ML18348A856

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Fuel Handling Accident in Containment
ML18348A856
Person / Time
Site: Palisades 
Issue date: 04/06/1977
From: Hoffman D
Consumers Power Co
To: Schwencer A
Office of Nuclear Reactor Regulation
References
Download: ML18348A856 (12)
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{{#Wiki_filter:* -~. consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201 o Area Code 517 788-0550 April 6, 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 Our March 22, 1977 submittal relating to the consequences of a refueling accident* inside the containment of the Palisades Plant contained typo-graphical errors in Appendices B, C and D. The entire report is resubmitted at this time and previous copies should be destroyed. 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 1. Appendix A gives all assumptions made in deriving Table 1. TABLE 1 SOURCE TERM (2 DAYS AFrER SHUTOOWN) B. PUFF RELEASE Isotope I-131 I-133 Xe-133 Xe-133m Xe-135 Kr-85 Activity (Ci) 3.48E01 1.81E01 7.13:ro3 5.02:ro3 2.46ID2 9.78:ro2 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 puf'f 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 puf'f release. C. UNIFORM MIXING OF ACTIVITY RELEASED WITH CONTAINMENT AIR If the release from the cavity is gradual and the activity uniformly mixed with containment air, the exposure rate at the two ARMs is suf'ficient 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 accident. Since these consequences 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.

I-APPENDIX A - SOURCE TERM Assumptions:

1.

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

2.

Reg Guide 1.25 assumptions for the radial peaking factor, gap activity, effective decontamination *factor for water are adapted.

3.

Accident occurs at 48 hours after shutdown (FSAR 14.19.2).

4.

Thirteen fuel rods were broken (FSAR 14.19.2).

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 100 3.48 EOl I-133 6.5 20.3 h 100 1.81 EOl Xe-133 6.5 5.27 d 1 7.13 E03 Xe-133m 6.5 2.26 d 1 5.02 E03 Xe-135 6.4 9.2 h 1 2.46 E02 Kr-85 1.3 10.76 y 1 9.78 E02

APPENDIX B - PUFF RELEASE Area Radiation Monitor 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 Exposure Isotope Activity (Ci) T (R/hr/Ci @ 1 m) l/D2 (9 m) Rate (R/h) I-131 3.48 EOl .22 .012 9.18 Fi-02 I-133 l.81 EOl .294 .012

6. 38 El-03 Xe-133 7.13 E03

.01 .012

8. 56 El-Ol Xe-133m 5.02 E03

.01 .012

6. 02 El-01 Xe-135 2.46 E02

.14 .012

4. 27 Fi-01 Kr-85 9.78 E02

.oo4 .012

4. 62 Fi-02 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 IV 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 6 4 3 Assuming uniform.mixing with containment air (1.684 x 10 cf or 4.77 x 10 m ), the exposure rate to the ARMs using semi-infinite cloud model is 8.44 R/hr. X (Ci/m3) Exposure Isotope Ci ~ EyX Rate (mR/h) I-131 3.48 EOl

7. 30 EJ-04

.38

2. 77 El-04 2.49 E02 I-133 1.81 EOl
3. 79 EJ-04

.50 1.90 E-04

1. 71 E02 Xe-133 7.13 E03 1.49 E-01

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

1. 05 E-01

.03 3.15 El-03 2.84 E03 Xe-135 2.46 E02 5.16 E-03 .25 1.29 E-03 1.16 E03 Total 8.44 E03 mR/h This is sufficient to close the isolation valve.

APPENDIX D -.DOSE AT SITE BOUNDARY WITHOUT ISOLATION A. Thyroid Dose From CHNG 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 Boay 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 X/Q x ~ ExX I-131 3.48 EOl 2.6 x 10-4 9.05 E-03 .38 3.44 E-03 I-133

1. 81 EOl 2.6 x 10-4 4.71 E-03

.50 2.35 E-03 Xe-133 7.13 E03 6 -4

2.

x 10 1.85 .03 5.56 E-02 Xe-133m 5.02 E03 6 -4

2.

x 10 1.31 .03 3.92 E-02 Xe-135 2.46 E02 6 -4

2.

x 10 6.40 E-02 .25 1.60 E-02 Total 1.16 E-01 D =.25 rnyx = 2. 9E-02 Rem = 29 mRem

THYROID OOSE AT SITE BOUNDARY FOR THE WORST FU'iL JL.1\\NDLTI-J"G ACCJJJENT WITHOUT US:UIG Tffi: CF..ARCOAL FILT~R A. Definition for the Worst Fuel Handling Accident and Limit for Th;-y.-.roid Dose. at Site Boundary: T'ne worGt accident that could occur in the spent f'uel pool during fuel handling is defined as the hottest fuel bundle dro~s onto the spent fuel pool f'J.oor and one outer row of fuel rods b::'eaks *'J_' The thyroid dose at site Bcundary due to this worst accident shall not e;<ceed 1.5 rem, which is the limit for a complete loss** of-load incident _(2) Since the proba-bilit~, of a fuel handling accident is less than the probabiiity 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 (1). 3.7 x 1010 (dis/sec/Ci) Where: ... ( l). q is the amount o:f isotope contained by the reactor at shutdmm (Ci). Po is the rated reactor power level (tvri.. \\), 2650. Y is the fission yield. Tr is the radiological half life of. the isotope.

  • ~

To is the effective full power *time, 3 years [3) Values for the reactor core inventory of the icx:line isotopes are given in Table 1. (J{.SAR, SecUon 14.19.2 ( 2~ech. Spec., Section 3.1.4 C3lssu.'!ling the entire core is irradiated for 3 years at :full po*;;e1*

  • 2 Table 1.

Reactor Core Inventory of Iodine Isotopes - At Shutdo~m Isotope Fission Yield (1o) Half-Life Inventory (Ci) I-129 . 9 7 1.7 x 10 y 2.52. I-131 2.9 8~05 d 6.65 E 07 t-132 4.4 2.26 h 1.00 E 08 I-133 6.5 20.3 h 1.49 E 08 I-13!+ 7.6 52.0 m

1. 74 E 08 r-*135 5.9 6.68 h 1.35 E 08 C.

Activity Released from the Damaged Fuel Assembly: 'rhe amount.of actiYity released from the damaged fuel e.ssembly under the worst accident is given b~r eq:iation (2). q' (Ci) = q(Ci) x (RPF) x (Fg) x (F) Where: .** (2) N q' is the activity released (Ci). q is defined in paragraph B. N is the number of fuel assemblies in the core, 2ot~. RPF is the radial peaking factor, 1.65(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\\*i~y R01.crrsed from the Damaged Fuel Assembl~r - At Shutdown Ii:otc~ Fg Activity re.) \\ ]. . I-129 .3 4.70 E.. 04 I-131 .1 4.14 E 03 I-132 .1 6.23 E 03 (l~ )US HRC, Re t:,uh to~:: G*.i ~de 1. 2'.). (5)FSAR, S*~ctio:l :J);.l:!.:?..

3 I-133 I-134 I-l35 .l .l .l 9.32 E 03 l.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 organic 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 99-75i .25% 133 1 The effective decontamination factor for spent fuel pool water is: EDF = 1 = 100 -. 9=9=7=57-r=1-=-3-=-3 -+---,. 0'"""'0""'"'25=-7,..,,...1 E. Activity Inhaled at Site Boundary: The activity inhaled at site boundary 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 botindary (Ci). q' is de~ined in paragraph C. I 6 -4 I 3 x Q is the dispersion coefficient at site boundary; 2. x 10 ~ec lil

  • B is the brec.thing 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.

4 Values of activity inhaled at site boundary for the different iodine isotopes are given in Table 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-06 I-132 5.62 E-06 I-133 8.39 E-o6 I-134 9.78 E-o6 I-135 7.57 E-o6 F. Thyroid Dose at Site Boundary: The dose to the thyi*oid* gland of a "Standard Man" due to the iodine isotopes inhaled is given by equation (4). D (rcra) -- 8.511 2 x 10 ft. E Te R (Ci ) *.* (!-!* )* M Where: D is the t~Toid dose (t(,em). fa is the fraction of the amount inhaled gets into the thyroid, 0.23(8). -E is the effective energy of the isotope (MeV). Te is the 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. (B)ICRP P"J.blica:.ion No 2.

G.

  • H.

5 e Table 5. Thyroid Dose at Site Boundary - At ShutdOim Isotope E (MeV) Te (days) Dose ~rem) I-129 0.068 138 3.38 E-06 I-131 0.23 7.6 5.53 I-132 0.65 0.097 3.01 E-01 I-133

o. 511-0.87 3.34 I-134 0.82 0.036 2.00 E~Ol I-135 0.52 0.28
9. 71 E-01 Thyroid Do9e 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 thyroid 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 i.s approximatel;r equivalent to total thyroid 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 1.5 rem and at 69th day post-removal, the thyroid dose is less than 15 m rem..All thyroid dose calculations are done vri thout t.he operation of the charcoal filter.

== Conclusion:== Fuel movement. without activating the charcoal filter should be allowed if it is more than 16 days sine~ its rer.1oval from the core. Should a worst fuel handling accident occur at this till]e, the thyroid dose at i?ite boundary would be less than 1.5 rem. CHNG: 56-75 10/17/75 t

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