Text

Forwards Supplemental Info to 880520 Radiological Impact Assessment Re Vantage 5 Fuel.Evaluation Performed of Environ Consequences for FSAR Chapter 15 Accidents Impacted by Fuel Change Using Reactor Coolant & Core Source Terms
	ML20155B530
Person / Time
Site:	Summer
Issue date:	09/30/1988
From:	Bradham O; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	NUDOCS 8810060328
	Download: ML20155B530 (67)
	v • d • e

,

\\

s Cargina Ekctric & Gas Company Sout th am gkjngvge SC 29065 Nuclear Operations SCE&G e -w Septemt,er 30, 1988 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Technical Specification Change VANTAGE 5 Fuel Gentlemen:

On May 20, 1988 South Carolina Electric & Gas Company (SCE&G) requested a revision to the Technical Specifications for the Virgil C. Summer Nuclear Station (VCSNS) in support of refueling and operating with VANTAGE 5 fuel.

This submittal contained a Radiological Impact Assessment in which it concluded that the transition from current fuel to VANTAGE 5, with its extended burnup characteristics, would have a small impact on thyroid and whole body doses.

To supplement the previous Radiological Impact Assessmer.t. SCE&G has perfoimed a detailed evaluation of the environment consequences for the Final Safety Analysis Report (FSAR) Chapter 15 accidents impacted by the fuel change. This letter forwards the results of the dose calculations.

This u

evaluation used the reactor coolant and core source terms for VANTAGE 5 fuel, previously supplied on August 31, 1988, in combination with the current NRC accepted methodology for dose avaluations as described in the FSAR.

Tables 1, 2 and 3 present the limiting analysis results from the FSAR and this evaluation for those transdents impacted by the fuel change.

Due to the revised source term methodology, the transition to VANTAGE 5 fuel generally results in a small decrease in gamma and beta doses and a small increase in thyroid doses.

In all cases, the dose results are well within applicable NRC acceptance criteria.

In addition to the limiting analysis results presented in Tables 1, 2 and 3 more realistic calculations have been performed consistent with the current dose consequence presentation given in the VCSNS FSAR.

These results are j

summarized in the FSAR markups attached.

i 8910060320 000930 DR ADOCK ObO( 3D I

Document C ntrol Desk September 30, 1988 Page 2 f

This letter completes the SCE&G assessment of environmental consequences from Chapter 15 events with VANTAGE 5 fuel.

If there should be any questions,.

please do not hesitate to call.

Very truly yours, h

0. S. Bradham H08/0SB:1cd Attachments c:

D. A. Nauman/J. G. Connelly, Jr./0. W. Dixon, Jr./T. C. Nichols, Jr.

E. C. Roberts W. A. Williams, Jr.

G. O. Percival J. N. Grace R. L. Prevatte J. J. Hayes, Jr.

J. B Knotts, Jr.

General Managers H. G. Shealy C. A. Prica/R. M. Campbell, Jr.

NSRC R. B. Clary RTS (TSP 880013)

K. E. Nodland NPCF J. C. Snelson File (813.20) 4 f

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TABLE 1 CURRENT CilAPTER 15 DOSES FOR LOPAR FUEL SITE BOUNDARY RESULTS LOW POPULATION ZONE RESULTS FSAR SECTIOM TITLE Gamuna Beta Thyroid Gamuna Beta Thyroid 15.2.9 Loss of Offsite Power 9.83E-4 2.04E-3 2.77E-2 2.27E-4 4.71E-4 3.82E-3 15.3.7 Instrument Line Break 3.18E-2 3.66E-2 6.72E-1 1.8SE-3 2.13E-3 3,90E-2 15.4.1 Loss of Coolant 4.45 2.99 1.52E+c 6.87E-1 4.04E-1 2.SE+1 Accident 15.4.2 Steam Line Break 1.46E-2 8.75E-3 9.70 3.26E-3 2.53E-3 2.05 15.4.3 Steam Generator Tube 1.40E-1 S.99E-2 3.67E-1 3.54E-2 1.52E-2 2.89E-1 Rupture 15.4.4 Locked Rotor 7.87E-1 1.11 6.99 1.83E-1 2.57E-1 1.62 15.4.5 fuel llandling Accident Inside Containment 1.22 1.62 1.35E+2 N/A N/A N/A Outside Containment 1.22 1.62 6.75 15.4.5 Rod Ejection 1.82E-1 9.98E-2 S.01E+1 2.64E-2 1.47E-2 1.39E+1 i

TABLE 2 REVISED CilAPTER 15 DOSES FOR VANTAGE 5 FUEL SITE BOUNDARY RESULTS LOW POPULATION ZONE RESULTS FSAR SECTION TITLE Gasuna Beta Thyroid Gamusa Beta Thyroid 15.2.9 Loss of Offsite Power 6.65E-4 1.30E-3 2.87E-2 1.53E-4 3.00E-4 3.95E-3 15.3.7 Instrument Line Break 2.85E-2 3.20E-2 7.60E-1 1.66E-3 1.86E-3 4.41E-2 15.4.1 Loss of Coolant 2.78 2.16 1.74E42 3.57E-1 3.07E-1 2.89E+1 Accident 15.4.2 Steam Line Break 1.23E-2 8.04E-3 1.27E+1 2.70E-3 1.78E-3 2.76 I

15.4.3 Steam Generator Tube 1.88E-1 2.17E-1 4.13E-1 4.71E-2 5.46E-2 3.19E-1 Rupture 15.4.4 Locked Rotor 5.63E-1 8.27E-1 8.01 1.31E-1 1.92E-1 1.85 15.4.5 Fuel liandiing Accident Inside Containment 1.40 1.65 1.53E+2 N/A N/A N/A Outside Containment 1.40 1.65 7.66 15.4.6 Rod Ejection 1.56E-1 7.20E-2 5.28+1 2.36E-2 1.15E-2 1.46E+1 i

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F Table 3 e

CONTROL ROOM DOSES FOLLOWING A LOCA 1

l Doses (Rem)

C'Jrrent FSAR VANTAGE 5 a

Gaussa 2.26 1.70 Beta 9.29 6.30 1

Thyroid 30.00 30.00 i

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ATTACHMENT This attachment contains marked-up pages for the environmental consequence sections of Chapter 15 of the FSAR as summarized below.

ACCIDENT ITEM FSAR PAGE NO Loss of Offsite Power Section 15.2.9.4 15.2-33 Table 15.2-5 to 15.2-8 15.2-55 to 15.2-58 Instrument Line Break Section 15.3.7 15.3-14 Table 15.3-6 15.3-22 Table 15.3-7 15.3-23 Loss of Coolant Accident Section 15.4.1.4.3 15.4-14 Tables 15.4-10 to 15.4-18 15.4-75 to 15.4-79, 15.4-81 15.4-83 to 15.4-85 Steam Line Break Section 15.4.2.1.4 15.4-22 Tables 15.4-23 to 15.4-27 15.4-91 to 15.4-95 Figures 15.4-61 to 15.4-63 Steam Generator Tube Rupture Section 15.4.3.4 15.4-34 Tables 15.4-29 to 15.4-33 15.4-97 to 15.4-101 Figures 15.4-66 to 15.4-68 Locked Rotor Section 15.4.4.4 15.4-39, 15.4-40 Tables 15.4-34a to 15.4-34e 15.4-103, 15.4-105 to 15.4-108 Figures 15.4-77a to 15.4-77c Fuel Handling Accident Section 15.4.5.1 15.4-41, 15.4-47, 15.4-48 Tables 15.4-35 to 15.4-37 15.4-109 to 15.4-111 Tables 15.4-39 to 15.4-41 15.4-114 to 15.4-116 Table 15.4-50 15.4-127 Table 15.4-51 15.4-128 Rod Ejection Section 15.4.6.4.4 15.4-61 Tables 15.4-44 to 15.4-46 15.4-121 to 15.4-123

6.

Defective fuel is equal to one percent.

7.

No noble gas is dissolved in steam generator water.

8.

The iodine partition factor in the steam generators is 0.01.

9.

During the postulated accident, iodine carryover from the primary side is uniformly mixed with the water in the steam generators and is diluted by the incoming feedwater.

10. The steam release for cooling down the plant is equally contributed by all steam generators.

11.

The 0-2 and 2-8 hour atmospheric diffusion factorsE given in Appendix 3

15A, and the 0-8 hour breathing rate of 3.47 x 10-m /sec are applicable.

12.

Dose model used to evaluate the environmental coasequence of this accident is given in Appendix 15A.

Steam releases to the atmosphere for the loss of offsite power are given by Table 15.2-5.

Using the previously listed assumptions, isotopic releases to the environment are summarized by Tables 15.2-7 and 15.2-8 for realistic and conservative assumptions, respectively.

,9 7.15X oo*

l. 'l0 X 40~0

/* 19 X I5' 019 A40 Camma, beta, and thyroid oses in ths first two hours of the loss of offsite power o elant auxiliari for_the realistic analysis at the site boundary are 7 23 :--19-9) rem, (M ; 10-9) rem and (9.09 r 1^-Drem. resoectively. The co esponding doses at the Low population zone are(4TM-s-W9 eem,(4x44-m _

4 remand 600.10-9eemrespectively.

1

(,,4,g3zfg G,c, gg,;i G 3.58 x d8 d

l.30 A to*3 (2.87xso amma, beta, and thyroid doses (iin the first twoThours of the loss of The l

offs te over to olant auxiliaries for the conservative analysis at the site boundary are 9T63-x-W9 tem,(2.0? x 10-3)res and(2." :--10-3 rem, respectively. Cor sponding doses at the low populacion r.one are $ r2 W

,-),0 *+) r e m, (%rH-s-1 rem and(4r82-.r-te*Drem, respectively, for the] duration oftheaccident.(3,3 p

L,, y g g - 3

( f,33 y,g Y The doses for this accident are well within the limits defined in 10 CFR 100 (25 Rem, whole body and 300 Rem, thyroid).

15.2.10 EICESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS 15.2.10.1 Identification of causes and Accident Description Addition of excessive feedwater causes an increase in core power by decreasing reactor coolant tempnrature.

Such transients are attenuated by i

the thermal capacity of the secondary plant and of the reactor coolant system (RCS). The overpower - overtemperature protection (neutron overpower, overtemperature and overpower AT trips) prevents any power increase which could lead to a DNBR less than 1.30.

15.2-33 r

TABLE 15.2-5 PARAMETERS USED IN LOSS OF OFFSITE POWER ANALYSIS Reali s tic Analysi s Conse rva t ive Analysi s core thermal power 29C P.W t 2900 MWt Steam generator tube leak rate prior to and during accident 100 lbs/ day (l) 1.0 gpm Fuel defects 0.12"ll) 1%

Iodine partition factor in steam generators prior to and during accident 0.01 0.01 l

Blowdown rate per steam generator prior to accident

,ht"gpm 15 gpm.

91 Ouration of plant cooldown by serandary system after scrident 3 hr 8 hr Steam release from three steam generators 447,900 lbs (0-2 hr) 629,300 lbs (0-2 hr) 757,700 lbs (2-3 hr) 757,700 lbs (2-8 hr)

Feedwater flow to three steam generators 629,300 lbs (0-2 hr) 429,300 lbs (0-2 hr) 341,300 lbs (2-3 hr) 341,300 lbs (2-3 hr)

Meteorology Annaa! average Accident (1) A.terican National Sta.9dards Institute, "Source Term Specification," ANSI N237, Revision 2.

~

15.2-55

f TABLE 15.2-6 4

SECONDARY SYSTEM EQUILIBRIUM CONCENTRATION (1) i i

Isotopes Loco.01 mom (2)

L,.0,1 g,,,

to.1,o,,.

2.L.S 2.GS I-131 '2.6 6 2T41x10-1 pCi/lb J A fx100 pCi/lb J At*x 101 pcl/lb 52+

5. e.'1 I-132 6.hf 1<99'ul0-2 Jag"x10-1

,la y, 100 2.')5 2.'15 i

I-133 2.96 JATx10-1 J<t71i100 1

I 4.77 4, g, 10

.77 I

gg y a to-1 I-134 6 77 M x10-3 FM110-2

,9 41 j

7 41 I

7.t/ K x10-/ 4 JAt:10k l jar a 10W I-135 i

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i (1) Using primary coolant parameters and activities in Tables 11.1-1 l

and 11.1-2.

1 (2) Lp a primary to secondary leakage rate.

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TABLE 15.2-7 LOSS OF OFFSITE POWER ACCIDENT ISOTOPIC RELEASE TO ENVIRONMENT REALISTIC ANALYSIS Activity Released te Environment by Accident (Cl)

Isotone (0-2 hr},

(2-8 hr) 1-131 4.80 4v44 x 10-5 7.91345' x 10-5 I-132 2.93 4149 x 10-5

'g.?2 y g ( x 1o-5 I-133

7. 2 5 -hw. x 10-5 1.23 jlAf x 10-/4 I-134 4.74 4r39 x 10-6

/.13 J.As'x 10-[6 I-135 3.44.2Ae x 10-5 6J$ g x 10-5 Xe-131m

/.00 64t'x10-/5 3.00 peg

x to-/3

/./8.1<93'x10-/f 3.555as'*x 10-/l Xe-133 Xe-133m 7.72 AM*x 10-[3 2.32 idf*x 10-[2 Xe-135 3.64 A 46*x 10-3

/.09,3d3'x10~/E Xe-135m 2.50 L 47'x 10'f4 7.50,1.<W x 10-4 Xe-137 0

1 79"x M C,1.ct1t* x g 4

Xe-138

3. 8 L e1 x 10~4 9.78J<615'x 10-4 Kr-83m O

Jur7' x /

O,,2<41 x /

Kr-85 4.JY.2dC x 10"f3

/,24J<$Tx10~/2 4

Kr-85m

9. /0.Ade x 10-4 2.73JJ5* x 10-3 Kr-87 6.99.2A t x 10-4

/.63,Idf x 10"[3 j

Kr-88

/. 6 3.249f* x 10'[3 4.96,248'x 10-3 Kr-89 0

h 11 xW 0.bdt* x W J

t a

g.

15.2-57 s

1 TABT.E 15.2-8 LOSS OF' 0FFSITE POWER ACCIDENT ISOTOPIC RELEASE TO ENVIRONMENT CONSERVATIVE ANALYSIS (1)

Activity Released to Environment by Accident (Cl)

Isotooe

[q-J. h r )

(2-8 hr)

I-131 l.ol %49 x 10-Il I.3g M x 10-1 I-132 MS M x 10-2 j,oo+:99xto-2 6

I-133 j.17 h::st x 10-1 1.65 W x 10-1 I-134 3.%7 W x 10-3

$3f b-1FJ x 10-73 I-135 412. ta9a x 10-2 6.38W Y 10-2 Xe-131m

~7.Qff 9:::::8 x 10-1 2.M M x 100

{

Xe-133 8 6 3.,ses c ic/ l M,cs x 102 Xe-133m

$.63 W x 100 l.61 +r*9 x 108 l Xe-135 26I W x 100 1,97,,:,,,44 x to/C Xe-135m 19ATert" x 10-1 d.5o e x 10-1 Xe-138 2 39 W x 10-1 1.20 W x 10/-I Kr-83m Kr-35 3.0'l ht? x 100 9,of ;,,+3. x to/o

/

10 'I l.19 W x 100 Kr-85m

( 63 h+6 x Kr-87 3.99 ST7t7 x 10-1

l. lie x 100 Kr-88 1.19 W** x 100 337w x 100 Kr-89 I

(1)

Primary to secondary leakage = 1.0 g;m f

9 15.2-$8 1

s; 3.

The iodine partition factor for activity released from the break is 0.1.

r 4.

The concentration of radioactive nuclides in the reactor coolant is listed in Table 11.1-2 for the conservative case and in Table 11.1-5 for the realistic c.ase.

[

I Using the previously listed assumptions, isotopic releases to the environment are determined to be those listed in Tables 15.3-6 and 15.3-7 1

for the realistic and conservacit*e cases, respectively.

Cach7% Yao#i.3G u dand thyroRea, to.

1. le,, Ren and 5.^1C G ores.

ectively. T' D'dVda 2

3.06 /Id' 177 h6 beta the sit undary for the realistic case i

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_g net i

W C

sponding doses aqhe low population zone are G.20. le Rea, N.?" _ r Ren and(4.63 = i_e y Rea,,,respectively.

C 5.07 g id '

j

.m ii' WZN

1. 2 ud*

r-~2 (o x i cd Camma eta and tii old dose [at tSa site boubdary for the conservative r

j case areiT!!~.

Rea,[T.46

_1G-$ Rem and !&. 72 IG-1) Rea, r f.ff vid l

low population zone arell e j

r yectively. Corrensonding doses at tha

-t9"81 Ree, p.11. ;;-E Ren andIO.00.10-M Rea, respectively.

1 A-- 416 t oc@

L <pyff y,gl Doses resulting from this accident are_well within the limits defined in 10 l

CFR (25 Rea whole body and 300 Rea thyroid).

Ivo

]

15.3.8 RRFRRENCR$

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Esposito, V.

J., Kesavan, K. and Maul B.

A., "WFLASH - A FORTRAN-IV l

Computer Program for Simulation of Transients in a Multi-Loop PWR,"

1 WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 1

(Non-Proprietary), July, 1974.

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Forsching, T. A., Murphy, J. H., Redfield, J. A., and Davis V. C.,

"FLASH-41 A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients in a Reactor Plant," WAPD-TM-848 Bettis I

Atomic Power Laboratory, March, 1969.

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3.

Bordelon, F. M., et al., "LOCTA-IV Progrant Loss of Coolant j

Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305

(

(Non-Proprietary), June, 1974 l

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Hellman, J. M., "Fuel Densification Experimental Results and Model j

i for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP-8219-A

{

(Non-Proprietary), March, 1975.

s a

l I

S.

Altamore, S. and Barry, R.

F., "The TURTI.E 24.0 Diffusion Depletion

{

Code," WCAP-7213-P-A (Proprietary) and WCAP-7758-A (Non-Proprietary),

l January,.1975.

j 6.

Barry, R.

F., "LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," WCAP-3269-26, September, 1963.

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15.3-14 3

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TA3LE 15.3-6 CREMICAL AND VOLUNE CONTROL SYSTEM LETDOWN LINE RUPTURE - ISOTOPIC RELIASE.

TO THE ENVIRONMENT - REALISTIC CA! E 2

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Activity Released Isotope (Cl)

)

I-131 g,7 3 2.46 x 10-1 I-132 gg,g,g trts x 10-1 I l

j I-133 y,5"7 }<46 x 10-1 I-134 c,, G,7 6;*1 x 10-2 I-135 g,g g Frf4 x 10-1 i

4 Xe-131m 15 1 6 tr*9 x 10'b o 4

I Xe-133 z, r si 4,4+ x 104 7-Xe-133m l,09 Set 6 x 1041 j

Xe-135

-3,q g 1,47 x 100 Xe-135m er,qq kre4 x 10-1 9

-~...

j Xe-138 71G kit 4 x 10-1 i --

Kr-85 9.oG ir&5 x 1040 j

Kr-85m i,act 9 t2 x 1040 j

Kr-87 1.L9 icF7 x 10'k O l

0 Kr-88 qs,gqg 1,74 x 10,

Er-69 C.ee 10 -

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O TABLE 15.3-7 CHEMICAL AND VOLUME CONTROL SYSTEM LETDOWN LINE RUPTURE - ISOTCPIC RELEASE TO THE ENVIRONMENT - CONSERVATIVE ~ CASE Activity Released Isotope (Ci)

I-131 7,.$7 w x too I-132 T 4 0

'*,-M x 104O I-133 3, gj h*6 x 100 I-134 76 /*res x 10-1 I-135 7, ; g g g x too Xe-131m Ii P2. h46 x 101 Xe-133 7.IG M 4 x 103 Xe-133m g,q g 2,4g 3 gg L Xe-135 6.M W x 1010 Xe-135m y.f(, arm x 10 Xe-138 G x 100 Kr-85

-7 SI h99 x 101

#-!,5 :'

l. % 1.S.1 x 101 l,'_-};

e.er wee x 10%o 1

a, c, g 1,44 x 10 I

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13.3-23 j

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will be bypassed around the Control Room Emergency Filter Plenum.

For the emergency mode of operation, the control room recirculation air flow will be routed through the Control Room Emergency Filter Plenum.

In the purge mode of operation, the system will supply 100 percent outside air to the control room.

The purge air inlet cover plate i

will be removed from the outside air intake plenum and the system l

relief dampers to allow an outside air flow of 21,270 cfm into the control room. The recirculation flow will be terminated for the duration of the purge mode.

Each control room air intake is provided with two isolation valves in series. One of the two valves restricts the outside air flow to a maximum of 1000 cfm flow for both normal and emergency modes. Upon receipt of an f

engineered safety features actuation signal, the control room ventilation system switches to the emergency mode.

For the purpose of this analysis, the maximum allowable air intake value j

was determined for the limiting total integrated dose to control room t

personnel. The results are presented in Table 15.4-18.

The maximum allowable air intake value was determined to be efm. The system operating flow of 1000 cfm provides adequate marg a for the protectir,n of control room personnel as specified under Genera Design Criterion 19 of 10 CFR 50, Appendix A.

I 2423 15.4.2 MAJOR SECONDARY SYSTEM PIPE RUPTURE 15.4.2.1 Ma for tuoture of a Main Steam Line I

15.4.2.1.1 Identificacion of Causes and Accident Description The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure.

In the l

presence of a negative moderator temperature coefficient, the cooldown

(

results in a reduction of core shutdown margin.

If the most reactive rot i

cluster control assembly (RCCA) is assumed stuck in its fully withdrawn positiori after roactor trip, there is an increased possibility that the core will become critical and retarn to power. A return to power tollowing I

l a steam line rupture is a potential problem mainly because of the high l

power peaking factors which exist assuming the most reactive RCCA to be i

stuck in its fully withdrawn position.

The core is ultimately shut down by the boric acid injection delivered by the safety injection system.

The limiting main steam line break was selected based upon the sensitivity l

studies performed in "Reactor Core Response rn Ru essive Sacondary Steam Releases," WCAP-9226, January, 1978.

The analysis of a main steam line rupture is performed to demonstrate that the following criterion is satisfied 15.4-14

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TABLE 15.4-10 tt RADI0 ACTIVITY RELEASE FROM THE RECIRCULATION LOOPS 1

(Emergency Core Cooling'and Reactor Building Spray Systems)

I d

Release Rate Isotope curies / min

[

i I-131 1 9 M x 10-2 t

j

~

I-132 Z.7 2 4 x 10-2 j.

I-133 3.S-J A x 10-2 i

I-134 4.l M x 10-2

)

I-135 3.5 x 10-2 1

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This table is based on 50 percent of core iodine inventory in the sump, i

71,454 ft3 of water in the sump, 5860 cc/Sr leakage and a 100 decontamination factor of iodine between Liquid and airborne phases.

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TABLE 15.4-11 PRIMARY COOLANT INVENTORY Activity Isotope (Cl) i 1-131 5".29 4,+f x 102 i

I-132 r.4 S 1,44 x 102 I-133

$ f,1 3,44 x 102

} n I-134 i.7-t kne x 102 I-135 4 41 4rH x 102 i

Kr-85 t.9 % krt1 x 103 f -

Kr-87 21,7 2A6 x 102 Kr-85m 1.76 4.46 x 102 i'

Kr-88 c go 3 r+f x 102 j

Xe-131m

q. i 6 4,44 x 102 p

Xe-133

( 41 5,40 x 104 Xe-133m 3.z.t 2 46 x 10b3

]

Xe-135 f,s-t LA9 x 103 Xe-135m t.oy 9,45 x 10% E-j Xe-138 g,3 L,44 x 102 j

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TABLE 15.4-12 FISSION PRODUCT ACTIVITY AIRBORNE IN CONTAINMENT REALISTIC CASE - LOSS OF COOLANT ACCIDENT Activity

_ Cl)

(

Isotope I-131 2.G g M 1 x 102 1 2.

I-132 g,;q &,40 x 10 I-133 43 / 1,44 x 102 I-134 c.3f Me x 101 I-135

2. 9 5~ 2d x 102 3

Kr-85 t. -r 2. W x 10 2

Kr-85m 3,, s I.,-H x 10 Kr-87

z. 2.7 M 6 x 102

~

Xr-88

's.80 7.-M x 102 l

Xe-131m 44 16 M 4 x 102 4

l Xe-131 w.9i 4 46 x 10 l

Xe-133m S,2, M&y 1043 i

Xe-135

,,fl 1,49 x 103 l

1 Xe-135m g,oy 4,4 x 10i 2.

2 Xe-138

,,g u l x 10 l

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O TABLE 15.4-13 FISSION PRODUCT ACTIVITY AIRBORNE IN CONTAINMENT CONSERVATIVE CASE - LOSS OF COOLANT ACCIDENT Activity Isotope (Cl)

I-131 5".'i G h ee x 105 I-132

f. 4 0 6,44 x ' 10' I-133 3 6 3 4,44 x 105 I-134 t o ( 6,45 x 10' I-135 t.9 o Mrf x 105 Kr-83m t.1 1 h*4 x 10' Kr-85 i.% L 40 x 105 Kr-85m 4.t o h +1 x 10' Kr-87 4.o7 5,44 x 10' Kr-88 81 01 h M x 10M Kr-89 f. 5"O h44 x 10' Xe-131m 9.Ib 3,46 x 101 Xe-133 I.G( h*4 x 106 Xe-133m t. r o 2 M x 10* #

Xe-135 1.o t kd f x 105 Xe-135m 1.sl' L.M x 10* I Xe-138

-7.o t L49 x 10* I 15.4-18 a.

e TABLE 15.4-14 FISSION PRODUCT ACTIVITY AIRBORNE IN CONTAINMENT REGULATORY CUIDE 1.4 CASE - LOSS OF COOLANT ACCIDENT Activity Isotooe (Ci)

I-131

1. V 1-r79 x 107 I-132 t.G h+3 x 107 I-133 3.e( 4v04 x 107 I-134 9 2 f Arte x 10

I-135 3.65 x 107 t.1 h+3 x 10%G Kr-33m d

Kr-85 c,.4 ara x 105 Kr-C5m 4.0 1r20 x 10 7 Kr-87 3,1 6rt? x 107 Kr-83

' r.3 bJ4 x 107 Kr-89 r,,f L ;1 x 10D7 Xe-131m f.4 h*3 x 105 Xe-133 a.f k,45 x 103 Xe-133m

z. 3 4r40 x loi ~I Xe-135 3,3 4 rM x 107 Xe-135m

?.i Gya x 107 Xe-138 g,3 L,J.4 x 108 I

15.4-79

TAliLE 15.4-15 (Cont inued)

PARAMETERS USED TO EVALUATE OFFSITE DOSES FOR TIIE LOSS OF COOLANT ACCIDENT

[

Regulatory Cuide Iaran.eter He.nlistic Analysis Conservative Analysis, 1.4 Analysis Forse o f todine Activity in Containment Ava i l atil e for Release Elemental lodine 91%

911 91%

Organic lodine 42 41 4I Particulate lodine SI 51 5%

flumpe r of Spray 1* umps Operating I of 2 1 of 2 1 of 2 Spray Femov.sl Coefficient for Iodine u

"a -

Elemental 12.55 hr-I 12.55 hr-I 12.55 hr-I h

Particulate 0.507 hr-I 0.507 hr-I 0.507 hr-I Elfcetive Decoutamination Factor of Spray on Etex ntal lodine 100 100 100 Containment Free volume 1.84 x 106 ft3 1.84 x 106 gt3 1.84 x 106 ft3 Containment Leak Rate 0.2% per day (0-24 hr) 0.2I per day (0.24 hr) 0.2% par day (0-24 hr) 0.11 per day (1-30 days) 0.1% per day (1-30 days) 0.1% per day (1-30 days)

Containment Recirculation fq,2 o<>

f 4,2.o d SN, zoo Flow 10,770 cfm 64 rNO c fm 64 rN'O c f m s

TABLE 15.4-16 0FFSITE DJO E,5 FROM LOSS OF COOLANT ACCIDENT Thyroid Dese (Rem) j Site Boundary Low Population. Zone (0-2 hours)

(0-30 days) 1609 meters 4827 meters Realistic Analy-is i. 9 Eb avdS x 10-3 3.o0 1<61 x 10-4 Conservative Analysis 7 67 Ldt x 100 f,og 1.!HS x to-1

)

Regulatory Guids 1.4 Analysis f,19 LM x 10 Z 69 140 x 101 2

10 CFR 100 Cuidelines 300 300 j

Canea and Beta Doses (Rem)

Site Boundary Low Population Zone (0-2 hours)

(0-30 days) 1609 meters 4827 meters Beta Beta Camma Skin C4mma Skin Dose Dose Dose Dose Realistic Analysis f.f&tT x 10-5 Mx to-4

'3 x to-5 d x 10-5 Conservative Analysis Gr?1 x 10~3 Sd5 x 10-3 Ldd x 10-3 L.M x 10-3 G.60 S. 3g scs gg,9g, Regulatory Guide 1.4 i

Analysi:

!*,45 x 100 24t7 x 100 L e x 10-1 f.,44 x 10-1 10 CFR 100 Guidelines 2 78

$ ( 1 )1.i b 3.r 7 25(1) 4 I

(1) 'Whole bory dose.

D 15.4-33

TABLE 15.4-17 PARAMETERS USED IN ANALYSIS OF CONTROL ROOM DOSE FOLLOWING A LOSS OF COOLANT ACCIDENT Parameters Control Room Free Volume 226,040 ft3 19,143 Filtered Recirculation Flow 2*y446 efm Recirculation Filter Efficiencies 95% for all species of iodine Maximum Control Room Filterec Air 2413 Infiltration Rate

.1790 cfm*

i Contro' doom Unfiltered Air Infiltra-tion Rate 10 cfm z 42. 3 Maximum Control Rcom Outleakage Equal to total inleaka3e (11&& cfm)

Meteorology 0-8 hrs:

9.35 x 10-' sec/m3 8-24 hrst 6.63 x 10-' sec/m3 i

1-4 days:

3.95 x 10-4 sec/m3 4-30 eys:

2.45 x 10-' sec/m3 Percent of Time Operator Is in 0 - 24 hrs 100 Control Room Following Accident 1 - 4 days 60:

4 - 30 days 40%

Duration of Accident 30 days Breathing Rate of Operators in 3

Control Room 3.47 x 10-' m /see Activity aelease Assumptiors Table 15.4-15 Method of Dose Calculation Appendix 15A

Actual system capacity 2500 cfn (single train)

I 15.4-84 3

O TABLE 15.4-18 CONTROL ROOM DOSES FOLL0t/ING A LOCA Doses (Red Thyroid Cama Beta Skin

7.

b 7. Z-

?, i' Realistic Analysis M x 10-4 AA T x 10-4 M I x 10-4

'0-1 b'

x 10-3 M x 10-2 Conservative Analysis x

Ultra-Conservative 3o g,.7 4,3 Analysis 3.08, x 101

. M 6 x 100 Mx 100 4

4 l

i 1

i l

13.4-85

8.

No condenser air removal system release and no steam generator blowdown during the accident.

9.

No noble gas is dissolved in the steam generator water.

'O.

The iodine partition factor in the unfaulted steam generators, 0.01, is determined as follows amount of iodine / unit mass steam amount of iodine / unit mass liquid 11.

During the postulated accident, iodine carryover frou the primary side in the two unfaulted steam generators is diluted in the incoming feedwater.

12.

In the faulted steam generator, all water boils off and is released through the break immediately after the accident. The partition factor for iodine released is assumed to be 1.0.

After this initial release, further iodine is released due to primary to secondary leakage in the faulted steam generator. A partition factor of 1.0 is also assumed for this release.

13.

The primary pressure remains constant at 2235 psig for 0-2 hours and then decreases linearly to atmospheric during the period 2-8 hours.

14.

The 0-2 hour and 2-8 hour atmospheric diffusion factors given in Appendix 15A and the 0-8 hour breaching race of 3.47 x 10-4 m3/sec are used.

15.

The dose model used to evaluate this accident ls given in Appendix 15A.

Steam releases to the atmoCp/.M-25 here in the first two hours of the steam line i

breakaregiveninTableQ5.b1T) Isotopic releases to the environment I

using these assumptions are sumarized by Tables 15.4-24 through 15.4-27.

g

b 2.05 f f 1.l%E 2.$ % 0 r the s eam line(\\ break accident, base The gama, beta and thyroid doses upon the realistic analysis, are 4.%. W Rem,0 47 s193) Rem and

-e-19,$ Rem, respectively@. _ at7t r-10"3' Rea, $5-1r-40$ Rem and 8T1 the site boundary. Correspondin

.q1sta he low population zone are Rem, respectively.

(,, q, q

-7 L q37 g7

( lo. 87 A 10

~

Carma, beta and thyroid doses at the site boundary and at the low population zone for the steam line break accident, based upon the conservative analysis, are given by Figures 15.4-61 through 15.4-63 as a 1

l function of primary to secondary leak rate. The doses resulting from this 1

accident are well within the limits defined by 10 CFR 100 (25 Rem, whole body; 300 Rem, thyroid) for the range of credible steam generator tube i

leakage.

15.4-22

TAIiLE 15.4-23 (Cont inued)

PARAMETERS USED IN STEAM LINE BREAK ANALYSES l'arameter 2ealistic Analysis Conservative Analysis I!owdown rate per steam generator prior to accident 4/gpm 15 gpm Initial steam and water release from faulted steam generator 165,000 lb (0-30 minutes) 165,000 lb (0-30 minutes)

Imng term steam release from faulted stean generator 12 lb (0-8 hours) 1,300 lb (0-8 hours)

Steam release from two unfaulted 332,700 lb (0-2 hours) 332,700 lb (0-2 hours) steam generators 665,400 lb (2-8 hours) 665,400 lb (2-8 hours)

C Feedwater flow to two unfaulted 453,900 lb (0-2 hours) 453,900 lb (0-2 hours) 4, -

steam generators 721,500 lb (2-8 hours) 721,500 lb (2-8 hours) e

~

Heteorology Annual average Accident l

l l

{

l

1 TABLE 15.4-24 STEAM LINE BREAX ISOTOPIC RELEASE TO ENVIRONMENT CONSERVATIVE ANALYSIS (I)

Activity Released to Environment by Accident (Cl)

J Isotone (0-2 hr_1 (2-8 hr)

I-131 f.90.AAr x 10-/'

/. 37.tae x 10-I o I-132 7.95 m x 10-74 2./4,6atr x 10-I' I-133 3.5J.5,41 x 10-f' 9./6.2af x 10-/8 I-134 6.fo.lar x 10-14 2.oV2ar*x 10-18 I-135

/.?4 245"*x 10-Il 4.fo L4e* x 10-/I Xe-131m 3.87.b W x 10 '3

/./6.htf x 10-2 Xe-133 6.95.1.Af x 10

2,et 447 x 100 8

Xe-133m 7.15.LAT x 10-2 2 jV 1Ac x 10-/l Xe-135 4.t33A9'x 10-2 l.21.lA 6 x 10-1 Xe-135m 6.772df x 10-52 58 2.0.%,b47 x 10 /8 Xe-138 2.96.h44 x 10-78 f.F6.L A f x 10-Kr-83m S.34.QA x 10-3

/.60 art A /o-A Kr-85 5.77,.L+r x 10-2

/,7 s pty :t to-1 Kr-85m 2

i

/.76.144 x 10 /2 5.29.'LAr x 10-2 Kr-87

/.7 /.541 x 10~

2 5./Sj.Afx10p Kr-88 3.7f LM" x 10-2

/./3 3ag x to-Kr-89 6.31 S # x jo*3

/.10 A 4 x to-2 1

i j

i l

4 l

i Il Primary to Secondary Leakage = 0.01 spm.

i 15.4-92 i

l TABLE 15.4-25 STEAM LINE BREAR ISOTOPIC RELEASE TO ENVIRONMENT CONSERVATIVE ANALYSIS (l)

Activity Released to Environment by Accident (Cl)

Isotoee (0-2 hr)

(2-8 hr) i I-131 4.9 o k+3'x10YO

/.37 kts x 10d% '

I-132 7.93 / # x 10'/l 9, /4 445 x log /0 1,.lf.,Lat x 10-I-133 J.55 L 47 x 10-f0 o

I-134 6.9o.br$r x 10'/l

/O 2.cV 2 +t x 10'/O I-135

/. 76 3,47 x 10 10

/

4,pp. tag x 10~

Xe-131m 3.T7 Aet x 10-2 1.16 A t3 x 10~1 Xe-133

(,,9f.1. M x 10/0 2.oTLtf x 101 Xe-133m

?,/f.t At x 10-1 2.N L M x 1060 Xe-135 1

4.28.341 x 10 $l

/.29 ketr x 100 Xe-135m to.17 h tt x 10*

2.0S ktt x 10*$0 Xe-138 2.16 A44' x 10~/O f.F6 L49' x 10~/O l

Kr-83m KJV At x so-*

/.60 At xto~'

Kr-85 S.71 L W x 10'1

/.7/j,M x 100 Kr-85m

/.76 Leer x 10-1 5.19 3.M x 10'*l Kr-87

/.7 / Att x 10-/l 5./f htT x 10-1 4

i Kr-88 3.71 Lw x 10-1

/./3 L97 x 10~/O o

Kr-89 G.3l Dd x to*%

/.90k6 g 40*l J

L

(

4 i

\\

l Primary to Secondary Leakage = 0.1 gpm.

I j

15.4-93 1

s

O TABLE 15.4-26 STEAM LINE BREAK ISOTOPIC RELEASE TO ENVIRONMENT CONSERVATIVE ANALYSIS (l)

Activity Released to Environment by Accident (Cl)

Isotope (0-2 hr)

(2-8 hr)

I-131 4.9c A,+S* x 10 +'

l.37 L44 x 10b2 f

I-132 7.9 3./*dt x 10&O 2.12 Ade x 10d+8 F

I-133 3.59.b4t x 10 +1 9./6 Let* x 10F+8 I-134 6.9o.h er x 10/#0 2o4 Att x 10&+8 I-135

/.76 A 4t x 10ff+1 4 r o.L A e x 10 + i j

f Xe-131m J.P7 Aef x 10-1

/./6 2d1 x 100 r

Xe-133 k.15 A4f x 10/+1 2.cf.44r x 102 i

Xe-133m 7./5.144t' x 100 y,jg L44 x toF+1 Xe-135 4 28.149 x 100 f,29 L e x 101 i

Xe-135m 6.77 AM x 10&O 2.o) 6,et* x 106+ 1 Xe-138 1.96 A 44 x 10 &+1 f.76 L49' x 107+1 1

Kr-83m S.3V A 4 X 10

/.60 A4 4 /O' 1

Kr-85 5.72. 3dt x 100 f,7f,3,a.yx 101 Kr-85n

/.76 het x 100 5.29.3d$~ x 100 Kr-87

/.7/.htt x 10&O 5./5.1 dt x 100

)

Kr-88 3.79.ht7 x 100

/,f3 5 37, loF+i Kr-89

b. 31 M x 10-!

/.90 M X to*

I i

i 1

i i

i I

1 l

(1) Primary to Secondary Leakage = 1.0 gpm.

l 4

i l

i 15.4-94 1

TABLE 15.4-27 STEAM LINE BRFC< ISOTOPIC RELEASE TO E!NIRCNMENT REALISTIC ANALYSIS Activity Released to Environment bv Accicent (Ci)

Isotoee (0-2 hr)

(2-3 hr)

I-13; 263 hie'x 10-3 t.~43 1rH x 10-3 I-132 5 2)W x 10-3

t. % 4 W x 10- 0 I-133 3.rq 2,-H 10-3 2.t61rH x 10-3 I-134 I Af ht) x 10-'

g.12. 2 79 x 10-4 I-135 i.so 1-+t x 10-3 i,z3 8 ?S x 10~/7 Xe-131m too 6720 x 10-I3 3 co* 3 r44 x 10-A,7 Xe-133 1 4 1-rH x 10-2I 3,$F ' fJ x 10-Il A

Xe-133m 713 Me x 10-9 3

7. 30 1,-M x 10-) ?-

Xe-135

2. 31-tt x 10-3 3 1.c3 Mt x 10-8 7-Xe-135m 1.49 h++ x 10-I 4 7.41 MC. x 10-'

Xe-137 o

2.J3 &.G-5 o

W,.;*

Xe-138 3 27 I W x 10-4 9.60 5 Ar8 x 10-'

K r-8 'ka o 0.07

.s

.0 7 o

2.H.

.0-"

Kr-85 4 6 2 !T x 10-I 3 114 J-tro x 10-52.

Kr-35m 9,ee 4-tt x 10-4 2 73145 x 10-3 Kr-37 5'44 W x 10-'

t.Ie3 7erts x 10-/3 Kr-88 1.s.3 ht* x 10~/3 4.40 3, 8 x 10-3 Kr-89 0 29 'A,.0-5 o C;4 c-1 15.4-95

30' i

i i i i iil l

I I I l 1

I l

p p

q 46 9

f ffgy

/

,Y O

j 10-3 Tp/

py o,s/

I

/

fpf7

/

3n-4

/

\\

/

Z

{

/'

~

j i

/

8

_/

(

/

J

/

l j

j i

ig-5 I

l/,

I I l!!

l I

i 1 ! ! II t

l i

10-2 10-1 100 PRIMARY TO SECONDARY LEAKAGE (GPM)

SOUTH CAROLINA ELECTRIC & GAS CO.

VIRGIL C. SUMMER NUCLEAR STATION Steam Line Break Whole Body i

Gamma Dose 00dbE2)AToJS OA3G Figure 15.4 61 1

o 10' 1

I I I Iiii I

i i i iIii i

1 2

~

b

+4 Of #

O'}f 10-3

/

0y 0'/

fBAk 3

=

I l

/

\\

10-4 Y

[

l y

l i

10-5 l

1 1 1111 i

f f I i11I I

I l

10-2 30 1 gno t

3 PRIMARY TO $ECONDARY LEAKAGE (GPM) i i

L SOUTH CAROLINA ELECTRIC & GAS CO.

i VIRGIL C. SUMMER NUCLEAR STAT 10N Steam Line treak Surface Body Seta Dose i

Coaneurae Case Figure 15.4 62 l

l l

l t

m 100 i

i i

iiii 1

i I

IIf I I

i V-2 HR.

$1TE BOUNDARY

~

0 8HR.LPL ~

/

i

/

)

/

1 l

lo-'

/

m i

~

/

I g

7 a

/

}

[

/

l 10 - 2 t

l

/

2

~

j

/

i

~

l

~

l

/

t

,/

[

i 1

l J

4 l

in-3 l

I I l IlII I

I l.

II III I

I l

l 10-2 10-1 100 i

j PRIMARY TO SECONDARY LEAKAGE (GPM)

\\

l i

SOUTH CAROUNA ELECTRIC & GA5 CO.

VIRGIL C. 5UMMER NUCLEAR 5TATION 1

Steam Line Break Thyroid Dose j

CodsCKYAToM CAs2 1

\\

f Figure 15.4 43 i

4 4

l i

The following information was required for the Steam Cenerator Tube Rupture Accident dose calculationi t

1.

The approximate mass of metal in contact with the RCS water is 1.33 x l

106 lb.

2.

The secondary side volume in one steam generator is $947 ft3 3.

The air ejector flow rate is I

i Main condenser:

576 lb/hr, air / vapor mixture I

)

Auxiliary condensers:

215 lb/hr, air / vapor mixture l

s 4.

The normal operation letdown rate is 60 gpm.

1

(

5.

The total RCS volume at hot conditions is 9410 ft3, including the j

pressurizer. The liquid volume in the RCS ddring normal operation is 8850 ft3 at hot conditions.

d 1

I a

i 6.

The volume fraction of liquid in the steam generators at normal j

operacion is 0.3.

l 7.

Emergency feedvater systen initiation time is less than 1, minute.

The flow rate to each steam generator is 190 spm.

i i

8.

Figure 15.4-83 shows the liquid volume fraction in the faulted and i

non-faulted steam generat' ors as a function of time after the tube I

3 rupture.

j I

i 9.

Pressure as a function of time in the primary system and in the

(

faulted an3 non-faulted steam generators is shown in Figure 15.4-84 i

The steam releases to the atmosphere for the postulated steams generator i

tube rupture are given in Table 15.4-29.

Isotopic releases to the environment based upon these assumptions are summarized in Tables 15.4-30 i

through 15.4-33.

s,196-4 2.cc, G-4 i

i

[

Cam.4, beta and thyraid doses k the site bound ry in the first two hours

(

1 of the postulated steam gener or tube ruptura ceident and ba " ~ " the 1,ME-4 6

I realistic analysis are G C 50 9 Rem,Cik...ej Rem and h IC-I)*

I Re respectively. Corresponding doses it the Low population zone art

(

)

Rem,C Z

_ ie7 Rem andc"~", ;_i-) Rem, respectively, for the 1

d ratton of the a cident.

N I

a.p d-r I.Zi6-5,

i j 1,0 go*(

Carr.a. beta and thyroid doses at the site boundary and low population tone

[

resulting from the postulated steam generator tube rupture accident and i

based upon the conservative analysis as a function of primary to sacer.Jary I

leak rate are given by Figures 15.4-ti through 15.4-68.

The doses from this accident are well within the limits defined in 10 CFR 100 (25 Rem, t

whole body, 300 Rem, thyroid) for the range of credible steam generator i

tube leakage.

l I

I a

i l

l 1

J 15.4-34

(

l

TABLE 15.4-29 PARAMETERS USED IN STEAM CENERATOR TUEE RUPTURE ANALYSES Realistic Analysis Conservstive Analysis Core thermal power 2900 MWt 2900 MWt Steam generator tube leak rate 100 lbs/ day (l) 0.01 to 1.0 gpm prior to and during accident Offsite pcwer Available Lost Fuel defects 0.12%(1) 1%

Iodino partition factors in 0.01 0.01 steam generators prior to and during accident 4/

Blowdown rate per steam at gpm 15 gpm generator prior to accident Time to isolate defective 30 min 30 min steam generator Duration of plant croldevn by 8 hr 8 he secondary system after accident steam release frei. defective 43,000 lbs 48,000 lbs steam generator (0-30 min)

(0-30 min)

Stes release fren 2 316,000 lbs (0-2 hr) 316,000 lbs (0-2 hr) unaffected steam generators 335,000 lbs (2-3 hr) 335,000 lbs (2-3 hr) feedwater flow to 2 346,000 lbs (0-2 hr) 346,000 lbs (0-2 hr) unaffected steam generators

$33,000 lbs (2-3 hr) 383,000 lbs (2-3 hr)

Reacter coolant released to 125,000 lbs 125,000 lbs the defective steam hip 7 generator Meteorology Annual average Accident

..c

,( 1 ) A?.crican National Standards Institute, "Source Term Specification,"

N237, Draft Revision 2.

15. -97

TABLE 13.4-30 STD.'i CENERATOR TUBE kUPTURE ISOTOPIC RE1. EASE TO ENVIRONMENT CONSERVATIVE ANAL'fst3(1)

Activity.'teleased to Environment bv Accident (Ci)

Isatoee (0-2 $d (2-8 hr)

I-131

'3. % M & x 10-1 I-132 2.7o 2. 4 2 x 10-2 3 49 5 44.x 10-1 2.e7 8. 56 x 10 I L I-133

66 hts x 10-1 I-134 4.16 3. 39 x 10-2 s

3.fo W x 10-2 4.8 2 5.30 x 10-3 I-135 3 31 r,M x 10-1 2.47 2.15 x 10-2 Xe-131m t.15 W x 10A Xe-133

,001,82 x 10-2 f.46 It-H x 10" 2 192.33 x 100 Xe-133m 9 f.6," cM x 102 g,gg.3.63 x lo-fi Xe-135

4. s.4 W x 102 7.27 7.17 x 10-2 Xe-135m 3 t t 2--t4 x 101 5.de 4. 54 x 10-3 Xe-133 4.c3 W x 101 Kr-5 3m -

6.54 7.2 6 x 10-3 Kr-85 s.ig W x 102 6.2 7 7. 5 4 x 10-2 Xr-85m t.a hto x 102 t.62 2.09 x 10-2 Kr-87 6 6)7.'? x 101 1 6 1.18 x 10-2 Xr-88 2..c4 h+1 x 102

.< r - 3 9 -

3.17 3. 5 4 x 10-2 i

i i

Primary to secensary leakage = 0.01 g;m.

l I

15.a-98 i

.s..

. o..

i

TABLE 15.4-31 STEAM CINERATOR TUBE RUPTURE ISOTOPIC RELEASE TO ENVIR0hhENT CON 3ERVATIVE_ ANALYSIS (l)

Ac:ivity Released to Environment bv Accidene (CL)

Isococe (0-2 hr)

(2-6 hr)

I-131 4.4 o *-rtt x 10- 1 c.1o ;i,+2 x 10-1 I-132 4 49 W x 10-1 t.<,7 S,-H x 10'11 I-133 7.19 M4 x 10-1

+ 36 A x 10~1 T.- 13 4 1.c).&vt$ x 10-21

,,,gs g + x to-2 I-135 4.M W x 10-1 2 42 2,+0 x 10-l Xe-131m hts-PTT7 x 102

2. co t-t! x 10-1 Xe-133 iAB M 7 x 10*

2.?6 Wp x 10 1 Xe-133m 9.65 2175 x 102 g,g w.g.x to+10 Xe-135 4.54 4 G x 102 g g.x to-1 7, g)

Xe-135m 3.12. 2,44 x 101

5,co ?,44.x 10-2 Xe-138 4.ci 4,44 x 101 G.5 t Mtr x 10-2 Kr-83m Kr-85

'5.86 w l x 102 g,27 g 4 x to-1 Kr-S$m 1.14 1790 x 102 t g2, 3.n39 x to-1 Kr-!?

' si 7 2 7

.v. 101 t.c3 W23 x 10-1 Kr-SS 7.04 2 rH x 102 3,g) wy:; x to-1 Kr-89 (1) Primary to secondary leakage = 0.1 gpe.

W 15.4-99

O TASI.I 15.4-32 STEAM CENER.ATOR TUEE RUPTURE ISOTOPIC RELEASE TO ENVIRCh?ENT CONSERVAT!VE ANALYS!S(l)

Activity Released t o E.a.vi ronxe.,t by Actident (C1)

Isocece (0-2 hr)

(2-3 hr) l I-131 t.2"7 b-H x 100 1.7o 3,+2 x 100 I-132 n.'t.5 ett x 10*$0

n S,.H x 10+l0 I-133 2.of IW x 100 4,g 399 x 100 I-134 2.6 ~7 2W x 10'l

r., l'i. 5 rM x 10'l I-135 1,14 ht x 100 1.4 2. 2,-M x 100

~

Xe-131m n:r. L+4 x 102 g,43 p9.x 100 4

Xe-133 i.41 W 104 2.24 2W x 102 Xe-133m 9 7o }v+1 x 102 g, n 3,+3, x to t e

Xe-135 4,g 4,40 x 102 7.27 77g x too l

Xe-135m 1,t 4 hts x 10 1 GC W x 10-1 l

Xe-138

4. t1 cts x 101 6.54 7-et x 10-l Kr-33m Xr-85 5.19 M x 102 p.27779 x ;go i

r.r-85m

[,gg hdt x 102 t 6 2. 2M x 100 x 101 1.ci 1 :-it x 100 Kr-37 4,3r 18r+i 2

l Kr-SS 2,og 2,41 x 10 S, t'7 3 rtt x 100

,j

(

Kr-89 l

l l

1 l

l l

1

.s I

(1) Primary to secondary leakage = 1.0 g;m.

l l

1 9

1 15. '. - L C O l

l

O o

TABLE 15.4-33 S T E.Ud CINER.ATOR TU3E RUPTURE ISOTCPIC RELEASE TO EWIRONMENT REALISTIC ANALYSIS Ac:ivi:y Released to Environmen:

by Accicent (Ci)

Isotoee (0-2 hr)

(2-8 hr)

I-131

.f.is-h +6 x 10-2 3,gi 7A 3 to-3 I-13' 412. 1 rw-3 x 10-2 3,4q g3 to-5 I-133 6 6 7 b+3 x 10 2

,,3 3, gar.,, t o-J V I-134 Y4 6,+F '

10~3

9. e kw+ x 10-6 I-135 Me

.W x 10-2 g33 4,44 x to-5 7 4

,44 x 10

d 3 Xe-131c s.63 M T2 x 1081 Xe-133 t.78 3ve9 x 1023 2.1"1 N M x 10-II Xe-133m f.!6 h +7.x 1047-i.52. h-?? x 10-d L 4e-135 S,4 W' x lotit 7,27 g-x to-3 Xe-135m Myw9x 10'f6

.t.9b W x 10-4

-I'

.0-I~

h !:7 W7

.0 Xe-138

4. To ard - x 100 g,33 gx to.4

-6 3 h

~Ms

.0-

. 73 4.-c-Kr-35 fe.24 k3t x 10 t l B.17 L.?;. x 10~4 3 Kr-35m I.34 O 1081 i.3 2. A44-x 10-4 }

Kr-37 s,g L 4t. x 100

,,39 4 79 x goa3

'*r-88

..is-h-t x 10 1 3 rpvH x 10~3 W

-5... -..sT e

i 9

l

!!.4-101 l

l l

i

100 g

g g g gggg g

g g i igg g

g

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t s,

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,nssA -AAs 0-2 HR.

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104 10-1 100 i

PRIMARY TO SECONDARY LEAKAGE (GPM) l l

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j SOUTH CAROUNA ELECTRIC & GAS CO.

[

VIRGIL C. SUMMER NUCLEAR STATION 1

Steeni Generator Tube Rupture Surface Body Gamma Dose couscev47WE Casa l

j Figure 15.4 64 1

i i

i

b 100 y

g g g ggg g

g g g g gg; h

~

e n 0-2 HR.

.A _n A _A A

_A r -

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-- yv- (f-yjy-y-

yyi

$1TE BOUNDARY s

L 10-1 i

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VIRGilC. SUMMER NUCLEAR STATION l

5 team Generator Tube Rupture Whole Body g Dose OcasKJAroJe Cipse r

i Figure 15.4 67 i

i

s

~

s o

100 I

I I I I Iii i

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i 0-2 HR.

~

$lTC BOUNDARY

~

0-8 HR. LPZ j

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1 s

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10-2 10-3 I ! lIfI I

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I 1

SOUTH CAROLINA ELECTRIC & GAS CO.

1 VIRGIL C. SUMMER 14UCLEAR STATION Steam Generator Tube A>apture

}

Thyroid Dose i

bAJ 5Edv4 77JE 0456 l

i Figure 15.4 68 I

l 4.

The iodine partition factor in the steam generators, 0.01, is determined as follows:

amount of iodine / unit mass steam l

amount of iodine / unit mass liquid 5.

No noble gas is dissolved or contained in the steam generator waters i.e., all noble gas leaked to the secendary system is continuously l

released with ster.E from the steam generators through the condenser air removal system.

6.

The blowdown rate from the steam generators is continuous at 15.0 spa per steam generator, The following conservative assumptions and parameters are used to calculate the activity releases and offsite doses for a steam line break 1.

Prior to the accident, an equilibrium activity of fission products exists in the primary and secondary systems due to primary to secondary leakage in the ste2m generators.

2.

Offsite power is lost and the main condenser is not available for steam dump.

1.

Eight hours after the accident the residual heat removal system starts operation to cool down the plant.

4.

Af ter eight hours following the accident, no steam and activity are released to the environment.

5.

Primary to secondary leakage is evenly distributed in the steam generators and varied from 0.01 spa to 1.0 spa.

6.

Defective fuel prior to the accident is one percent.

M fee,d 7.

As a result of the accident, Jen' percent of the fuel rods in the core are considered to be failed and their gap activity is considered to be released to, and instantaneously 71xed with, the reactor coolant.

The gap activity consists of 10 port int of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine in the daaaged rods at the time of the accident.

I 8.

No condenser air removal systes release and no steam generator blowdown occurs during the accident.

9.

No noble gas is dissolved in tha steam generator water.

10.

The iodine partition factor in the steam generators, 0.01, is determined as follows:

amount of iodine / unit mass steam amount of iodine / unit mass liquid l

15.4-39

11.

During the postulated accident, iodine in the steam generators is diluted with the incoming feedwater.

12.

The primary pressure remains constant at 2235 psis for 0-2 hours and then decreases linearly to atmospheric during the period 2-8 hours.

13.

The 0-2 hour and 2-8 hour atmospheric diffusion factors given in 3

Appendix 15A and the 0-8 hour breathing rate of 3.47 x 10~4 m /see are used.

14.

The dose model used to evaluate this accident is given in Appendix 15A.

Steam releases to the atmosphere for the reactor pump locked rotor accident are given in Table 15.4-34a. Assumptions for the realistic analysis are also presented in Table 15.4-344.

Isotopic releases to the environment using these assumptions are summarized by Tables 15.4-34b through 15.4-34e. ~0 n

1.40 x.0 ft. t 9 x s o* 7 (7.I5 vo The gamma, beta and thyroid doses for the reactortecolant eump lock d rN or a

dent, based iupon the realistic analysis, are 6~38 x 10-9) Rem,

.S T 10' Rem and G.98 x 10~3 Rem,respectively,at the site hmmdary.

Co responding doses at the low population zone are

.05 x 10'3 Rem, %

10' Rem, and d.98 x 10-5) Rem, respectively.

L 3.58 x M 3

O * *. 3

{ b' N X

q Cansna, beta and thyroid doses at the site boundary and at the low population zone for the reactor coolant pump locked rotor accident, based upon the conservative analysis, are given by Figures 15.A-77a through 15.4-77c as a function of primary to secondary leak rate. The doses retulting from this accident are well within the limits defined by 10 CFR 100 (25 Rem, whole bodyl 300 Rem, thyroid) for the range of credible steam generator tube leakage.

15.4.5 FUEI. HANDLINC ACCIDDtTS A fuel hasdling accident (FEA) during refueling could release a fraction of the fisaien product inventory in thi plant to the environment. Two accident teenerlos are consideredt (1) a r6 fueling accident occurring inside containment and (2) a refueling accident occurring outside containment.

15.4.5.1 Fuel l'sadling Accident Inside of Containae.nt Tra postulated fuel handling accident inside containment is the dropping of a spe0t fael assembly onto the core during refueling which results in damage t; the fuel assemblies. For this postulated accident, two analyses bases are evaluatedt (1) a realistic case and (2) a conservative case. ne conservative case analysis is based on Regulatory Guide 1.25 assumptions.

The assumed analysis parameters and radiological consequences associated with these cases are discussed below.

ANENI? MENT 4 15.4-40 AUGUST. 1988

l l

l TABLE 15.4-34a PARAMETERS USED IN thCKED ROTOR ACCIDENT ANALYSIS Realistic Analysis Conservat ive Analysis l

Core Thermal Power 2900 MWt 2900 MWt l

Steam Gene:ator Tube Leak Rate Prior t o Accident and for First Eight Hours Following Accident 100 lb/ day (I) 0.01 to 1.0 gpm Offsite Power Lost Lost Fuel Defects 0.12 percent I percent Failed Fuel 0.0 10 percent U

Activity Released to Reactor Coolant from Failed Fuel 0.0 10 percent of gap inventory

.,ow Parcent of Activity in Damaged Rods in the Cap Nob!c Cases (except Kr-85) 10 percent Kr-85 30 percent Iodines 10 percent Iodine Partition Factor for Steam Generators 0.01 0.01 l

Durat ion of Plant Cooldown by Secondary System after Accident 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours Blowdown Rate per Steam Cenerator prior to Accident

[gpa

[gpa l

'll if i

i i

.J

4 O

TABt.E 15.4-34b REAC*CR COOLMIT PLHP LCCXE0 ROTOR

!$0 topic IEt.!ASE TO !WIR0hHDIT REALISTIC ASALYsts Activity Released to Envircr.mant by Accidene_(CI)

A.8 Isoceeg (0-2 hr)

(_.4=9 hr)

I-131 4,50 J.At x 10~3 1Al +rH x 10"# A I-132 3At L49 n 10-5 f.sz 6mM x 10-3 I-133 1.33 b 37,x 10-5 t.:3.h*5 x 10-4 I-134 L H 3r!8 x 10-6 nig w x 10-5 I-135 3M W x 10-3 sif-6:*7 x 10-3 Xe-131m 1.co 6-r&t x 10-30 too4 M x 10"* 3 Xe-133 Lif b +3 x 10-2 l t$s 2,44 x go-J i Xe-133=

1% h+9 x 10*# 2 14.t rM x 10~J E Xe-135 S'H-t,43 x 10-3 1.0% 4,44 x 10-3' 1 Xe-135m SD.L67 x 10*I 4 m 4, N 's 10*'

Xe-137 0

2, ? ?, d * *'

o M-2-x-100-Xe '. 3 3 J.34 W x 10*'

9.1R h +7 x 10~'

Kr-33m 0

h&7 4 av I O

.M 7-d o~'"

Kr-35

't.t9 4A ', ;0-3 3 1.:.4 t-02 x 10-4 1 Kr 85m 0,10 4tte x 10-* '

1.% trst x 10~3 Kr-67 h i M 6 x 10-'

143 Gr84 x 10*# 3 Kr-!!

1.61 ht$ x 10-'1 do art $ x 10-3 Kr-39 0

- _9 :M-o.8v77 x 10-5 I

r

[

l I

i 15.4-105 l

i w.

Q TABLE 15.4-34c REACTOR COOLANT P'.HP LOCXED ROTCR n oTOPic artEAsE ro rsvineswrN: censtavAr:VE _ ANALYSIS (l)

Activity Released to Envircr. ment by Accident (ci) isotece (0-2 hr)

(0-8 hr) j I-131 2,13. W x 10'1 1.'% W x 10"I I-132 3.03 1,44 x 10'l

1. 2.1 1-res x 100

.~.

I-133 SM Frt1 x 10-1 1.'11 h-ta x 100 I-134 46% e 46 x 10-1 Ljs he x 100 v

I-135 0.% Sitt x 10-l l.Si he$ x 100 Xe 131m 1.$1 k*trx 10-l 4.31'5-17 x 10*l Xe-133 4 2.0 W x 101 n.69 6770 x 102 Xe-133m 4 334-ree x 100

M 4,33 x to f s

Xe-135 (A1 M2 x 10f0 3/A e *4 x 101 r

Xe-135m 1.% M t x 1020 32k4ttt x 101 Xe-133 Ja'1M2 x 101 idl. ht$ x 102 Kr-53m 3.y Srto 100 q.11 1,4-2 x 10 M Kr-85 for 6,42 x 10-1 2.17 2-37 x 100 Kr-85m

&M3 4v92 x 100 2.11 STt? x 101 Xr-87 t.oa,4r5-3 x 101 2

tea.47t3 x 101 j

Kr-88 LM9 MS x 101

$.15 8,+2 x 101 i

Kr-89 t,T[2,42 x 101

1. cal,-13 x 10f I 1

.i l

t I

I i

)

Note:

(1) Primary to secondary leakage equal to 0.01 gym.

L5.4-106

O TABLE !$.4-344 REACTOR CCOLANT PMP LUCXED ROTOR ISOTOPIC RELEASE TO ENVIRCNENT CCNSERVATIVE UlALYSI5(1)

Activity Released to Enviter. ment bv Accident (Cl)

Isoroie (0-2 hr)

(0-9 hr) 1-131 2.12 hv+$ x 100 g,qq w +g x toO I-132 3 03 3rh3 a 100 t.it Wx 101 I 133

%M ht1 x 100 171 +:f6 x 101 I-134 t42 W x 100 i,gg 4,a9 x 101 1-135 1A; M *. 500 ili 4,*$ x 101 Xe-131m L9 irv4+ x 100 c,3g' 9 +9 x 100 Xe-133 8 4 W x 102 f.G7 M x 103 Xe-133m 4.53 Iv4+ x 101 1!.4 'vt5 x 4

10 3.

Xe-135 1.91 W x 10 1 343 e--r9 x 102 2

Xe-135m 1.40 21 3M +W x 10 33 at. 3 to2 Xe-133 r&t x 102 gi.;.,,$ x to3 Kr-83m 3.44 M x 101 til b-H x 102 i Kr-85 147 W x 100 2.17 Iv W x 101 Kr-85m 5,g 4,42 x 101 2.tt 3.cH 2 102 Kr-57 1.ea h+3 x 102

%e2.4cir,1 x 102 Kr-88 t.W Ma x 102 fg $ m x to2 Kr-89 1.17 2Ttt x 102 7.es t 73 x 108A i

Note:

(1) Prie.ary to secondary leskage equal to 0.1 gtn.

15.4-107

I h

TAlt.E 15.4-34e REAC*CR C001. ANT PtTMP t.0CXED ROTOR i

ISOTOPIC ret. EASE TO ENVIRONMENT CONSERVATIVE ANAf.YSIS(l) j Activity Released to Environment by Accident (Cl)

{

l i

Isotoce (0-2 hr)

(0-9 hr) t

/!

I-131 2.u. 1 W x 101 S.44 h e 101 I-132

.le3 4,M a 101 1.11 +tts 102 1

0 I-133 y M7 x 101 4*. i,j f.H HM x 102 i

2-134 41 a,H x 101 1.lf M 4 x 102 4

i I-135 3.% Mt x 101 tA as*5 a 102 1

6.31' M7 x 101 l

xe-131m t,f8 m x 101 i

I Xe-133 M.2 W x 103 a.se hM x 104 1

i 1

Xe-133m 6.31 M75 x 102

fy 6,+5 x 10 3 Xe-135 t.11 M 1 x 1081 3.4 W x 103 l

Xe-135m t.% W x 10

A.

3.24 4,4 x 103 i

3 I

xe-138 334 M1Q x 103 1,42 4,+9 x 10' l

Kr-83m 2.44 M c : 102 i

i g,gg 4,91 x 1031 Kr-85

&&1 6:f2 x 101 2.21 4r$4 x 102 i

l 9

Kr-853

.i.%) J +2 x 102 g,ig.+;;9, to3 l

Kr-87 1.03 ast3 x 103 ha &cH x 103 s,T AM x 103 Kr-88 1.W +rts 103 Kr-89 1.7) 4,42 x 103 7.og ht$ x 10#3 j

I

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15.4-108 l

1

l 100 s

h,#

9, l

f-f O'

I 1

10-1

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4' 0'

3 I

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l 10-2

.?7' r!

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i i

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10-3 j

10-2 10-1 100 i

i PRIMARY TO SECONDARY LE AKAGC (GPM)

SOUTH CAROUN A ELECTRIC & GA5 CO.

VIRGIL C. 5UMMER NUCLEAR STAT 10N Locked Rotor Accident Whole Body Gamma Dose OcQ$itlAT.J5 CAsl Figure 15.4 77a

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SOUTH CAROLIN A ELECT 7UC !. G A5 CO.

l VIRG3L C. SU'.1MER NUCLE AR ST ATION l

i i

l Lecked Ret 0r A<c. cent Sur# ace Boay j

i Beta Dose l

Co.11cv rAQ t C. : e.

{

f Fg.te 15 4 776 i

101 l

4

f 4t-l t

i 1o0

-b

/

k l

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b s

J J

10-1 t

i d

l t

l k

10 - 2 10 - 2 10-1 tg0

{

PRIMARY TO $ECCNCARY LEAKAGE (GPM) j SOUTH CAROUNA ELECTRIC & G A5 CO.

l KRGIL C. 50MMER NUGEAR STAT 10N i

Locked Rotor Accident Thyroid Dose de@E4v4Tws d4sg Figwre 15.4 77c

{

I l

t

9 15.4.5.1.1 Identi'ication of Causes and Accident Description There are numerous administrative controls and physical limitations which 4

are imposed to prevent a fuel handling accident from occurring during refueling operations.

Nevertheless, an accident sequence has been postulated with the objective of assessing the potential risk to the public health and safety.

It is postulated that a spent fuel assembly is dropped onto the core during refueling resulting in breaching of the fuel rod cladding. As a result of the damage, a portion of the volatile fission gases are released to the water pool covering the core.

Subsequently, a fraction of the water soluble gases are absorbed in the pool with the remainder being transported through the water and into the reactor building atmosphere.

The escaped gases are assumed to be released instantaneously to the environment via the reactor building purge system and dispersed into the atmosphere.

j 15.4.5.1.2 Analysis of Effects and Consequences 15.4.5.1.2.1 Method of Analysis

)

The following assumptions are postulated in the calculation of the radiological consequences of a fuel handling accident inside containment:

Realistic Analyses 1.

The accident occurs at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months af ter reactor shutdown, which is the i

minleum time after shutdown that refueling operations could commence.

Radioactive decay of the fission product inventory for this time period is taken into account.

2.

All 264 pins in the dropped spent fuel assembly are damaged.

3.

The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full power operation at the end of core life insediately preceding shutdown. For the realistic analysis, nuclear characteristics of the highest-rateddischargedassemblyaregiveninTable(3424-44. The i

model discussed in Section 15.1.7.2 is used to determine these

]

fuel pellet-cladding activities.

j f[ t/.33 4

All activity in the clad gap of the damaged fuel, given in Table 15.4-36, is released to the reactor cavity pool.

5.

The maximum fuel rod pressurisation is 1200 psis.

J 6.

The minisua water depth between the top of the damaged fuel rods and the reactor cavity pool surface is 23 feet.

7.

Noble gases released to the reactor cavity pool are immediately released to the reactor building atmosphere.

l 1

AMENDMENT 4 15.4-41 AUCUST. 1983 4

The response time for the gas channel of RM-A4 to provide a closure signal directly to the interlocked reactor building purge isolation valves is based upon the following assumptions.

(a) Sample line length of 30 feet.

(b) Sample cavity of 0.04 ft3 (c) Sample flow of I cfm.

(d) Electronic and relt.y response of 0.56 seconds.

This results in a 13.2 second transient time for the sample from the duct to the detector. Adding the electronic and relay response produces a total of 13.8 se onds.

Combining this with the closure time of the reactor building purge isolation valves (see Technical Specifications) of less chan five seconds and adding an air flow time of 22.4 seconds from the fuel handling accident puff through the purge exhaust duct to the monitor sample point produces a total of less than 41.2 seconds from the occurrence of the postulated fuel handling accident puff inside containnunt until the reactor building purge is isolated.

15.4.5.4.2 Environmental Cansequences of a Postulat.ed Fuel

' andling Accident Outside of Containssent J

Following a postulated fuel handling accident Outside containment, a

quantity of airborne radioactivity would be released to the environment via the fuel handling building charcoal exhaust system.

The dose received by an individual standing at the exclusion area boundary for the accident duration has been evaluated for both a conservative and a realistic case.

The bases for the conservative Regulatory Guide 1.25 evaluations are as follows:

1.

The accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown, since plant technical specifications require the reactor to be subcritical 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the movement of spent fuel.

Radioactive decay of the fission product inventory during the interval between shutdown and the start of refueling activities is taken into account.

2.

The number of pins broken is a total of 314 pins.

Being equivalent to 1.19 assemblies, this quantity of pins broken represents 50 pins broken in the impacted assembly as well as the 264 pins of the dropped assembly.

3.

The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damsged assembly are calculated assuming full power operation at the end of core life immediatelg preceding shutdown. Th: per:</terperature d! tribut!^a sn !f used in both analyses are given in Table (15.'-3f) feaking factors Peak,_e44141 4eetenwgiver, i-FOAP Table --15. '-39.

i IS, Y-M AMENDMENT 3 15.4-47 AUGUST, 1987

4.

The maximum fuel red pressurization is 1200 psig.

5.

The minimum water depth between the top of the damaged fuel rods and the spent fuel pool surface is 23 feet.

6.

All activity in the clad gap, given in Table 15.4-37, is released to the spent fuel pool in the conservative analysis. This activity consists of 10 percent of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine in the rods at the time of the accident. Activity available for releare is given in Table 15.4-37.

7.

Noble gases released to the spent fuel pool are immediately released to the atmosphere.

8.

In the conservative analysis, the iodine gap inventory is composed of inorganic species (99.75%) and organic species (0.25%).

9.

The spent fuel pool decontamination factor is 133 for inorganic iodines and 1 for all other radioisotopes.

10.

No credit is taken for non-safety ventilation and purge systems.

11.

No mixing of the pool releases with the fuel handling building atmosphere is assumed.

12.

All iodine escaping from the spent fuel pool is immediately available for release to the environment.

13.

No credit is taken for natural decay due either to holdup in the fuel handling building or after the activity has been released to the environment.

4 14.

Isotopic data, breathing rate, and meteorology is given in Appendix 15A.

Assumptions used to evaluate the offsite dose of the realistic case are identical to those used for the conservative case with the following exceptions:

1.

The number of pins broken in the realistic case is to 264.

This is equivalent to one assembly. Gap activities are listed in Table 15.4-36.

I. W I S* *}-3').

Radialpeakingfactorisj4TaslistedinTable/45.4-35-r.d 1

2.

45. 4-3^.

3.

The effective decontamination factor for iodines in the spent fuel pool is 500 as given in Reference (41).

j The activities released to the environment for the conservative and realistic models ara given in Tables 15.4-36 and 15.4-37.

The doses to the thyroid, whole body, and skin are given in Table 15.4-50.

f AMENDMENT 3 15.4-48 AUGUST, 1987 i

1)EL& TE TABLE 15.4-35 NUCLEAR CHARACTERISTICS OF HICHEST RATED DISCHARCED ASSEMBLY ore Power, MWt 2900 Num e of Assemblies 157 Core Av (age Assembly Power at 102% of Full Power, NWt 18.84 Highest Power D charmed Assembly Axial Peak to Averag Ratio 1.55 Radial Peak to Av3 rage Ratio 1.55 Peak Linear Power Densit kW/ft 13.65 Maximum Centerline Fuel Te aratu e, F

3639 Power-Tes

ure Distribution for H est sted Discharged Assembly Percent o Assembly Per nt of Assembly Fuel Temperature Vol Within Po r within Pante. OF Temos ature Ranae Temper ture Range 1600 - 3800 0.04 0.07 3400 - 3600 0.6 1.1 3200 - 3400 1.4 2.2 3000 - 3200 2.2 3.3 2800 - 3000 3.1 4.4 2600 - 2800 4.1 5.5 2400 - 26 5.4 7.0 3200 -

00 6.9 8.5 2000 2200 8.9 10.3 2000 67.36 57.63 1

l 2

1 i

i AMENDMENT 3 15.4-109 AUCUST, 1987 t-

8 TABLE 15.4-36 REALISTIC CASE i co Ho# 5 MMR ACTIVITIES IN HICHEST RATED ASSEMBLY AT TIZ Cr REACTOR SHUTDOWN Curies in

.L-Percent of Activity Curies in Isotope Assembly M in Cap Cap Kr-83m 1riHir o

4,444 o.iy G,496- 0 Kr-85 OrtM 6.7 y 6 0 3 24,4-Z.t Z.

2,44-1 4 Ylo 3 Kr-85m 47&3 o Q 134 0 2.(

W C

Kr-87 4,44 0 4,+ih7 o ti OrMt 0 Kr-88 S,49 o h467. o. L 7 Iv5$- 0 Kr-89 14,7 0

0x0266 c.o t*5 0,4 49- 0 3

i Xe-131m 0r45it 9" fx10 4,86-t. ~1 Gre954 'l V >10 Xe-133 Hv6 i. t yto "

1,45

1. t 145*--

b 3 YI04 3

Xe-133m Gv396 I.oy10 i Orttt 0 '4 A,444-7 4 x 0" Xe-135 4r26 7.tyi03 Gr3+

8. *- l 1r45-c & xio" Xe-135m 4rts G. 3 > io 0 O r&5M o st.

M

3. 3 y s o"'

Xe-138

+S77 o

Or0599 c.>N N O 3

I-131 4rM f.?*/0*

1,59 l.4 14r4-8.I 110 I-132 lett c/. 9 y 10 3' OrH-d. I S' M

7 4 y./o -

Z I-133 Mrl

c. / x lo4 GrH 0 46 M s 2.S xio2-I-134 47,4-O Orte5 o.o* r 1-re?- o I-135 1M eg o y,o A49-- o. 2.4 w g.o 39-1 I

AMENDMENT 4 15.4-110 AUGUST, 1988

TABLE 15.4-37 CONSERVATIVE CASE ACTIVITIES IN HIGHEST RATED ASSEMBLY AT II.": CT REACTOR SHUTDOWN

'I 00 M ut-6 /r MR.

Curies in Percent Activ ty Curies in Isotooe Assembiv in Cap (1 Cap Ca.14 l+

Kr-83m h40-0 10 444-o Kr-85 age'r G.7 v s o' 30

-2d Z..o yio3 Kr-85m 3,46-o 10 She o Kr-87 Gr46 0 10 646 o Kr-88 AG4- 0 10 9h* o Kr-89 1 2D- 0 10 1-he O

~

Xe-131m he59 0f x 10 10 k'.rt $". t X to 1 7. y, s o "I 10 1.74,4. t. z. x s o f Xe-133 IJ,44 Xe-133m h4+

1. o y io 10

-4,4 t, o y to 4 3

Xe-135

-4773 2 8 > #o 10 4h5 z.. I x to 2 Xe-135m 4 r67-

6.. ~6 3 # o 10 44 r7 (,. w so -l Xe-138

+h o 10 1-5$-4 o

7 rra f S y'o' f

10 M rs d"". 5 x f o Y I-131 I-132 14 r4-4. ci y to I 10 144 r5-4. 'i y 30 V I-133 14r64 (. I y lo 'i 10 164,-2 G. I x /0 3 I-134 1-SM-o 10 147,4 o I-135 14r3+ 4. o y s o '

10 153,4 ty, o y fo o l

NOTE:

(1) In accordance with Regulatory Guide 1.25.

l 1

AMENDMENT 4 15.4-111 AUGUST, 1988

3.

TABLE 15.4-39 REALISTIC AND CONSERVATIVE CASE NUCLEAR CHARACTERISTICS OF HICHEST RATED DISCHARCED ASSEMBLY

(

I.

CONSERVATIVE CASE Core Power, MWt 2900 j

Number of Assemblies 157 Highest Power Discharged Assembly Axial Peak to Average Ratio (l) 1.65 r

Radial Peak to Average Ratio (1}

1.65 II.

REALISTIC CASE Core Power, MWt 2900 Number of Assemblies 157 Highest Power Discharged Assembly 1'

Axial Peak to Average Ratio

.L,56 1.fe'i' Radial Peak to Average Ratio kv44-1 G,dI' 1

l i

l 1

4 NOTE:

i 4

I (1)

In accordance with Regulatory Guide 1.25 (see Appendix 3A).

1 4

I I

AMENDMENT 4 l

15.4-114 AUGUST, 1988 l

l t

t

,,, - - - _., - - - ~... _ _ _ _ _ _ _ - _,. _

. c

a e

TABLE 15.4-40 ACTIVITI RELEASES FROM A FUEL HANDLING ACCIDENT INTO 3UILDING Activity Released Activity Released Conservative Case Realistic Case Isotope (Curies)

(Curies)

Kr-83m 0

0 3

3 Kr-85

7. 3 9 x t o

}vett 1 42.110 W Kr-85m O

5.00.; 10-3 o 1.00

.s 10 ^

Kr-87 0

0 Kr-88 o

1.50 x 'O~'

2.51.,10-E o

Kr-89 0

0 i

Xe-131m

(.,. V S t r o

Milr 9 3 9510'

+36-4 Xe-133 i. y 5 x s0

IS i,3 01-t. 6 t.5 #c PmWt Xe-133m

, # A 5 no 4 3,-H4

-1 H o no

H t.

Xe-135

r. 50 s to **'

fft

f. ' f8 * 'o Gr1-Xe-135m 7, o sjo-0,949 5 3, o p o '

4,70

0--3 Xe-138 O

O I-131

(.,,qoxia m

g,4, ty,o '

y 7

I-132

( G 33 ao -

1, 5,, ; 040 j, y, y,o 5,33 3 ;;-13 I-133

-7,z.5 x,o' 4-3 --

f4, wo 3 os I-134 0

0 I-135 q, w,,o '

6.15., ;c-2 g, i o y,,' V 3,;

g A t

1 i

i i

i

_i AMENDMENT 3

]

15.4-115 AUGUST, 1987

l TABLE 15.4-41 OFFSITE DOSES CUE TO POSTULATED FUEL HANDLINC ACCIDENT INSIDE CONTAINMENP Conservative (R.C. 1.25)

Realistic Dose Tvoe Case (Rem)

Case (Rem)

Thyroid

/. 53 4,44 x 102 1 4 3 W x 100 Camma

/. 4 o 1-rti x 100 L. I 3 4,44 x 10~1 Beta

/.45FLv61 x 100 g,40 x 10-1 j,.7 3

~

NOTE:

Dose receptor point located 1 mile away at exclusion boundary.

AMENDMENT 3 15.4-116 AUGUST, 1987

)

TABLE 15.4-50 0FFSITE DOSES DUE TO FUEL HANDLING

<sC0IDENT OUTSIDE CONTAINMENT w

)

Conservative Realistic (R.C. 1.25)

Case Dose Type Case (Rem)

(Rem)

Thyroid

~7. C,(, &rf5 x 100 t, l7._ h&3 x 10-1 Cama 1 4 o +.ft x 100 i, i3 4ro6-x 10-1 0

2.30.x 10-1 Bets

, c, f 1,43 x 10 a3 N

NOTE:

Dose receptor point located 1 mill away at exclusion boundary.

AMENDMENT 4 15.4-127 AUGUST, 1988

4 s

TABLE 15.4-51 t

FUEL HANDLINC ACCIDENT OUTSIDE CONTAII. MENT -

ISOTOPIC RELEASE TO ENVIRONMENT Conservative Realistic (R.C. 1.25)

Case Isotope Case (Cl)

(Cl)

I-131 7 4 r

-M+ x 101 S.10 SrM x 10-1 I-132 a,q z x io '

0 1 3 n id*' z.s I av34 x 10-0 I-133 3. c, 5 3,44 x 100 I-134 0

0 I-135 Z.39 h46 x 10-3 i.Io 4,40 x 10-5 Xe-131m 6. r 5-7v7+ x 102 y,35 &v46 x 10%l Xe-133 ty3 1-r42 x 105 i. 3 2. 4v&9 x 104 Xe-133m J.q 3r19 x 10d

-* t') -h+2 x 102 Xe-135 2.g-o 2r7t x 102

,,, ) er36 x 100 Xe-135m

-r,so 9249 x 10-1

s. 3 0 +r79 x 10-11 Xe-138 0

0 Kr-83m 0

0 Kr-d5 L.39 3ve9 x 103 f.vE Gr64 x 103 Kr-85m o 5,40 ; 103 o 1.00. 10 '

Kr-87 0

0 Kr-88 o 1.50 : 10- 0 o 2.51 : 10-8_

Kr-89 0

0 6

l

(.w AMENDMENT 4 15.4-128 AUGUST, 1988

4' 2.

It is assumed that 50 percent of the iodines and 100 percent of the noble gases in the fuel that melts are released to the reactor coolant. This is a very conservative assumption since only centerline melting could occur for a maximum time period of six seconds.

3.

The fraction of fuel melting is conservatively assumed to be one quarter of one percent of the core, determined by the following methods a.

A conservative upper limit of 50 percent of the rods experiencing clad damage may experience centerline melting (a total of five percent of the core).

b.

Of rods experiencing centerline melting, only a conservative maximum of the innermost ten. percent of the rod volume will actually melt (equivalent to 0.5 percent of the core that could experience melting).

c.

A conservative maximum of 50 percent of the axial length of the rod will experience melting due to the power distribution (0.5 of the 0.5 percent of the core equals 0.25 percent of the core).

The remainder of the assumptions and parameters used to calculats the activity release from the plant and the subsequent offsite doses for the ultraconservative analysis are identical to those used for the conservative analysis.

15.4.6.4.4 Results Isotopic releases to the containment are summarized in Tables 15.4-44 through 15.4-46.

I

-1

-f

(.03 Y tcI

- J 4 2. X 10 f - 3. 6 7. D o For the reali ftie analysis, the gamma, beta and thtroid do g at the site

.r boundary arer4.30 x 10-5 Rem,(T." x 10 s) Rem and(2.10 m 1G41 Rem,

(~~ l.T ft 'O Corrg' Rem and {0.32 xonding doy s at the low population zone are (0.00 4 re4 ectively._

10,

10-bRem,respectively.

- Rem,8.54 x 7

L-6.2*/ / <8 w8 'l.Cl00" betagithvrokd doses at the / @lO X'O

)

f-6. i f For the conservative analys(s,g[1e gammgS Rem, w.16 x

/

t t

si oundary (0-2 hours) arrli.33 x 15 10 aRem and @ -

i 4

respectively. Co (0-30 days) are(2,2^

10 gresponding.__ doses _at the low population zone l Rem, Li.23 IG-2l Rem and [1.23 10/ Rem, 1

respectively.

l. 2.og g io -

L qqrpo-i f., ge,, g,o

-t t

I. % r o* '

~12o 4to**

t beta nd thyroid doses atigzIgm For the ultraconservative analysi the a.,

the boundary (0-2 hours) are

.--x 10~- Rem, ^ ^3 x 10 3 Rem and h

~

.a -10 Rem, respectively.

Corresponding doigp at the _ low populgtion zone

-30 days) are[2.54 c10's Rem,[5.57 x 10~ j Rem and/1.3^. 10 i Rem, respectively.

L 2.% xto '

(.f tyi iv '

d Nfxso'

~

/

These doses are well within the limits defined in 10 CFR 100 (25 Rem, whole body 300 Rem, thyroid) at the site boundary and low population zone for the two hour and thirty day periods, respectively, after the accident.

15.4-61

e

~,

TABLE 15.4-44 i

CONTROL ROD EJECTION ACCIDENT ISOTOPIC RELEASE TO CONTAINNENT REALISTIC ANALYSIS-Isotooe Activity Released by Accident (Ci)

I-131 0.35' 4,43 x 101 I-132 G.5'8 4,40 x 101 I-133 i. 0'i Mt x 101E-I-134 t.f 2. %44 x 10W I-135

f. 99 ht5 x 101 Xe-131m di. 9 'i h40 x 10%'

Xe-133 61619,94 x 1013 Xe-133m

~3. s ( he$ x 1041 Xe-135 1 81 Svf6 x 10* E'.

i Xe-135m 1.z f 3 -M x 10Y'

,a,

._ ino

'ke-133 f.G 3 h x 10%'

0 a 0 3..

4.00 m 10 Kr-85

2. 0 (* 1,44 x 1052.

Kr-85m e/,SV 4.42 x 101 Kr-87 z M Z L,44 x 101 i

Kr-88 Sil6 h45 x 101 0

L 0^

1.00. 10

?

a L

l

)

4

)

i 4

l

'l i

4 1

l 1

)

1 15.4-121 i

i

s' TABLE 15.4-45 CONTROL ROD EJECTION ACCIDENT ISOTOPIC RELEASE TO CONTAINMENT CONSERVATIVE ANALYSIS Isotope Activity Released by Accident (Ci)

I-131 7, yo Jde x 105 I-132 g,i2. L e9 x 106 I-133

/,5 6 1,44 x 106 I-134 1.7o 1.,48 x 106 i

I-135 1.46 x 106 Xe-131m 5,9O b ec x 103 Xe-133

/,s-O hti x 106 Xe-133m z.30 4,41 x 10* I 3

Xe-135 3,3 o 4,43 x 105 Xe-135m

3. i o 4-,44 x 105 Xe-138 i,30 LA6 x 105 i

Kr-83m q,so 1re9 x 105N Kr-85 6.y o %fr5 x 103 Kr-85m

z. o o 3,44 x 105 Kr-87 3 1o 6,+7 x 105 Kr-88 f.3o 8.,00 x 105 Kr-89 6 50 Lt.4 x 10%I i

k 1

i f

r

{

1 3

I s

)

15.4-122

e I

e e

TABLE 15.4-46

{

CONTROL ROD EJECTION ACCIDENT ISOTOPIC RELEASE TO CONTAINMENT ULTRACONSERVATIVE ANALYSIS _

Isotope Activity Released by Accident (Cl) i 5

I-131 g.--;8 4re5 x 10 I-132 i. t 6, L,44 x 106 I-133 178 1,44 x 106 I-134 i.4l 3,42 x 106 I-135 1.64 x 106 Xe-131m 6' 7f-7vf4 x 103 Xe-133

1. F g 2rM x 106%I Xe-133m z.66 5,44 x 10 Xe-135 4 13 Sv4fr x 105 5

Xe-135m 3 86 5,45 x 10 Xe-138 i 43 1 43 x 106 5

Kr-83m

),a4 he6 x 10 Kr-85 6.oo LA4 x 10* 3 Kr-85m a.ro 4v00 x 105 Kr-87 q.(,31A-1 x 105 L,46 x 10*I Kr-88 c,4 5 Kr-89 g,3 3 LAs x 106I i

l I

1 i

i 15.4-123 i

ML20155B530

Text

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