those rods which expurienced - l cladding perforation riuring normal operation. Measurements performed (Referenceft) at operating BWRs'during reactor shutdown have been used to develop an analytical model for the prediction of iodine and noble gas spiking as a consequence of reactor scram and vessel depressurization. Based on the 95th per-centilet(i.e., only 5% of the time will the release be greater)
~
h obabilith the I-131 release is calculated to be 2.14 Ci/ bundle and Xe-133 to be 11.55 Ci/ bundle. Other iodine and noble gas isotopes are determined in accordance with their cumulative fission yields and are tabulated in Table 15.6-13.
Whsde$NCE 3no measurements have been obtained during a pressure transient as rapid as the LOCA, it is difficult tio predict the utual release rate from the fuel as a consequence of iodine -
)
spiking. Therefore, it-is arbitrarily assumed that 100% of the spiking source term 6e released Hhs esta To TH a.F ORYM...CLL n a n, SEFCRE tn; Tkl Bicapown p;;ica th:t TRANSdNTISCVER.
t00 u: th; di;;h;;;; ;;;1 = t i: ::=:hia- t: =t; .
pg. lon,at FRecnous INVOLVED IN THIS RftlAst ARE CONStSTENT WITH THcsc sPCciFIEO in NUREG 0772 be. 99.97% ta PPRnCULATE FORT 1 AND O.03% AS CRGMic tobrNE),
It is also assumed that plate-out and condensation removeI 50%
PARncutArs of the airborne 3 iodine activity. The total ;;;ivity cirt :n: in .
I l.- centai... eat i; p;;;;nt 0 in T;bi; 15.0 14.
l l
15.6.5.5.2.2 Fission Product Transport to the Environment INSERT M t
E. Th: 1::t ::t: c := e: pri : 2 ;;;t:i=;at ; tt; ;;;;ad;;1 i m attir : t i: 1_0t/dzy, t :: 1005 ciring i . :: r:f t w .
- .
- 1:::: f:: S: ::: -? y =nhirc t te 2; ::vircr::t vi :
00.00 i:dia: Offici::t ST i: p:::: t:d i T:ble 15.5-15. he int:g :t:d i::t:pi: ::tivity :01:22 d t: 2: ::vircr_ : t ir pr -
i se.. Led in T;il: 15.5--10. -
i 15.6-18 .
?
4 NSERT D p.lS.(w-I7)
The chemical forms of iodine used in this 1 realistic analysis.are con'sistent with those specified in NUREG-0772. As l 1
in the case of the design basis analysis, appropriate credit is given for :
suppression pool scrubbing. Specific values of the parameters used in the evaluation are presented in Table 15.6-7. The specific models, assumptions and the program used for computer evaluation are describcd in Reference 2. .
i l+
i i
I
-l I
NSERr E ( p'lS.6-lS')
The activity released in this analysis is directed to the drywell and from there into the soppression p ol. Any fission products not retained in the suppression pool due to scrubbing are assumed to be distributed in the wetwell airspace. This t.ansport process is discussed in detail in Section 15.6.5.5.1.. As in the case of the design basis analysis, credit was taken for a decontamination factor of 10,000 for particulate iodine due to the suppression pool, The activity airborne in the containment wetwell airspace is presented in Table 15.6-14.
Parameters applicable to the primary and secondary containmentSare given in Table 15.6-10. The integrated isotopic activity release to the environment via the SGTS is presented in Table 15.6-15. The SGTS has a 99%foYl"nhromoval efficiencyAno A 99.7% PAcrtC.ddlTE tooiNE Rtrie<AL EFFtCsLHCf, i
l t
j O
b 9
I
, . GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 0 15.6.5.5.2.3 Results !
As swa m Tasa: 15.6-16, 3 The calculated radiological exposures for this event are 7 :::rt M 1.. M ie 10.0 10 c.0 :: M= ::: a small fraction of 10CFR100.
15.5.5.5.3 control Room lNSEgr ._+
F A dose analysis has been performed to demonstrate that the ventilation system satisfies the NRC radiation guidelines. The
, results of the analysis show that the ventilation system design does satisfy their guideline. A schematic of the control room intake vents is shown in Figure 15.6-3.
The doses received during a 30-day period after a loss-of-coolant accident are:
Dose U.S. NRC Limit
Whole Body 2.56 5 Thyroid 29.4 30 Beta 53.8 75 A factor of 1/4 was taken into account for a dual inlet with manual override capabilities. The methods used to calculate these doses are presented in Reference 5. A complete list of assumptions and input data follows:
1 (l' Source Terms The source terms used in this analysis are consistent with R.G. 1.3 (i.e., 254 halogens and 100% noble gases airborne in the containment) and were presented in Table 15.6-8.
15.6-19
GESSAR II 22A7007 238 NUCLEAR ISIAND Rev. 0 15.6.5.5.3 Control Room (Continued) i
~
hDDITIONAL ASSut1PIIONS ok) VEhlTILADOM j% par 1ETERS AND VtETEcROLOGscAL DAEh ARE GIVEN BELOW Vent 11atton Parameter (Sgg fgcogg (5,f-] fog VENTIMTION lNTAKE
'SCt1ENATIch Inlet air flowp filtered 0.944 m3 /sec unfiltered 0.0014 m3 /sec INLErFilter NNeiency 981 Control Room Volume 1.102 E+4 m3 Occupancy factors 0-2 hrs 1.0 2-8 hrs 1.0 8-24 hrs 1.0 1-4 days 0.6 4-30 day 5 0.4
% Meteorology Data
-)
3 X/Q Values sec/m 0.-2 hri 8.0 E-3 2-8 hrs 3,2 M E-3 8-24 hr5 2.9 M E-3 '
24 hrP4 days 2.2 M E-3 4-30 days 2.2 M E-3 15.6.6 Feedwater Line Break - Outside Containment i
i In order to evaluate large liquid process line pipe breaks outside containment, the' failure of a feedwater line is assumed to evalu-ate the response of the plant design to this postulated event.
~
The postulated break of the feedwater line, representing the largest liquid line outside the containment, provides the. envelope 1
15.6-20 .
- ~ , - - , ~~ ,-.,-r,,, , , . - ,
~
INSERT F ( a c 15.6- @
t1 ave The control room and its associated ventilation system hes been3 designed with the objective of continuous occupancy following a LOCA. An analysis
! has therefore been performed to demonstrate that the ventilstion system satisfies the NRC's control room habitability guidclines relative to radiation exposure. The potential doses to control room personnel during a 30-day period after a L0tA are shown below. These are based on: 1) the source terms, iodine fractions and scrubbing discussed in Section 15.6.5.5.1 ; 2) a factor of 0.25 to take credit for a dual inlet with manual override capabilities; and 3) the calculation methed: presented in Reference 4.
Dere U.S. NRC Limit M (Rem) ,
~
Whole Body 2.94 5 Thyroid O.02 30 Beta 36 75 x A St1 As s- INCRErtEs;T To rite (CHouE Sch'Y MciE REGu'ur$ WM GPmA SccRCES kTERNAL. Ib TttE CcN % Room. TTits efin cr is Ev^tunrec
,a Secnos 6.%2.5.
~ .
CESSAR II 22A7007 238 Nt. ; LEAR ISIAND A;v. '6
'15.6.7 References
]
- 1. r. J. Moody, " Maximum Two-Phase Vessel Blowdown from Pipes",
ASME Paper Number 65-WA/HT-1, March 15, 1965.
l
- 2. -r. 7. Ot;nc;;;;; :nd . J. M: ;:n, "Centerv:tiv: "-d -
1;;ical ?.ccid:nt I';;1 :tica --Th: COF?.C^1 0 d;", M ::h 1975 !
'M200-211f2).
-3 . O,. M;;y:n, "" .
licti: ?.ccid nt ?.n:lvcic The n 1.AC Cede",
Octcher l?" 'F " ^-21112).
K.3 r.a.krueschy,c.R. nills,N. R. norton, A. a. Levine,
" Behavior of Iodine in Reactor Water During Plant Shutdown and Startup", August 1972 (NEDO-10585).
~
- 5. L. G. ";eic; and V. O. 1;guycn, "Ocatrol nee 7. Accid:nt l' "'
Exp;;; : Ev:12 tion C"0C l' :g ;r.", Janu;ry 1070
-fMED^ 2 2 0 0 0 } '
- 2 D. Nauvm, et al, "RaorotocacAt Acc,osar Evntonnon~ Tne CONAC 03 Cooe" Decer*see I9et (NECO~2If3-)),
t
# 4 D. Nouven, etd "Conrest Rose Accioem Ex PoSORE EVAWAnod-CRDoS' PROGR^n: FE6k"**Y IR8I NN'D0'13909 Ab 9
t l 15.6-27/15.6-28 1
Y
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 4 (% Table 15.6-6 STEAMLINE BREAK ACCIDENT (REALISTIC ANALYSIS) RALTOLOGICAL EFFECTS Whole Body Dose Inhalation Dose (rem) (rem) ~ Exclusion Area 1.lE-2 5.2E-1 4 Low Population Zone 5.4E-3 2.6E-1 ed .i i
, s. * * '
- t r, 0 .7 - '
( . 15.6-35 i
~
- ~ ~
GESSAR II 22A7007 238 NUCLEAK ISLAND Rev. 0 4 Table 15.6-7 ,I LOSS-OF-COOLANT ACCIDENT - PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSES-Design Realistic Basis Basis Assumptions Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level 3651 MWt 3651 MWt B. Burnup NA NA C. Fuel damage 1004 0 D. Release of activity by nuclide icoL MentO.s too/,r AlesuEC% 50% Teo,as. So?'c feh,~E
'E. Iodine fractions (1) Organic y 0.0003 X O.Occ3 (2) Elemental 20 EO (3) Particulate 50.9997 X O.9997 F. Reactor coolant activity NA 15.6.5.5.2.1 before the accident II. Data and assumptions used to
( estimate activity released s A. Primary containment leak rate (^./J.ej ) M6d l$.6-10 1 re-TABLE J5.6-/O 1,4-
)
B. E6condary containment leak 0.2 hre 3?.0.2 123 rate (t/ dei) TABd 15.6-10 TARd 15.6-10 C. :10 hr; 43.0 122 O K. Valve movement times NA NA F E. Adsorption.and filtration ! efficiencies (t) l (1) ' Organic iodine -NA 99 NA 99 (2) Elemental iodine 99 NA 99A MA (3) Particulate iodine 43h 99 NA 99,3 (4) Particulate fission products 43h 99 43A-97.9 F g. Recirculation system parameters (1) Flow rate (CFM) 5000 5000 (2) Mixing efficiency 50 100 (3) Filter efficiency NA NA G f. Containment spray parameters (flow rate, drop size, etc.) NA NA H 4. Containment volumes NA NA { jf. All other pertinent data , and assumptions None None s
}
C, S'0PPRE55(otd cot.
% RUBBING DEcoNmanca VAcnat (1) ORGAN IC. IO DI^E 15.6-36 l I I l0,000 (2) PARTICCL6TE IODn0E IOiCOO
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 1, . Table 15.6-7 (Continued) Design Realistic Basis Basis Assumptions Assumptions TII. Dispersion Rate A. Boundary and LPZ distance (m) *
- B. X/Q's for-time intervals of (1).0-2 hr - SB/L's e
- 2. 0E-3/1. 0E-3 2. 0E-3/1. 0E-3 1.0E-3 1.0E-3 (2) 2-8 hr - LPZ 3.8E-4 3.8E-4 (3) 8-24 hr - LPZ l.0E-4 1.0E-4 (4) 1-4 days - LPZ (5) 4-30 days - LPZ 3.42-5 3.4E-5 7.5E-6 7.5E-6 IV. Dose Data A. Method of dose calculktion ' Reference 2 Reference J 2 B. Dose conversion assunptions Reference 2 Reference gZ
.C. Peak activity concentrations Table Table in containmentWcrwcc A#RSPACE 15.6-8 15.6-14 D. Doses Table Table 15.6-12 15.6-16
- Applicant to Supply .
15.6-37 !
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?5 N M LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS). X (, % ,
l' ACTIVITY AIRBORNE IN PRIMAAY CONTAINMENT (Ci) h -N l- N i Isotope 1 min 30 min 1 hr 2 hr 4 hr 8 hr 12 hr 1 day 4 dayL 30 day Il31 '2.1E 07 2.lE 07 2.lE 07 '2.lE 07 2.lE 07 2.lE 07 2.OE 07 .1.9E 07 'l.5E 07 1.2E 06 1132 3.5E 07 3.0E 07 2.6E 07 1.9E 07 1.0E 07 3.lE 06 9.2E 05 2.4E 04 7.3E-06 ' O. l Il33 3.3E 07 3.2E 07 3.lE 07 3.0E'07 2.8E.07 2.5E 07 2.2E 07 1.4E 07 1.3E 06 9.3E-04 1134 5.5E 07 3.7E 07 2.5E 07 1.lE 07 2.3E 06 9.8E 04 4.lE 03 3.0E-01 O. O. 1135 4.6E 07 4.3E 07 4.lE 07 3.7E 07 3.0E 07 2.0E 07 1.3E 07 3.6E 06 1.8E-03 0~ l w Total I 1.9e 08 1.6E 08 1.4E 08 1.2E 08 9.2E 07 6.8E 07 5.6E 07 3.7E 07 1.6E 07 1.2E 06 $ Kr83m f .5E 0'i 8.OE 06 6.6E 06 4.5E 06 2.lE 06 4.8E 05 1.lE 05 1.2E 03 2.2d-09 0.- Eo i gl Kr85m 2.3E 07 2.0E 07 1.7E 07 1.2E 07 6.6E 06 3.6E 06 5.5E 05. 0M e 2.lE 07 7.6E 00 O. .Q$ 3 6 Er85 5.9E 05 5.9E 05 5.92.05 -5.9E 05 5.9E 05 5.9E 05 5.8E 05 5.8E 05 5.6E 05' 4.3E 05 $$ Er8'P 4.7E 07 3.6E 07 .2.7E 07 1.6E 07 5.3E 06 5.9E 05 6.6E 04 '9.2E 01 6.5E-16 0. (( , Kr88 6.7E 07 5.9E 07 5.2E 07 4.1E 07 2.5E 07- 9.2E 06 3.4E 06 1.7E 05 2.9E-03 0. N 2 Kr89 6.7E 07 1.2E 05 1.6E 02 3.OE-04 1.lE-15 O. O. O. O. O.- O Xe131m 5.7E 05 5.7E 05 5.7E 05 5.7E 05 5.7E 05 5.6E 05 5.5E 05 5.4E 05 4.4E 05 7.5E 04 ' Xel33m 2.3E 07 2.3E 07 2.3E 07 2.2E 07 2.2E 07 2.lE 07 2.0E 07 1.7E 07 6.4E 06 1.5E 03 ' Lvl33 1.3E 08 1.3E 08 1.3E 08 1.3E 08 1.3E 08 1.3E 08 1.2E 08 1.2E 08 7.6E 07 1.9E 06-
~'
Xe135m 3. tie 07 9.8E 06 2.5E 06 1.7E 05 7.2E O2 1.4E-02 2.6E-07 0. O. O.
- Xel35 2.4E 07 2.3E 07 2.3E 07 2.lE 07 1.8E 07 1.3E 07 9.8E 06 3.9E 06 1.6E 04 0. i Xe *,. 'l7 1.5E 08 7.8E 05 3.4E 03 6.8E-02 2.6E-11 0. O. O. O. O.
Xel38 1.6E 08 3.9E 07 9.0E 06 4.8E 05 1.4E 03 1.lE-02 8.9E-08 0. O. O. w 7.4E 08 3.5E 08- 3.0E 08 2.5E 08 2.2E 08 1.8E 08 1.6E 08 55 Tctal NG 1.4E 08 -8.3E 07 2.4E O fy
- o8.
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GESSAR II 22A7000 238 NUCLEAR ISIAND Rev. O Tr$9t96 I
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GESSAR II 22A7007-238 NUCLEAR ICIAND R;v. 0 (v 2 Table 15.6-9 SliIELD BUILDING EXHAUST RATE t Time Average Exhaust Flow Rate to SGTS (hr)~ - (SCFM) lj 0-2 480 2 - 10 90
>10 66 Table 15.6-10 LEAKAGE RAT 5'S AND MIXING RATIO Numerical Value Parameter Design Basis Realistic
{- A. Primary to Secondary Containment (t/ day) 0 - 2 hr 0.832 1re 0.932 2 - 10 hr .' 0.903 -i reO.903
>10 hr O.908 1re o.909 B. Primary Containment Leakage to SGTS (t/ day) 0 - 2 hr 0.168 ** 0.169 2 - 10 hr 0.097
>10 hr NA0.097 0.092 4MrO.092 C. Secondary Containment Leakage to SGTS (t/ day) 0 - 2 hr -
2 - 10 hr 319.3 i+9 IN b
>10 hr 59.9 i+3 29.9 43.6 149- 22.0 D., Mixing Efficiency (%)
i Primary Containment 100 100 Shield Building Annulus 50 100 15.6-39
1
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Table 15.6-11 bm P t _L,. g b LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS) ACTIVITY RELEASE TO ENVIRONMENT (Ci) gy ( Isotope 1 min 30 min R ;2 1 hr 2 hr 4 hr 8 hr 12 hr 1 day 4 day 30 day ' Il31 2.5E-Ol 8.7E 00 2.0E 01 4.8E 01 7.5E 01 1.5E 02 2.4E 02 5.7E 02 3.9E 03 1.8E 04 i Il32 I 4.lE-01 1.3E 01 2.8E 01 5.7E.01 7.5E 01 9.6E 01 1.0E 02 1.1E 02 1.lE 02 1.lE 02 I133 3.8E-01 1.3E 01 2.9E 01 7.0E 01 1.lE 02 2.OE 02 3.0E 02 6.0E 02 1.6E 03 1.7E 03 I134 6.4E-01 1.8E 01 3.4E 01 5.7E 01 6.4E 01 6.6E 01 6.6E 01 6.6E 01 6.6E 01 6.6E 01 I135 5.4E-01 1.8E 01 4.0E 01 9.lE 01 1.3E 02 2.2E 02 2.9E 02 4.lE 02 4.8E 02 4.8E 02 w w Total I " 2.2E 00 7.lE 01 1.5E 02 3.2E 02 4.6E 02 7.4E 02 1.0E 03 1.8E 03 6.2E 03 2.0E 04 [ Kr83m 1.lE 01 3.5E 02 7.3E 02 1.4E 03 1.9E 03 2.2E 03 2.3E 03 2.4E 03 2.4E 03 2.4E 03 Eo m Kr85m 0g 2.7E 01 8.9E,02 1.9E 03 4.3E 03 6.2E 03 9.4E 03 1.2E 04 1.4E 04 1.5E 04 1.5E 04
$ Kr85 6.9E-01 2.4E 01 5.4E 01 1.3E 02 2.lE 02 4.2E 02 6.7E 02 1.7E 03 1.3E 04 1.4E 05 gg Kr87 ""
5.5E 01 1.7E 03 -3.3E 03 6.0E 03 7.2E 03 7.9E 03 8.OE 03 8.lE 03 8.lE 03 8.1E 03 Kr88 7.8E 01 2.5E 03 5.4E 03 1.lE 04 1.6E 04 .2.lE 04 2.4E 04 2.5E 04 2.5E 04 2.5E 04 E O Kr39 8.8E 01 4.7E 02 4.7E 02 4.7E 02 4.7E 02 4.7E'02 4.7E 02 4.7E 02 4.7E 02 4.7E 02 Xel31m 6.7E-01 2.3E 01 5.3E 01 1.3E 02 2.0E 02 4.0E 02 6.5E 02 1.6E 03 1.lE 04 6.4E 04 . Xe133m 2.7E 01 9.3E 02 2.lE 03 5.lE 03 7.9E 03 1.6E 04 2.4E 04 5.58 04 2.6E 05 4.5E 05 Xel33 1.6E 02 5.4E 03 1.2E 04 3.0E 04 4.7E 04 9.3E 04 1.5E 05 3.5E 05 2.2E 06 7.3E 06 . i Xel35m l l 4.4E 01 8.2E 02 1.lE 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 Xe135 l 2.8E 01 9.7E 02 2.1E 03 5.0E 03 7.5E 03 1.3E 04 1.8E 04 2.8E 04 4.0E 04 4.OE 04 ' Xel37* 1.32 02 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 Xel38 u 1.9E 02 3.5E 03 4.5E 03 4.8E 03 4.9E 03 4.9E 03 4.9E 03 4.9E 03 4.9E 03 4.9E 03 gg Total NG << 9.0E 02 1.9E 04 3.5E 04 7.1E 04 1.0E 05 1.7E 05 2.4E 05 4.9E 05 2.6E 05 8.lE 06
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GESSAR II . 22A7007 238 NUCLEAR ISLAND Rev. 4
. Table 15.6-12 IDSS-OF-COOLANT ACCIDENT (DESIGN BASE ANALYSIS) I i RADIOLOGICAL K*FECTS -
I whole Body Dose Inhalation Dose (rem) (ren) Xxclusion Area 19./l 46r90.OS l Low Population Ione 14.1 64re 0,ll
. ~
I Table 2 5.6-13 ISOTOPIC SPIKING ACTIVITY The 95th Cumulative Probability Spiking Isotope Name Activity (Ci/ bundle) I131 2.14 1132 3.21 I133 5.03 I134 5.44 1135 4.79 Kr83m 9.04-l* Kr85m 2.23+0 Kr85 4.90-1 Kr87 4.33+0 b8 6.12+0 Kr89 7.96+0 Xel31m 6.60-2 Xe133m 3.26-1 Xel33 1.16+1 Xel35m 1.80+0 Xel35 1.10+1 Xel37 1.05+1 Xel38 1.06+1 -
-*9.04-1 = 9.04 x 10-1 -
15.6-41
. _ _ . - - - tn [6 TJ Table 15.6-14 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS) bE$
e r ACTIVITY AIRBORNE IN THE CONTAINMENT (Ci) q 'D Q 3> Q
- R E. -
Isotope 1 min 1 hr 2 hr ' 8 hr 1 day 3 day 26 day 3131 8.59E 01 8.55E 01 8.52E 01 8.332 01 7.84E 01 5.96E 01 5.64E 00 1132 1.41E 02 1.05E 02 7.73E 01 1.25E 01 9.69E-02 2.99E-11 f . 0. 1133 2.08E 02 2.01E 02 1.94E 02 1.59E 02 9.28E 01 8226E 00 0.91E-09 1134 2.39E 02 1.10E 02 4.97E 01 4.29E-01 1.34E-04 0 O. 1135 2.03E 02 1.83E 02- 1.65E 02 8.75E 01 1.62E 01 8.167.-03 0. 8.77E 02 6.84E 02 5.71E 02 3.43E 02 Total 1.88E 02 6.79E 01 5.64E 00 $ Er83m 6.72E 02 4.65E 02 3.20E 02 3.37E 01 0.38E-02 1.53E-13 l
!
- Er85m 1.67E 03 1.43E 03 1.23E 03 4.83E 02 4.03E s01 5.55E-04 0.
0. 88 Er85 3.67E 02 3.66E 02 3.66E 02 3.65E 02 3.63E 02 3.52E 02 . 2.73E 02 h Er87 3.21E 03 1.87E 03 1.08E 03 4.05E 01 6.35E-03 0 O. I Er88 4.56E 03 3.57E 03 2.79E 03 6.29E 02 1.19E 01 2.02E-07 0. $U Erst 4.78E 03 1.13E-02 2.14E-08 0. O. O. O. , Xe131m 4.94E 01 4.92E 01 4.31E 01 4.83E 01 4.61E 01 3.77E 01 5.53E 00 g Xe133m 2.44E 02 2.41E 02 2.37E 02 2.19E 02 1.77E 02 6.75E 01 1.63E-02 l Me133 8.64E 03 8.59E 03 0.54E 03 8.2'4E 03 7.50E 03 4.91E 03 1.26E 02 Xe135m, 1.20E 03 8.87E 01 5.85E 00 4.83E-07 0. O. O. Xe135 8.19E 03 7.60E 03 7.04E 03 4.46E 03 1.32E 03 5.51E 00 0. Ee137 6.54E 03 1.54E-01 3.03E-06 0. O. O. O. Xe138 7.58E 03 4.25E 02 7.27E 01 5.23E-07 0. O. O.
- Total 4.78E 04 2.47E 04 2.17E 04 1.45E 04 9.46E,03 5.37E 03 4.06E 02
/ u i
' so u
/ O>
's *
. O 4
- w. .. .
Table 15.6-14 ,, f .- LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS) i
- ACTIVITY AIRBORNE IN THE CONTAINNENT3(Ci)
WETWELL . I ISOTOPE 1-MIN 10-MIN 1-H0llR 2-HullR 4-HOUR 8-HOUR 12-HR 1-DAY 4-DAY 30-BAY i' ~ I-131 5.63E-01 5.63E-01 5.61E-01 5.59E-01. 5.55E-01 5.46E-01 5.37E-01 5.12E-01 3.83E-01 3.14E-02 I-132 8.60E-01 8.22E-01 6.39E-01 4.73E-01 2.59E-01 7.73E-02 2.31E-02 6.19E-04 2.27E-13 1 00E-20 - l I-133 1.33E 00 1.32E 00 1.29E 00 1.24E 00 1.16E 00 1.02E 00 8.87E-01 5.92E-01 5.21E-02 3.74E-11 1 I-134 1.45E 00 1.29E 00 6.6'iE-01 3.01E-01 6.20E-02 2.62E-03 1.11E-04 8.35E-09 1.00E-20 1.00E-20
' 1.02E-01 5.21E-051.00E-20 I-135 1.28E 00 1.26E 00 1.15E 00 1.03F 00 8.3PE-01 5.50E-01 3.61E-01 w TOTAL' I 5.48E 00 M.2SE 00 4,30E 00 3.41E 00 2.8RE 00 2.19E 00 1.81E 00 1.21E 00 4.36E-01 3.14E-02
,. - R.
I I KR-83M 6.72E 02 6.35E 02 4.63E 02 3.17E 02 1.4BE 02 3.26E 01 7.15E 00 7.55E-02 1.05E-13 1.00E-20 h
!* KR-85M 1.67E 03 1.63E 03 1.43E 0.1 1.22E 03 8.98E 02 4.83E 02 2.60E 02 4.03E 01 5.68E-04 1.00E-20 l KR-85 3.67E 02 3.67E 02 3.67E 02 3.67E 02 3.66E 02 3.66E 02 3.65E 02 3.63E 02 3.52E 02 2.70E 02 "U
,4 KR-87 3.21E 03 2.96E 03 1.8PE 04 1.09E 03 3.66E 02 4.13E 01 4.65E 00 6.69E-03 1.00E-20 1 00E-20
.N KR-88 4.56E 03 4.40E 03 3.59E 03 2.81E 03 1.72E 03 6.48E 02 2.44E 02 * ,.30E 01 2.93E-07 1.00E-20
! KR-89 4.78E 03 6.68E 02 1.19E-02 2 40E-08 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1.00E-20 XE131M 4.94E 01 4.94E 0. 4.93E 01 8.91E 01 4.88E 01 4.83E 01 4.77E 01 4.61E 01 3.76E 01 6.38E 00 XE133M 2.44E 02 2.43E 02 2.41E 02 2.37E 02 2.31E 02 2.19E 02 2.07E 02 1.76E 02 6.60E 01 1.35E-02 XE-133 8.64E Ob 8.63E 03 8.59E 03 8.54E 03 8.44E 03 8.24F 03 8.05F 03 7.49E 03 4.89E 03 !221E 02 XE135M 1.28E 03 8.A0E 02 9.39E 01 6.58E 00 3.23E-02 7.R1E-07 1.88E-11 1.00E-20 1.00E-20 1.00E-20 '
; XE-135 8.19E 03 8.10E 03 7.59E 03 7.03E 03 6.03E 03 4.44E 03 3.27E 03 1.30E 03 5.18E 00 1.00E-20 XE-137 6.54E 03 1.28E 03 1.51E-01 2.SiE-06 1.0EE-15 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1.00E-20 XE-138 7.58E 03 4.88E 03 4.23E 02 2.24E 01 6.33E-02 5.03E-07 4.00E-12 1.00E-20 1.00E-20 1.00E-20 TOTAL NG 4.78E 04 3.47E 04 2.47E 04 2.17E 04 1.83E 04 1.45E 04 1.24F 04 9.43E 03 5.35E 03 3 onE 02 u
.f 5
.8
e m. i
.m p 5D-mh,
" n, Table 15.6-15 bqh LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY RELEASED TO THE ENVIRONMENT (Ci) r-
'h i Isotope 1 min E
1 hr 2 hrs 8 hrs 1 day 4 days 30 days k I131 2.54E-10 8.98E-07 3.52E-06 5.04E-05 3.44E-04 2.24E-03 8.20E-03 I , Il32 4.19E-10 1.22E-06 3.95E-06 2.0?E-05 2.80E-05 2.81E-05 '2.81E-05 i
- I133 6.?.4E-10 2.13E-06 8.152-06 1.04E-04 5.39E-04 1.46E-03 1.56E-03 '
I134 7.12E-10 1.53E-06 3.76E-06 7.66E-06 7.74E-06 7.74E-06 ! 7.74E-06 Il35 6.01E-10 1.99E-06 7.30E-06 7.15E-05 2.2E-04 2.63E-04 2.63E-04 { Total 2.60E-09 7.76E-06 2.67E-05 2.55E-04 1.13E-03 4.00E-03 1.01E-02 l l g Kr83m- 1.99E-06 5.54E-03 1.72E-02 7.48E-02 9.03E-02 9.04E-02 9.04E-02 Eo i Kr85m 4.94E-06 1.58E-02 5.62E-02 4.64E-01 1.03E 00 nn 1.11E 00 1.llE 00 m l 5 Kr85 1.08E-07 3.84E-03 1.51E-02 2.19E-01 1.55E 00 1.15E 01 9.29E 01 ' Kr87 9.53E-06 2.38E-02 6.68E-02 1.99E-01 2.llE-01 2.llE-01 2.11E-01 g[ Kr88 1.35E-05 4.07E-02 1.37E-01 Kr89 1.52E-05 7.28E-04 7.28E-04 8.40E-01 7.28E-04 1.30E 00 1.32E 00 1.32E 00 {j 7.28E-04 7.28E-04 7.28E-04 0 Xel31m 1.46E-07 5.16E-04 2.03E-03 2.91E-02 2. ole-01 1.36E 00 6. ole 00 l Xel33m 7.22E-07 2.53E-03 9.88E-03 1.36E-01 8.48E-01 3.95E 00 6.07E 00
$ Xel33 2.56E-05 9.02E-02 3.53E-01 5. ole 00 3.36E 01 2.03E 02 5.44E 02 1 i l Xe135m 3.86E-06 2.89E-03 3.70E-03 3.81E 3.81E-03 3.81E-03 3.81E-03 Xel35 2.42E-05 8.17E-02 3.06E-01 3.33E 00 1.21E 01 1.81E 01
' 1.81E 01 ; Xel37 2.06E-05 1.42E-03 1.42E-03 1.42E-03 1.42E-03 1.42E-03 1.42E-03 I Xel38 2.28E-05 1.54E-02 1.90E-02 'l.94E-02 1.94E-02 1.94E-02 '1.94E-02
\ m Total 1.44E-04 2.85E-01 9.88E-01 1.03E 01 5.10E 01 6.70E 02 EN 2.40E 02 ,g, i
, s ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0-l
~
Table 15.6-16 :
~
1 LOSS-OF-COOLAlfr ACCIDENT (REALISTIC ANALYSIS) RADIOLOGICAL EFFECTS Whole Body Dose Inhalation Dose (rem) (rem)
- 1. 2E -3 5.y:1. 0E-4 Exclusion Irea 2.2:; 4 9.7E-+ 2.9E-G 0.0 Low Population Zone 3. ;;;;-4 5 Table 15.6-17 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK OUTSIDE CONTAINMENT Time Event O sec One feedwater line breaks.
0+ sec Feedwater line check valves isolate the reactor from the break.
<30 sec At low-low water reactor level RCIC would initiate, HPCS would initiate, MSLIV closure would initiate, reactor scram would initiate and recirculation pumps
.would trip. ;
%2 min The safety / relief valves would open and close and main-tain the reactor vessel pressure at approximately 1100 psig.
1-2 hr Normal reactor cooldown prc;3 dure established.
- Applicant to Supply 15.6-44
~
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a GESSAk II 22A7007 238 NUCLEAR ISLAND R3v. O SECONDARY g CONTr *NMENT / j i 4 SHlELD SulLDING 40B0 CPM MAX 4000 CPM MIN , j g TO SHIELD BUILDING RECIRC , 5000 W M l FAN FROM SH!!LD SulLDING T 1000 CFM MAX PRIMARY CONTAINMENT
- 360 CFM MIN 7
ggg %. AUXILIARY E000 CFM 1000 CFM BUILDING ' MAX FROM FUEL 1650 CFM BUILDING MON 3000 CFM MAX O J L 200 CFM MIN FROM AB M N I-INSOARD OUTBOARD STOP MSIV MSIV VALVE RPV
. \
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t SMEAK d
- CONTAINMENT LEAKAGE (1.0E/ DAY)
TO FE 43NO CFM MAX
. _ _ . . ', 300 CFM MIN
~'._ __
FROM ECCS j AJtEA SUPPRESSION POOL i f Figure 15.6-2. Post-LOCA Fiss.0a
';;k;;;fbouct Rc cancaSysrENS g ";thr:y;'
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15.6-50
% ^
. . i TURSINE TUnetwE STOP REACTOR VESSEL SUILDING VALVE MAIN STEAM LINE TURetfE FLOW AP ISOLATION VALVES ""
VALVE w P., j 4 N^ 1 M S X X X X >4-J5 x, m
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FLOW LIMITER FEEDWATER CHECK VALVES MA. - CONTAINMENT FEEDWATER . MOLATION VALVE SYPASS LINE r "A" ' CONDENBER u FEEDWATER PUMP - E
.a Figure 15.6-1.
. ou Steam Flow Schematic for Steam Break Outside Containment s
_ _ _ . _ _ _ _ _ _ _ _ _}}