ML18348A706

Letter Reevaluation of Consequences of a Refueling Accident Inside the Containment

ML18348A706
Person / Time
Site:	Palisades
Issue date:	03/22/1977
From:	Hoffman D Consumers Power Co
To:	Schwencer A Office of Nuclear Reactor Regulation
References
Download: ML18348A706 (12)

Text

il consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201

• Area Code 517 788-0550 March 22, 1977 Director of Nuclear Reactor Regulation Att:

Mr Albert Schwencer, Chief Operating Reactor Branch No 1 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255, LICENSE DPR PALISADES PLANT -

FUEL HANDLING ACCIDENT IN CONTAINMENT As requested by your January 14, 1977 letter, the consequences of a refueling accident inside the containment of the Palisades Plant has been reevaluated.

The attached analysis confirms that the potential consequences of such an ac-cident are well within 10 CFR Part 100 guidelines.

The analysis was performed per the requirements of your referenced letter and provides the necessary documentation to close this issue.

David P Hoffman Assistant Nuclear Licensing Administrator CC:

JGKeppler, USNRC

REFUELING ACCIDENT INSIDE CONTAINMENT FOR THE PALISADES PLANT A. _ SOURCE TERM The worst refueling accident that could occur inside containment is defined as the hottest bundle dropping onto the cavity floor at two days after reactor shutdown with the immediate consequency of one outer row of fuel rods breaking{l)

The radioactive material released is as listed in Table l. Appendix A gives all assumptions made in deriving Table l.

TABLE l SOURCE TEEM ( 2 DAYS AF'rER SHUTOOWN)

B.

PUFF RELEASE Isotope I-131 I-133 Xe-133 Xe-133m Xe-135 Kr-85 Activity (Ci) 3.4am1 l.81m1 7.1JED3 5.02m3 2.46m2 9.78m2 Assuming the activities listed in Table 1 are released simultaneously, the Area Radiation Monitors will alarm immediately and initiate containment isolation valve closure.

Since the time duration for closing the isolation valve is less than the transit time for the radioactive puff to reach the exhaust grill, re-lease of radioactive material through the containment stack does not occur. Ap-pendix B gives detailed description of the puff release.

C.

WUFQRM_ :MIXTITG OF ACTIVITY ~:EA.S,l!:D WITH _CONT/i::t'.NMENT AIR If the release from the cavity is graduai and the activity unif'ormly mixed with containment air, the exposure rate at the two ABMs is sufficient to signal con-tainment isolation using semi-infinite cloud model.

The amount of activity re-leased from the stack prior to isolation is less than the case analyzed in (D).

Details are in Appendix C.

D.

WITHOUT ISOLATION Assuming the release rate from the cavity does not cause containment isolation and consequently all the activity is released through the stack, the whole body and thyroid doses to individuals at site boundary during the entire -period of release are 29 mRems and 6.2 rems respectively.

This case is not believed credible, however, it does represent the upper limit of the consequence -of a refueling acc*ident.

Since these c0nsequences are far less than Part 100 limits, no detailed dose consequences are necessary for Case (C) which is a more realistic evaluation of a probable accident description

• NOTES:

(1)

FSAR, Section 14.19.2.

.APPENDIX A - SOURCE TERM Assumptions:

l. Reactor has been in operation at 2,650.MWt for 3 fuJ.l years prior to shutdown.

2.

Reg Guide l. 25 assumptions for the radial peaking factor, gap activity, effective decontamination factor for water are adapted.

3.

Accident occurs at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown ( FSAR l4.19. 2) *

4.

Thirteen fuel rods were broken (FSAR l4.l9.2).

5.

Fission yields from U-235, and half-lives of radionuclides, are adapted from Battelle Northwest Laboratory's Chart of the Nuclides.

Isotope Y(%)

T1/2 df Activity (Ci)

I-131 2.9 8.05 d lOO 3.48 EOl I-133 6.5 20.3 h lOO l.8l EOl Xe-l33 6.5 5.27 d l

7.l3 E03 Xe-133m 6.5 2.26 d l

5.02 E03 Xe-135 6.4 9_.2. h l

2.46 E02 Kr-85 1.3 10.76 y l

9.78 E02

APPENDIX B - PUFF RELEASE Ar.ea Radiation Mani tor There are two area radiation monitors (one-out~of-two logic for isolation) inside containment during refueling period.

The set point to initiate containment iso-lation valve.closure is 20 mR/hour above background for both ARMs.

Maximum dis-tance from the cavity to either ARM is less than 9 meters.

Exposure Rate at ABMs Isotope Activity (Ci)

' (R/lm/Ci @ 1 m) l/D2 (9 m)

Exposure Rate (R/h)

I-131 3.48 EOl

.22

.012 9.18 E02 I-133 1.81 EOl

.294

.012 6.38 E03 Xe-133 7.13 E03

.01

.012 8.56 EOl Xe-133m 5.02 E03

.01

.012 6.02 EOl Xe-135 2.46 E02

.14

.012 4.27 EOl Kr-85 9.78 E02

.oo4

.012 4.62 E02 Total 2.03 R/h The exposure rate at the ARMs is 2.03 R/h from the puff released to the water surface.

This exposure rate is sufficient to signal rl closure.

Transit Time for the Puff From the Cavity Surface to the Exhaust Grill The distance between the cavity surface to the exhaust grill is 43 feet.

The air movement speed inside containment is 3 feet/second.

Therefore, it takes 14.3 seconds for the puff to get to the exhaust grill assuming the air is pushing at that -direction.

IV Closure Time The Tech Specs require routine closure testing of this valve and plant procedure Q0-5 limits the closure time to 10 seconds or less.

A 4-second closure time is routinely experienced.

Conclusion Since the transit time is greater than both the Tech Spec and typical IV closure time, there will be no radioactive material released from the exhaust duct.

APPENDIX C - UNIFORM MIXING ARM Response Assuming uniform mixing with containment air (l.684 x 106 cf or 4.77 x 104 m3),

the exposure rate to the ARMs using semi-infinite cloud model is 8.44 R/lm.

X (Ci/m3)

• Exposure Isotope Ci

~

EyX Rate (mR/h)

I-131 3.48 EOl 7.30 E04

.38 2.77 E04 2-.49 E02 I-133 l.81 EOl 3.79 E04

.50 l.90 E04

l. 71 E02 Xe-133 7.13 E03 l.49 EOl

.03 4.47 E03 4.02 E03 Xe-133m 5.02 E03

l. 05 EOl

.03 3.15 E03 2.84 E03 Xe-135 2.46 E02 5.16 E03

.25 l.29 E03 l.61 E03 Total 8.44 E03 mR/h This is sufficient to close the isolation valve.

.APPENDIX D - DOSE AT SITE BOUNDARY WITHOUT ISOLATION A.

Tliyroid Dose From CHMG 56-75 (10/17/75) attached thyroid dose to individuals at site boundary during the entire :period of radioactive :plume traveling is 6.2 rems.

B.

Whole Bod,y Dose Using the semi-infinite cloud model, the whole body dose to individuals at site boundary during the entire :period of radioactive :plume traveling is 29 mRems.

IsotoEe Ci XLQ x

~

E:r::X I-131 3.48 EOl

-4 2.6 x 10 9,05 E03

,38 3.44 E03 I-133 1.81 EOl 2.6 x 10 -4 4.71 E03

.50 2.35 E03 Xe-133 7.13 E03 2.6 x 10 -4 1.85

.03 5.56 E02 Xe-133m.

5.02 E03 2.6 x lO -4 1.31

.03 3.92 E02 Xe-135 2.46 E02 2.6 x 10 -4 6.40 E02

• 25 1.60 E02 Total 1.16 EOl D =

• 25 EEyX

. >= 2.9 E02 Rem= 29 mRem

\\..

9*

THYROID DOSE AT SITE BOUNDARY FOR THE WORST FUEL HANDLIN"G ACCIDENT WITIDUT USING THE CHARCOAL FILTSR A.

Definition for the Worst Fuel Handling Accident and Limit for Thyroid Dose at Site B~undary:

Tb.e worst accident that could occur in the spent fuel pool during fuel handling is defined as the hottest fuel bundle dr'.Lf s onto the spent fuel pool floor and one outer row of fuel rods breaks.' The thyroid dose at site Bcundary due to this worst accident shall not exceed l.5 rem, which is the limit for a complete loss-of-load incident.C2) Since the proba-bilit~{ of a fu.el handling accident is less than the probability of a loss-of-load accident, the assumption is justified.

B.

Reactor Core Iodine Inventory:

A simplified formula for the reactor core inventory, q, for a specific isotope is given by equation (l).

3.7 x 1010 (dis/sec/Ci}

Where:

... ( l).

q is the amount of isotope contained by the reactor at shutdown (Ci).

Po is the rated reacto.r power level (Ml-\\), 2650.

Y is the fission yield.

Tr is the :::-adiological half life of the isotope.

• ~

To is the effective full power -time, 3 years [3)

Values for the reactor core inventory of the iodine isotopes are given in Table l.

<Ji.SAR, Sect:i.on 14.19.2 C

2~ech. Spec., Section 3.1.4 C3Assu.'!ling the entire core is irradiated fo:!' 3 years at full po~*ie11 1-~.........,~

......,.-~--*-"'***** --........ _._...,JllJ~~,.._.,... -.*r--rc-o::-"'---'I"!""~.,,,.-.!".-.. _.--~-**:-..~'""*.. _.":""""'..,..~.........,...-... *-*-.. -*

... ~**-**.,.,..*-~*****--**-,~-*-*.---... ---.*-..* ***-****-.. **

\\_ ___ -

2 Table 1. Reactor Core Inventory of Iodine Isotopes - At Shutdmm Isoto.ee Fission Yield ~%)

Half-Life Inventory (Ci)

I-129

• 9 1.7 x 107 y 2.52.

I-131 2.9 8~05 d 6.65 E 07

!-132 4.4 2.26 h l.oo E*oa I-133 6.5 20.3 h l.49 E 08 I-134 7.6 52.0 m l.74 E OB r-*135 5.9 6.68 h l.35 E 08 C.

Activity-Released from the Damaged Fuel Assembly:

T'ne amount pf activity released from the damaged fuel assembly under the worst accident is given by eq:iation (2).

q' (Ci) = q(Ci) x (RPF) x (Fg) x (F)

Where:

• • • (2)

N q 1 is the activity released (Ci).

q is defined in paragraph B.

N is the number of fuel assemblies in the core, 204.

RPF is the radial_peaking factor, l.65(4).

Fg is the fraction of fuel in the gap.

F is the fraction of the assembly damaged, 0*.077(5).

Values of activities released.for the iodine isotopes are given in Table 2.

Table 2. Acti\\*ity Rel.cn.sed from the Darr'.aged Fuel Assembly - At Shutdown Izctooc Fg Activity {cq

. I-129

.3 4.70 E... 04 I-131

.1 4.14 E 03 I-132

.l 6.. 23 E.03

(!~>us NRC, I\\egulator:: G*.i~de 1.25.

( 5 )FSAR, Scct:!.o:;. ll~. l~"1. '.?.

-~~----.. ~-~-----*- *--___,.~-..... *.-**-!"" !'=o,...~.. ~-Y":"""'... :; -* --*-**** ----** _....., '

• ~. -- --~ -

... - *- *---*~**.... -*.~-:-*~-:*--***-,,.....-..... ~

.... ~

• • ~* ***..,...**.,:::*.*-~.. ~"I."'.~;"":"'*....... :**":.:--.. -.-......

*. ~

c_:

3

. I-133

_I-134 I-135

.1

.1

.1 9.32 E 03 1.09 E 04 8.39 E 03 D.

Effective Decontamination Factor for Spent Fuel Pool Water:

The radioactive iodines released is composed of inorganic and orga..~ic species.

The composition and the decontamination factors of these two species are given in Table 3.

Table 3.

Co~position and Decontamination Factors for the Iodine Species(G)

Species Inorganic Organic Composition 99.75"/a

.25$

133 l

The effective decontamination factor for sp7nt fuel pool water is:

EDF =

1

= lOO

.9975/133 +.0025/1 E. Activity Inhaled at Site Boundary:

The activity inhaled at.site boundar:,r by a "Standard Man"(7) is given by equation (3).

R (Ci) = q' (Ci) x (x/Q) x B *** (3)

• EDF

)

Where:

R is the activity inhaled at site borindary (Ci).

q' is defined in paragraph C.

x/Q is the dispersion coe:fficient at site boundary; 2.6 x l0-4 eec/ro.3.

B is the breathing rate of a "Standard Men" 3.47 x io-4 m3/sec.

EDF is defined in paragraph D.

(6)us NRC, Regulatory Guide 1.25 *

(7)ICRP Publication No 2.

F.

4 Values of activity inhaled at site boundary for the different iodine isotopes are given in Taple 4.

Table 4. Activity Inhaled at Site Boundary -

At Shutdown Isotooe

• Activity Inhaled (Ci)

I-129 4.24 E-13 I-131 3.73 E-o6 I-132 5.62 E-06 I-133 8.39 E-o6 I-134 9.78 E*o6 I-135 7.57 E-o6 Thyroid Dose at Site Boundary:

The dose to the thyi 1 oid* gland of a "Standard Man" due to the* iodine isotopes inhaled is given by equation (4).

2

'D (re::a) = 8.54 :~ 10

fa E Te R (Ci) *** (4)*

M Where:

D is the t~Toid dose (R,em) *

. f'a is the fraction of the* amount inhaled gets into the thyroid, 0.23(B).

E is the effective energy of the isotope (MeV).

Te is tha effective half life of the isotope (sec).

M is the thyroid weight, 20 grams.

R is defined in paragraph E.

Values of the thyroid dose for the different iodine isotopes are given in Table 5.

(S)ICRP P;.;.blica~ion r:o 2 *

.r*;*-:**.~-.*, *'3'F'~---***---**~*:*.. ~*-*.... "o*.*o.... :.*:

~.*._*:;.:..*:~:.+.. ;*,**"w... **"*"T°' **,* *..-.,_,,..

• -***;.a..--.-*---~........,,.....,...... -:-...-.*~--

e

\\.... -

G.

H.

5 Table 5.

Thyroid Dose at Site Boundary*- At Shutdown*

Isoto,Ee E ~MeV~

Te (days)

Dose ~rem)

I-129 0.068 138 3.38 E-o6 I-131.

0.23 7.6 5.53 I-132 0.65 0.097 3.01. E-01.

I-133 0.54 0.87 3.34 I-134

. 0.82 0.036 2.00 E-01.

I-135 0.52 0.28 9.71 E-Ol Thyroid Dose at Site Boundary as A Function of' Post-Removal Time:

The thyroid dose due to I-131 and I-133 as a function of post-removal time from the core is6plotted in Figure 1. Since I-129 has a very small thy"roid.

dose, 3.38 x 10-rem and I-132, I-134 and I-135 all have very short half-lives, the thyroid dose due to I-131 and I-133 is approximately equivalent to total thyToid dose during post-removal time 5-150 days. From Figure 1,.

it can be seen that at 16th day post-removal, the thyroid dose from a worst fuel handling accident is less than l.5 rem and at 69th day post-removal, the thyroid dose is less than 15 m rem..All thyroid dose calculations are done without the operation of the charcoal :filter.

==

Conclusion:==

\\

Fuel movement without activating the charcoal f'ilter should be allowed if' it is more than 16 days sine~ its removal from the core.

Should a worst fuel handling accident occur at this tin).e, the thyroid dose at ~i te boundary would be less than 1.5 rem.

CHNG:

56-75 10/17/75

r i

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