ML20071E274

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Documents Agreements on Credit for Suppression Pool Scrubbing in GESSAR-II,Chapter 15 Re Radiological Dose Calculations.Potential Decontamination Capability Provided by Scrubbing Will Be Recognized
ML20071E274
Person / Time
Site: 05000447
Issue date: 03/09/1983
From: Sherwood G
GENERAL ELECTRIC CO.
To: Eisenhut D
Office of Nuclear Reactor Regulation
References
JFN-7-83, JNF-007-83, JNF-7-83, MFN-019-83, MFN-19-83, NUDOCS 8303140179
Download: ML20071E274 (2)
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Text

'

.T GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 f1C682,(408)925-5040 MFN-019-83 JNF-007-83 March 9, 1983 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C.

20555 Attention:

Mr. D. G. Eisenhut, Director Division of Licensing

SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II) DOCKET NO. 50-447 CHAPTER 15 RADIOLOGICAL DOSE CALCULATIONS WITH POOL SCRUBBING CREDIT

Reference:

Glenn G. Sherwood to D. G. Eisenhut, " Revision of Chapter 15 Padiological Dose Calculations,"

i January 25, 1983.

The purpose of this letter is to document agreements made with the Staff regarding credit for suppression pool scrubbing in Chapter 15 radiological dose calculations.

On January 13, 1983, GE personnel met with Mr. L. G. Hulman and other members of the Staff to discuss credit for pool scrubbing for GESSAR II Chapter 15 accident evaluations.

In that meeting, the Staff indicated that if GE provided the analysis of reduced radiological doses resulting from pool scrubbi: g, the Staff would recognize the potential decontamination capability attributed to the pool. This would be reflected in the GESSAR II SER in March, 1983.

In addition the Staff would provide a detailed review of the submittal and would prepare a supplement to the SER which would include pool scrubbing credit.

It was indicated such a supplement could be issued by September 1983.

In accordance with this approach, GE submitted the required information on the GESSAR II docket (reference letter).

As discussed in this meeting the January 25 submittal culminates several technical meetings with the Staff on the subject of suppression pool decontamination factors.

[000 C303140179 830309 i

PDR ADOCK 05000447 A

PDR

GENERAL $ ELECTRIC U. S. Nuclear Regulatory Commission

-Page 2 If there are any questions on the information provided herein, or in the referenced letter, please contact J. N. Fox of my staff at (408) 925-5039.

Very truly yours, Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation GGS: pes /112R Attachment cc:

F. J. Miraglia, NRC D. C. Scaletti, NRC L. G. Hulman, NRC C. O. Thomas, NRC R. M. Ketchell, GE (Washington Liaison Office)

L. S. Gifford, GE-(Bethesda Liaison Office)

GEN ER AL h. ELECTRIC NUCLEAR POWER i

o SYSTEMS DIVISION l GENERAL ELECTRIC COMPANY,175 CURTNER AVE SAN JOSE. CALIFORNIA 95125.

MC 682, (408) 925-5040 FN-OH-83 January 25, 1933 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 Loss 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 in Section 15D. 2, the suppression pool provides an extmmely effective fission product retention nechanism. General Electric personnel have been working with your staff during 1982 to facilitate the review of these test results and their application to GESSAR's Chapter 15 analysis.

This revision demonstrates the capability of the suppmssion pool to significantly reduce offsite radiological doses for Design Basis Accident conditions.

Following your preliminary review we will schedule the submittal of an amendment to implement this revision.

Very truly yours, 4.<

Glenn G.

erwo a a er Nuclear Safety and Licensing Operation Attachments cc:

F.J. Miraglia, w/o att.

4.C. Scaletti L.G. Hulman C.0. Thomas, w/c/ att.

L.S. Gifford, w/c/att.

i3U/c5NN3

mMias gg 22A7007 238 NUCLEAR ISIAND Rav. 0 3

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6.4.2.4 Interaction With other zones and Pressure-Containing Equipment (Continued) outdoor air are mixed and drawn through filters, a cooling coil and zone electric reheating coils.

~

There are 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 and 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 HEPA and charcoal filtering system for cleanup before being used for pressurization.

(*

The control room must remain habitable during emergency condi-tions.

In order to make this possible, potential sources of danger such as steam lines, pressure vessels, CO fire fighting 2

containers, etc. are located outside of the control room and removed from the compartments containing Control Building life support systems.

i 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.

All dampers except the mixing dampers in the air conditioning units are of the two position (open or closed) type.

6.4-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l

6.4.2.5 Shielding Design j

l 6.4.2.5.1 Design Basis The Control Build'ing shi~elding design is based upon adequately protecting against the radiation resulting from an incid;;t ;;

pcsroATEb a LOCA.

H e dose rates received under nor-3 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 anticipated 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.

R'dioactivity released by an inadequate response to LOCA can a

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" on the Control Building.

(Those remaining in the l

reactor vessel, however, contribute negligibly to this effect.)

l (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 fiss, ion products released from the contain-ment will be taken into the Control Building via the building ventilation system air intake's.

The majority l

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 gas'es 6.4-12

~

GESSAR II 22A7007 238 NUCLEAR ISIAND R2v. O E

(

6.4.2.5.1 Design Basis (Continued) collect on the filters.

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 i

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.

I i

The DBA analysis is structured on the conservative NRC assumptions.

Theh$EN;;;; ;f 105", r t:0 powee fission products;quilibri=

l ine: teri:: releaseda_,Ron THE REACTOR VESSEL.

F

ter ;;;;;l :n;4K CCCQqHEqT To D
tt:

c.;;nc..

1:::: fre-the centai-- ent are given 5:1 = ls PRESEwrED eM

{.

sesscc.nm 15.6.S o u g ;,;;

R:10:::d frc:

":ler:: free rie:irn " refect Et::t;

" :::1 C;ntain nt yd,1.

r...

g

al;;;..;

4M 3%

6 emes ":;1i;ib1:

The containment leak rate assumed for the design analyses is 1.0%

of the containment volume per day.

Radioactive decay during I

transport through the containment is taken into account.

The j

leaked radioactivity goes into the4;Q..cNP/yQ CM., :6WMEAJ T'lu;,

SE

,u ng nnu and 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 gssum?

to be 99%

d onwc ws funcumit res.rs o efficient for filtering 4see6eiodinep, and none of the vented gas is assumed to bypass the' filter.

I.

6.4-13

---m

___.,-m---e,,,_.,,,m

,_.,~..g.------.m.-v-ww sev we

--sm...--v,------w

-ne-.--o-c-<--w-------,-.-.--y------v-y-

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 6.4.2.5.2 Source Terms wgTWELL AMO PoTENnALLY AVAILABLA Containment sources " shining" on the Control BuildingAare listed fcR RELEASE 3

in Table 15.6.

Source terms for the cloud and filter are

",vnrui-g consistent with the activity releases of Table 15.6 b.

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/squecemeteM).

T;m... 1.ec, 20.1. d fre.

" ::. c l u : 7:r : il nt C:nt :1 n :: ":ntil:ti:n Cy t = 0 :i;n f,,; ":: ting C; :::1 0::ign Criteri:n 10", by "=p: :nd." rphy, cr pre:ented h:ler:

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, _ _ _ i_3,

s ae w. - - -,.,

Ti...

rcrica LOO?. in C :: " nit

^ ^ hre 0r4944-1 0 2-y:

4s-9996-o,..

For the cloud source term, no credit was taken for decay between i

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:1cul:ted rh:1: h dy d:::: t: ::ntr:1 ::e; p;r nn:1 ::: gi::n h:1.; f

n ;;;;;;d :in 2: nth p;;t LOCT. peri;d.

0;; hund::d p:::::t ::: p:::y f::t:: i: ::: = di

%LE bow GArttin DOSES CALCULATED BY COMSEgvArts/ELY A S Sur*sk)G l00% OccoPAnc.V FOR A SIX - r10MTH foST-LOCA PERIOD ARC GivCN RELou FOR THE COMThrNt4ENT, CLOUD AND flL TEk $///NE CcNTKtBor/ 0N S, 6.4-14

238 NUCLEAR 2SLhMD R3v. 0

..e 6.4.2.5.3 Results (Continued)

Dose Location (Dose in Rem)

Component Control Room Equipment Room Containment Shine 0.0054 0.0063 Clotid O.19 0.20 Charcoal Filter 0.0012 -

0.078 0.20 0.28

%dnoLL cod 6T*tYkomb,no scia DOGES LT AG IRen nt AttBORNE A cT1 wry wirkon nt cou*c Room ARE 7Cl) IN SECnoN ANO/SRC TD WELL WITHIN T~ ore Lot 4 tTS WL 19 c{f 50, _e cej1e_,ggo wnosc t __c y gu

_,,,,,,_., L@99 of_qc sn A v.

gggigg.w. 6 hic % ALW 70Vselisis"ciZs,MkEE5&. _ _If5d55 $5E75 UYo[~~~ '

impli;; th; ;;;;pt;tility ;

.lgoccupancy of these areas for the ocRario.

orTHEpost-LOCA perio$'Yu'pancy will, however, have to be restricted somewhat in the chiller rooms at El (+)28'-6" during the first day post-LOCA, 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 l

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 considered to be significant for l

the following reasons.

The penetrations are all relatively small (cabletrayslessthan24in.ypipeslessthan6in.)[ndnot i

radially aligned with the containment.

Also, the overall con-tribution of the containment source component is quite small; thus, an increase due to streaming 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-tion of this figure shows no credible streaming for the external cloud source.

Finally, arrangement of the filter cubicle precludes t

6.4-15

i t

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.4.2.5.3 Results (Continued)

}

door streaming.

There are no floor penetrations of any consequence, cnd wall penetrations are kept well away from the source region. -

Syst'm Operational Procedures 6.4.3 e

During normal operation, the control room group operates with mixed recirculated and fresh air, which pressurizes the subject spcces.

Emergency conditions such as LOCA or high radiation cause automatic change over to reduced outside air and charcoal filtering of all outside air to effectively isolate operating parsonnel 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).

Datection of radioactivity is instrumented, and changeover to rcduced circulation and charcoal filtering is automatic.

Redun-

~

dancy of instrumentation and air handling systems ensures against system failure due to single component failure.

The above operational description is brief.

For a more detailed doccription of normal and emergency operation of the control room habitability systems, see Subsections 9.4.1, 9.5.1, 9.5.3, 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 i

Accumptions used in the generation of post-LOCA radiation source terms are described fully in Subsections 0.2.0.; er.0 15.6.~5.

s

)

6.4-16

GESSAR II 22A7007 c,

238 NUCLEAR ISLAND R3v. 6

- ' 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 consequence 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.

\\

l 15.6.5.4 Barrier Performance l

The design basis for the containment is to maintain its integrity cnd =T ExCGEO LNEL C ACctPTAmE CRorthA no 3

p:n:n:: n:rr:1 etrerer

fter the instantaneous rupture of the largest single primary systern piping 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 result in exceeding the containment design limit l

(see 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 l

cccident:

(1)

The first is based on conservativt assumptions con-i sidered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet l

1 l

15.6-14

,--s-g e-,w---

---ww-- - - - - - -.-

. t i tttl LQ3 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 alistic estimate of radiological consequences.

a This analysis is referred to as the " realistic analysis",

As.THon.N HMC CONSCRVATivE A550MEDOWS STN. EfM4W.

A schematic of the transport pathway is shown in Figure 15.6-2.

Additional parameters and information for specific design basis l

e accidents are provided in subsection 19.3.15.1.

m

.s 15.6.5.5.1 INSERT Design Basis Analysis 4

b O.; ;;;thed;, ;;;;;.ptir.; r.d senditien; e;;d te ;;;1;_t; thi;

id;;t =; in
: f=;; rith th;;; s;id:lin;; ;;t f::th in i s;1;t; j C;id;; 1.2 = d 1.7.

Th; :p;;ifi; x 2:1:, ::: r;tir;

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t;r: :::d in thi; ;;;1;; tic; 7.r; pre;;;ted in Tahi; 1";.:; 7.

15.6.5.5.1.1 Fission Product Release from Fuel Insws$ is assumed that 100% of the noble gases and 50% o B

(are released from an equilibrium core operating at a power level of 3651 MWt for 1000 days prior to the accident.

While not specifically stateil in Regulatory Guide 1.3, the assumed release

/

of 100% of the core noble gas activity and 50% of the iodine b

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 of this accident.

Of this release, 100% of the noble gases and 50% of the iodine.become airborne.

The remaining 50% of the iodine is removed by plate-out and condensation, therefore, it is not available for airborne release to the environment..The

(

~

vity airborne in the containment is presented in Table 15.6-8 15.6-15

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se-------,------,,,r-

I i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev.0-

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i 6.5.5.1.2 Fission Product Transport to the Environment

..}

'The transport pathway consists of leakage from the containment 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 l

prinary containment and its penetrations (excluding the 1

main steamlines) is 1.04/ day for the duration of the accident.

All of this leakage is to the secondary con-

[

l 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 l

rate, leakage rate, and mixing ratio are given en Tables 15.6-9 and 15.6-10.

(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 y rgo may be j

utilizedisonehouraftertheaccidenthateof100scfm l

minimum.

The purge would be processed by SGTS prior to releah to the environment. S,ncE THE HYDRoGE4 CONTROL EdviPt(CN IS E EQustr1EMT, IT Is Mar ASGurtEO TO FAIL. N EVAt.OArtAIG rhE PorEnrohn RAOoLOGICAI EXf0SURES ASSOGATED WirH TWIS ACCsDENT, Fission product release to the environment based on the above assumptiorsis given in Table 15.6-11.

15.6.5.5.1.3 Results

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15.6-1s

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/NSERT A (page 15,6-IS)

The matheds, assumptions and conditions used to evaluate the potentia radiological exposure to onsite and offsite personnel are consistent with the guidelines set forth in Regulatory Guide 1.3 except as noted below The guidance of Regulatory Guide 1.3 has been supplemented using free:

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 fission 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.

e--

--~-.__-,,----we--

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m,,,ww,,

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p_

o f NSE.RT 3 ( p f5.6-15)

It is assumed that 100% of the noble gases and 5C% of the iodine are released to the drywell from an equilibrium core operat at a power level of 3651 W t for 1000 days prior to the accident. These assumed releases imply that severely degraded ECCS performance has resulted 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 iodine 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 iodina released, 50% is assumed to be removed by plateout and condensation and therefore is not available for potential release to the environment.

s,6.S.S.I,2 Fiss,oa PRecocr Tiuwsecar To rue Esmear i

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 mass through the suppression pool before they can reach the containment wetw11 alirspace. 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 MeeeTransportpathwaysandanyassociatedretentionmechanismsare identified below:

1 1

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e=-+..--,,--------.-r-.

wwm,

--.--e

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lNSERT]b CONTNo$b 1)

Suppression Pool - All fission products discharged from ge drywell or safety / relief valves enter the suppression pool 7 to3 + feet 4

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 discharged into the pool under these conditions.

Based upon the testing and modeling documented in Section 15 D.2, a suppression pool scrubbir:g 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 forms 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 Leakane - 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 re w ire r rteri to,the primary and secondary containments are given on Tables 15.6*9 and 15.6-10,;nd tr.: ::th'ty :idrx h it.: ;;rin y :=t:f rnnt S l

,r ;;;ted in T d e 10.0 0.

i 3)

Engineered Safety Feature Leakage - Engineered safety feature (ESF) components located outside the primary containment contribute insignificant leakage to the secondary containment.

e w

=w-

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wyw,,-w,c--

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,,,.,,,, _ _.,,9_, _., _,,,.,,.,,,,

=

gggg7 C ( 4 6-15 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 consdrvative assumption 5used in the design basis analysis.

For example, the calculated exposures 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 art present,in addition to the suppression pool,which would act to limit the release of fission products to the environment.

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.

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 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 anyincreaseintheparticulatereleasebyafactorof108,theeffecton theoffsitekeNuldbenegligible. Only the thyroid dose would be l

calculated to change, with an increase from 0.05 rem to 01 rem in the Exclusion Area and 0.1 rem to 0.2. res in the low population zone.

Thus, the results would still remain well within the guidelines of 10CFR100.

l I

l t

GESSAR JI 238 NUCLEAR ISIAND 22A7007 l

Riv. 0

},

15.6.5.5.2 Realistic Analysis The realistic analysis is based on a realistic but still conserva ti lNSW,ve assessment of this accident.

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12:ti:n :::

pr ::nt:d in T_il. 15.0 7.

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 in the reactor coolant plus any additional activity which may be released as a corisequence of reactor scram and vessel depressurization.

r While there are vp ous activation and corrosion products contained

(

in the reactor coolant, the products of primary importance are I

the iodine isotopes I-131 to I-135.

The coolant concentration for these isotopes is:

I-131

?:t3 2 2."2 01,';. O.02./t Ci/gm I-132 1. ;^ 0 1%.0i,'; - O.2.6 XC/Sm I-133 -1.40 1.l5,.Oi,'s. O.lS M.Ci/9m i

I-134 4.75 1."* Ci,';m 0. 6 n.C; e

I_135

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e centerv:tir:1y ::: = :d th:t 'Ot :f th: r:102:06 l

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1 I

l l

15.6-17

i

.o GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15.6.5.5.2.1 Fission Product Release from Fuel (Continued)

'}

As a consequence of reactor scram and depressurization, additional.

iodine activity is released from those rods which experienced -

cladding perforation during normal operation.

Measurements performed (Referenc

9) 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)

(drobabilb 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.

Wh44+$NCE 3no measurements have been obtained during a pressure transient as rapid as the LOCA, it is difficult tio predict the actual release rate from the fuel as a consequence of iodine

~}

spiking.

Therefore, it is arbitrarily assumed that 100% of th HAS REED 7

SEMT A S spiking source term % released 9 r4 DRWELL SEFCRE ThE Bicw0cwap;;ief th:t t00 O/fR.

3

..ng...;

mf the it;;h;;5; ;;;1:at i: fiz ht - m.

TisE. Ico.He FRacnous IrlVOLVED IN ntIS RnlAst PRE CONStGTENT WITH TNcsc sPEcirtEO lM hlGREG 0772 De. 9987% IM PARRCUATE FoRtf AND O.039f AS CRGAAttc toDrNE),

It is also assumed that plate-out and condensation removef 504 PARDCnME of the airborne 3 iodine activity.

... t;tel ;;tivity ;ich:::: in th centeinaeat is pz;;;at;0 in T;hi; 15.0 14.

+

15.6.5.5.2.2 Fission Product Transport to the Environment INSERT 4 E

Th; 1::h ::t: fr S: p:' T; ;;;tain ;nt te th; ;;;;ncerj

trir : t i: 1. 0 5,'d a y

--h e r : 100t ri=ing i:,2:: r:f 'a :::::.

"ci:::: f :: Se :::: fr-; :::trir t 'a 2: :: irch : t vi 00.00 i:fia: Offici::t STS i: p ::: td in T:ti: 15.5-15.

The int:g :t:f'i :t:;i: ::tivity ::lered t: $2 : ricer: t ir pr -

se,. Led in T;hi; 15.5-15.

~

9 15.6-18

luseer 3 (p.154-17)

The chemical forms of iodine used in ~this realistic' analysis are con'sistent with those specified in NUREG-0772. As 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 described in Refetence 2.

l l

l 1

y

NGERT E WI5.$~IS The activity released in ihis analysis is directed to the drywell and from there into the nGppression pool. Any fission products not retained in the suppression pool due to scrubbing are assumed to be distributed la the wetwell airspace.

This transport 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 watwell 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 efficiencyMD A 99,9%. PennCddTE /cosNE REmv4L EfftC dNCf.

l 1

l 1

l t

,,7-,-.----y.---r w

v=-

=--

v n--5 m--

4

  • -n-*

--e-e*--

--e-'-

M ssa mero 238 NUCLEAR ISLAND R3v. 0

,h 15.6.5.5.2.3 Results As sw w Tesa: 15. 6 - 14, The calculated radiological exposures for this event are W 3 L. M 'e 10.0 10 rf :: ci r. r; a small fraction of 10CFR100.

15.6.5.5.3 control Room lNSW -+

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 c 30-day period after a loss-of-coolant accident are:

Dose U.S. NRC Limit (Rem)

(Rem)

Whole Body 2.56 5

Inyroid 29.4 30 Beta 53.8 75 A factor of 1/4 was taken into account for a dual inlet with j

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)

Source Terms The source terms used in this analysis are consistent with R.G. 1.3 (i.e., 251 halogens and 1004 noble gases l

airborne in the containment) and were presented in Table 15.6-8.

(

l l

l 15.6-19 l

,c.

~

GESSAR II 22A7007 238 NUCLEAR ISLAWD Rev. 0 15.6.5.5.3 control Room (Continued) s hDDITIONhl ASSUMPrloMS OM VEh)TILArron FAPArtETERS Ab)D HEWoROLOGscAL DarA ARE GIVEN BELOW

% ventIIation Parameter (SEE Focu ScrtENATIC.

Inlet air flowp 3

filtered 0.944 m /sec unfiltered 0.0014 m3/sec INLErFilter NNeiency 99%

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 days 0.4 Meteorology Data

')

3 X/Q Values sec/m 0.-2 hrs 8.0 E-3 2-8 hrs 3,2 M E-3 8-24 hrs 2,7 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 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 postulatied break of the feedwater line, representing the largest liquid line outside the containment, provides the, envelope 15.6-20

NSERT C(),*lS.b-0)

~

tim E The control room and its associated' ventilation system hee been designed 3

with the objective of continuous occupancy following a LOCA. An analysis has therefore been performed to demonstrate that the ventilation system satisfies the NRC's control room habitability guidelines relative to radiation exposure. The potential doses to control room personnel during.

a 30-day period after e LOCA 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 methods presented in Reference 4.

Dose U.S. NRC Limit Heml (Rem)

~

Whole Body 2.94 5

Thyroid 0.02 30 Beta 36 75 x A St4ALs. INCRErtEsir TO THE WHouE 80s.4 C0%E REGari fROM gar 1MA SccRCES EsTEMAl. Tb THE CUNTKbd Room. TTits EMCr (S E VA t.GATCO

.a secnov 6.V.2.5.

  • O

(

GESSAR II 22A7007 1

~

238 NUCLEAR ISLAND R;v. '6

)

]

~15.6.7 References 1.

F. J. Moody, " Maximum Two-Phase Vessel Blowdown from Pipes",

ASME Paper Number 65-MA/HT-1, March 15, 1965.

2.

7. 7. Ot;nc;;;;;

nd E. 2. M: ; =, "C :::: :tiv; "-df*-

1;gic;l 1;;id;nt T;;1; tic: - The C^"?.C'01 0:d;", :::h 10 75

'!!!"^- 211 f ? ?.

I 3.

";;y::, ""::licti: ?.;;id;nt 7.n:1y;is T:.; n:U.C Ced;",

^

Ortche 19 '"'

'M"^-211 e ? ?.

l K3 F. J.

rutschy, G. R. Hills, N. R. Norton, A. J. Id. vine,

" Behavior of Iodine in Reactor Water During Plant Shutdown and Startup*, August 1972 (NEDO-10585).

t L. G. *,;;i;; and 'J.

O.

!; ica, "C nt ;l 0;;;. A;;ident Exp;;; : 0;;1 :ti:n 0""O " :g

.", J;;;;.ry 1070

'!:: " ^ 2 ? ? ? ? ?

I

  • 2 D. NGuYEs), ed al, EADtos.0GILAL Acc,oeur Evatonrion-Tu e CoNAc 03 Cool, Decs"ses 1981 (m-2w3-0, a y D Nauven, etaI, "Coartet Roen Accioem Ex PoSORE EVAwAnod-CR005' PRosene,' FesRanRY R81 (NM-23909 A)

E

~

e 15.6-27/15.6-28 l

I r

l t

GESSAR II 22A7007 1

238 NUCLEAR ISIAND Rav. 4 i.

I Table 15.6-6 STEAMLINE BREAK ACCIDENT (REALISTIC ANALYS15)

RADIOLOGICAL EFFECTS.

Whole Body Dose Inhalation Dose (rem)

(rem)

Exclusion Area 1.1E-2 5.2E-1, Low Population zone 5.4E-3 2.6E-1 m.

o,..

'4 G

15.6-35

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 15.6-7 l

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 lool ocuk loc /, NesstGAE c

S0% Tcoua 50Je' ico,~E

  • E.

Iodine fractions (1) Organic

  1. O.0003 XO.0003 (2) Elemental 20 EO (3) Particulate EO.9997 XO.9997 F. Reactor coolant activity NA 15.6.5.5.2.1 before the accident II.

Data and assumptions used to estimate activity released A. Primary containment leak TAgg l$.6-10 TA8g 15.WO

)

l rate (^./de,)

t re-B. Secondary containment leak 0.2 hre 210.2 120 rate (^/ day; 7ABuf IS.6-lO TAgg 15,6-/O C.

10 h;; ::,0 122 DK. Valve movement times NA NA g E. Adsorption and filtration efficiencies (4)

(1) Organic iodine

-NA 99 NA 93 (2) Elemental iodine 99 NA

!MMJ MA (3) Particulate iodine 4ih 99 NA 99,3 (4) Particulate fission products 43h99

-NA *!7,3 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 para:naters (flow rate, drop size, etc.)

NA NA H E. Containment Volumes NA NA g Jr. All other pertinent data and assumptions None None s

C, S>PPREsstoA) Poot

)

RUBBING DEcouranAnoa VA@t (0 ORGANIC.

ICDtdE 15.6-36 (n PAf ttfDr Arc loAsur 10,000

_ lopCC - _

[

GESSAR II 22A7007 238 NUCLEAR ISLAND t

Rsv. O k,

Table 15.6-7 (Continued)

Design Realistic Basis Basis Assumptions Assumptions III. Dispersion Rate A. Boundary and LPZ distance (m)

B. x/Q's for time intervals of (1) 0-2 hr - SB/LPS

2. 0E-3/1. 0E-3 2. 0E-3/1. 0E-3 1.0E-3 1.0E-3 I

(2) 2-8 hr - LPZ 3.8E-4 3.8E-4 i

(3) 8-24 hr - LPZ 1.0E-4 1.0E-4 (4) 1-4 days - LPZ 3.4E-5 3.4E-5 (5) 4-30 days - LPZ 7.5E-6 7.5E-6 IV.

Dose Data A. Method of dose calculation Reference 2 Reference 5 2 B. Dose conversion assumptions Reference 2 Reference gZ C. Peak activity concentrations Table Table in containmentWcrwrcc AIRSPACE 15.6-8 15.6-14 D. Doses Table Table 15.6-12 15.6-16

  • Applicant to Supply 15.6-37

L M (O.

.-g th

~

Q. f L ?

Table 15.6-8 2.

LOSS-OP-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS) mM#

ACTIVITY AIRBORNE IN PRIMARY CONTAINMENT (Ci)

(

x Isotope 1 min 30 min 1 hr 2 hr 4 hr 8 hr 12 hr 1 day 4 day 3C day \\

I131

'2.lE 07 2.lE 07 2.lE 07 2.lE 07 2.lE 07 2.lE 07 2.0E 07 1.9E 07 1.5E 07 1.2E 06 I132 3.5E 07 3.0E 07 2.6E 07 1.9E 07

'l.0E 07 3.lE 06 9.2E 05 2.4E 04 7.3E-06 0.

I133 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 0.

O.

1135 4.68 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.

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 y

Kr83n 9.5E 06 8.0E 06 6.6E 06 4.5E 06 2.lE 06 4.8E 05 1.lE 05 1.2E 03 2.2E-09 0.

$o om Kr85m 2.3E 07 2.lE 07 2.0E 07 1.7E 07 1.2E 07 6.6E 06 3.6E 06 5.5E 05 7.6E 00 0.

Q$

1 Kr85 5.9E 05 5.9E 05 5.98 05 5.9E 05 5.9E 05 5.9E 05 5.8E 05 5.8E 05 5.6E 05 4.3E 05 Kr87 4.7E 07 3.6E 07 2.7B 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.lE 07 2.5E 07 9.28 06 3.4E 06 1.7E 05 2.9E-03 0.

Nz Kr89 6.7E 07 1.2E 05 1.6E 02 3.0E-04 1.lE-15 O.

O.

O.

O.

O.

jD X 131m 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 X:133m 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 X133 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 X135m 3.6E 07 9.8E 06 2.5E 06 1.7E 05 7.2E 02 1.4E-02 2.6E-O*1 0.

O.

O.

X3135 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.

X 137 1.5E OS 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.

y a>

Total idG 7.4E 08 3.5E 08 3.0E 08 2.5E 08 2.2E 08 1.8E 08 1.6E 08 1.4E 08 8.35 07 2.4E

,4 y o8.

,r

GESSAR II 22A7000 l

238 WUCLEAR ISLAND R0v. O i

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GESSAR II 22A7007 238 NUCLEAR ISIAND Rsv. 0

.-(</

Table 15.5-9 SHIELD BUILDING EXBAUST RATE Time Average Exhaust Flow Rate to SGTS-(

,hr)~-

(SCFM) 0-2 480 2 - 10 90

>10 66 Table 15.6-10 LEAKAGE RATES AND MIXING RATIO Numerical Value Parameter Design Basis Realistic

{-

A.

Primary to Secondary Containment (t/ day) 0 - 2 hr O.832 h+0.932 2 - 10 hr 0.903

-h*0.903

>10 hr 0.908 h40.909 B.

Primary Containment Leakage to SGTS (t/ day) 0 - 2 hr 0.168

    • O.169 l

2 - 10 hr 0.097 45A. O.097

>10 hr 0.092 MO.CR2 C.

Secondary Containment Leakage to SGTS (t/ day) 0 - 2 hr 319.3 i+3 IN b 2 - 10 hr 59.9 i+3 ~28.9

>10 hr 43.6 1+3-22.0 D.,

Mixing Efficiency (t)

Primary Containment 100 100 Shield Building Annulus 50 100

(

15.6-39 v-y r

,,-_,--,-----...,m._

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

C Z. D

~

m

~ ~

Table 15.6-11 WQY i

L.

1 LOSS-OP-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS) f ACTIVITY RELEASE TO ENVIRONMENT (Ci) 03 b I

I h5T-l Isotope 1 min 30 min I hr 2 hr 4 hr 8 hr 12 hr 1 day 4 day 30 day 1131 2.5E-01 8.7E 00 2.0E 01 4.8E 01 7.5E 01 1.5E 02 2.4E 02 5.7E 02 3.9E 03 1.8B 04 1132 4.13-01 1.3E 01 2.8E 01 5.7E.01 7.5E 01 9.6E 01 1.02 02 1.lE 02 1.1E 02 'l.lE 02 1133 3.8E-01 1.38 01 2.9E 01 7.0E 01 1.lE 02 2.0E 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 6.42 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 u

w Total I 2.2E 00 7.lE 01 1.5E 02 3.2E 02 4.6E O2 7.4E O2 1.0E 03 1.8E 03 6.2E 03 2.0E 04 I

bo Er83m 1.15 01 3.5E 02 7.3E 02 1.4E 03 1.9E 03 2.2E 03 2.3E 03 2.4B 03 2.4E 03 2.4E 03 0g Er85m 2.75 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 l

Er85 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 Er87 5.5E 01 1.7E 03 3.3E 03 6.05 03 7.2E 03 7.9E 03 8.0E 03 8.lE 03 8.1E 03 8.1E 03 gg "M

E Er88 7.8E 01 2.5E 03 5.4E 03 1.lE 04 1.6E 04 2.lE 04 2.4E 04 2.5E 04 2.58 04 2.55 04 g

Kr89 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 O2 4.7E 02 Xe131m 6.75-01 2.3E 01 5.3E 01 1.3E 02 2.0E 02 4.05 02 6.5E 02 1.6E 03 1.lE 04 6.4E 04 Xel33m 2.7E 01 9.3E 02 2.lE 03 5.lE 03 7.9E 03 1.6E 04 2.4E 04 5.5E 04 2.6E 05 4.5E 05 Mel33 1.6E 02 5.4E 03 1.2E 04 3.OE 04 4.7E 04 9.3E 04 1.5E 05 3.5E 05 2.2E 06 7.35 06 Xel35m 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 Xel35 2.85 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 4.0E 04 l

Xel31 1.9E 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 1.25 03 l

Xel38 1.95 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 g

Total NG 9.05 02 1.9E 04 3.5E 04 7.lE 04 1.0E 05 1.7E 05 2.4E 05 4.9E 05 2.6E 05 8.1E 06, a

d

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238 WUCI2AR ISIAND R v. 0

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238 NUCLEAR ISIAND Rev. 4 Table 15.6-12 IDSS--OF-COOLANT ACCIDENT (DESIGN RASE ANALYSIS)

RADIOLOGICAL EFFECTS.

Whole Body Dose Inhalation.Do.se (rem)

(rem) 46,+0.05 Exclusion Area 19./l Low Population Ione 14.1

&+reO.ll

~

Table 15.6-13 ISOTOPIC SPIKING ACTIVITY The 95th Cumulative Probability Spiking Isotope Name Activity (Ci/ bundle)

I131 2.14 1132 3.21 I133 5.03 1134 5.44 1135 4.79 I

Kr83m 9.04-1*

Kr85m 2.23+0 Kr85 4.90-1 Kr87 4.33+0 b8 6.12+0 Kr39 7.96+0 6.60-2 Xe131m Xe133m 3.26-1 Xe133 1.16+1 Xel35m 1.80+0 Xe135 1.10+1 Xe137 1.05+1 1.06+1 Xe138

~

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Tablo 15.6-14 h&

LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS) 8 ACTIVITY AIRBORNE IN THE CONTAINMENT (Ci)

% 3)

Q>h a.c Isotope 1 min 1 hr 2 hr 8 hr 1 day 3 day 26 day 1

~

3131 8.59E 01 8.55E 01 8.52E 01 8.33E 01 7.84E 01 5.96E 01 5.64E 00 I

f I132 1.41E 02 1.05E 02 7.,73E 01 1.25E 01 9.69E-02 2.99E-11 0.

3133 2.08E 02 2.01E 02 1.94E 02 1.59E 02 9.28E 01 8.26E 00 6.91E-09 I134 2.39E 02 1.10E 02 4.97E 01 4.29E-01 1.34E-06 0.

O.

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M 8 77E 02 '

6.845 02 5.71E 02 3.43E 02 1.88E 02 6.19E 01 5.64E 00 total y

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Q Br85m 1.67E 03 1.43E 03 1.23E 03 4.83E 02 4.03E 01 5.55E-04 0.

Ob s

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O.

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O.

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/

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Tcblo 15.6-14 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)

ACTIVITY AIRBORNE IN THE CONTAINMENT 3(C1) i VETkIELL ISOTOPE 1-MIN 10-MIN 1-H0llR 2-HOUR 4-HOUR R-HOUR 12-HR 1-DAY 4-DAY 30-DAT j-l 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 R.60E-01 R.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 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 i

1-134 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 I-135 1.28E 00 1.26E 00 1 1SE 00 1.03F 00 B.3PE-01 5.50E-01 3.61E-01 1.02E-01 5 21E-051.00E-20 5.4RE 00 5.25E 00 4.30E 00 3.61E 00 2.*:RE 00 2.19E 00 1.81E 00 1.21E 00 4.36E-01 3.14E-02 TOTAt' I a

i -

II h

KR-83M 6.72E 02 6.35E 02 4.63E 02 3.17E 02 1.49E 02 3.26E 01 7.15E 00 7 55E-02 1 05E-13 1.00E-20

!I KR-RSM 1.67E 03 1.63E 03 1.43E 03 1.22E 03 8.9AE 02 4.83E 0'? 2.60E 02 4.03E 01 5.69E-04 1.00E-20 jk 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 7CE 02 "U'

KR-87 3.21E 03 2.96E 03 1.88E 03 1.09E 03 3.66E 02 4.13E 01 4.65E 00 6.69E-03 1.00E-20 1.00E-20 i

,h KR-88 4.56E 03 4.40E 03 3.59E 03 2.R1E 03 1.72E 03 6.48E 02 2.44E 02 1.30E 01 2.93E-07 1.00E-20 g

KR-89 4.78E 03 6.6RE 02 1.19E-02 2.40E-0R 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1 00E-20 1

XE131M 4.94E 01 4.94E 01 4.93E 01 4.91E 01 4.88E 01 4.83E 01 4.77E 01 4.61E 01. 3.76E 01 6.3RE 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 03 8.63E 03 8.59E 03 8.54E 03 8.44E 03 8.24E 03 8.05F 03 7.4*E 03 4.89E 03 1.21E 02 XE135M 1.2RE 03 R.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 B.10E 03 7.59E 03 7.03E 03 6.03E 03 4.44E 03 3.27E 03 1.30E 03 5.1PE 00 1.00E-20 XE-137 6.54E 03 1.28E 03 1.51E-01 2.91E-06 1.08F-15 1.00E-20 1.00E-20 1.00E-20 1.00E-20 1.00E-20 XE-13R 7.58E 03 4.88E 03 4.23E 02 2.2iE 01 6.33E-02 5.03E-n7 4.00E-12 1.00E-20 1.00E-20 1 00E-20 TOTAL. NO 4.78E 04 3.47E 04 2.47E 04 2.17E 04 1.93E 04 1.45E 04 1 24E 04 9.43E 03 5.35E 03 3 98E 02 w

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m ct, Table 15.6-15 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)

. bqh h

ACTIVITY RELEASED TO THE ENVIRONMENT (C1) r*

C Isotope 1 min 1 h--

2 hrs 8 hrs 1 day 4 days 30 days k,

1131 2.54E-10 8.98E-07 3.52E-06 5.04E-05 3.44E-04 2.24E-03

.8.20E-03 I132 4.19E-10 1.22E-06 3.95E-06 2.07E-05 2.80E-05 2.81E-05

~ 2.81E-05 1133 6.14E-10 2.13E-06 8.19E-06 1.04E-04 5.39E-04 1.46E-03 1.56E-03 1134 7.12E-10 1.53E-06 3.76E-06 7.66E-06 7.74E-06 7.74E-06 ' 7.74E-06 I135 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 g

w y

Kr83m-1.99E-06 5.54E-03 1.72E-02 7.48E-02 9.03E-02 9.04E-02 9.04E-02 Eo

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gg Kr85 1.08E-07 3.84E-03 1.51E-02 2.19E-01 1.55E 00 1.15E 01 9.29E 01 EE Kr87 9.53E-06 2.38E-02 6.68E-02 1.99E-01 2.llE-01 2.11E-01 2.11E-01 g[

Kr88 1.35E-05 4.07E-02 1.37E-01 8.40E-01 1.30E 00 1.32E 00 1.32E 00

{

Kr89 1.52E-05 7.28E-04 7.28E-04 7.28E-04 7.28E-04 7.28E-04 7.28E-04 0

Xel31m 1.46E-07 5.16E-04 2.03E-03 2.91E-02 2.01E-01 1.36E 00 6.01E 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

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Xel35m 3.86E-06 2.89E-03 3.70E-03 3.81E-03 3.81E-03 3.81E-03 3.81E-03 1

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

\\

Xel38 2.28E-05 1.54E-02 1.90E-02 i.94E-02 1.94E-02 1.94E-02

'l.94E-02

\\

w E5 Total 1.44E-04 2.85E-01 9.88E-01 1.03E 01 5.10E 01 2.40E 02 6.70E 02

<,g 00

/

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0-Table 15.6-16 LOSS-OF-COOLANT' ACCIDENT (REALISTIC ANALYSIS)

RADIOLOGICAL EFFECTS Whole Body Dose Inhalation Dose' (rem)

(rem) 5.y:1. E-6 1.2E-3 Exclusion Area 2.2 0

0 9.7E-4 23E-6 Low Population Zone

3. ;;;-4 6M I

Table 15.6-17 l

SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK OUTSIDE CONTAINMENT Time Event O see One feedwater line breaks.

l 0+ sec Feedwater line check valves isolate the reactor from l

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 procedure established.

  • Applicant to Supply 15.6-44

r

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GESSAR II 22A7007 i.-

238 NUCLEAR ISIAND R3v. 0 2

4 SECONDARY CONTAINMENT I

SNGELD BUILDING 4000 CFM MIN,

j ( TO SNtELD SUI 1 DING REciRC 3000 CrM FAN FROM SHIELD SulLDING l

i 1000 CFM MAX PRIMMtY CONTAINMENT 350 CFM M4N gNtYWFELt.

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