ML20235R053

Determination of Potential Radiological Consequences to Members of Public from Fuel Handling Accident at La Crosse Nuclear Facility

ML20235R053
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Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	09/30/1987
From:	Shafer P DAIRYLAND POWER COOPERATIVE
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ML20235R020	List: ML20235R016 ML20235R053
References
LAC-TR-132, NUDOCS 8710070778
Download: ML20235R053 (10)

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. .d DETERMINATION OF THE POTENTIAL RADIOLOGICAL CONSEQUENCES TO MEMBERS OF THE PUBLIC FROM A FUEL HANDLING ACCIDENT AT THE LA CROSSE NUCLEAR FACILITY by Paul W. Shafer Radiation Protection Engineer LAC-TR-132 l

September 1987  !

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The Determination of the Potential Radiological Consequence to Members of the Public from a Fuel Handling' Accident at the La Crosse Nuclear Facility INTRODUCTION On April 30, 1987, the La Crosse Boiling Water Reactor (LACBWR) was shut down permanently. By the end of June, 1987, all spent fuel assemblies had been moved into the fuel element storage well (FESW), and the reactor vessel head was replaced onto the reactor. The maximum credible accident scenarios shifted to events in the FESW. Therefore, this study was conducted to determine the radiological consequences of a postulated fuel handling accident in the FESW resulting in damage to fuel cladding and subsequent release of radioactive material.to the atmosphere.

DISCUSSION A fuel handling accident during refuelling operations could release a fraction of the fission product inventory to the Containment Building atmosphere and subsequently to the environment. Two illustrative accident scenarios were conservatively analyzed to determine the potential doses to Members of the ,

Public at the Effluent Release Boundary. These two postulated accidents-are '

described below:

1) Cask Drop Accident - This accident conservatively assumes the rupture of the cladding of all 333. spent fuel assemblies located in the FESW, due to the droppage of a spent fuel shipping cask on top of the two-tiered spent fuel storage racks.

2) Fuel Handling Accident - This accident conservatively assumes the cladding rupture of all the fuel rods in the two fuel assemblies with the highest fission product activity.

SCOPE The principle fission gas of concern for any potential fuel damage accident as of October 1987 is Kr-85 8 The other Krypton and Xenon fission gases have decayed to stable elements. The radiobromines and radioiodines, including I-131, have decayed to stable elements. Some residual I-129 remains in the spent fuel assemblies. This has been computed 2 to be only 435 mci. Usings the fuel pool water decontamination factor of 100 and the 30% release fraction , the amount of I-129 available for release to the atmosphere would only be 1.31 mci, which would result in an immeasurable thyroid and whole body dose to a Member of the Public, beyond the Effluent Release Boundary.

Therefore, Kr-85 is the only fission product considered in this analysis of offsite doses due to atmospheric transport.

It has been computed 2 that there sre about 119,200 Curies of Kr-85 remaining in the 333 spent fuel assemblies stored in the FESW. Using the criteria of U. S. NRC Regulatory Guide 1.251, it is assumed, in the Cask Drop l

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i accident scenario, that 30% of the 119,200 Curies of Kr-85, or 35,760 Ci is released to the CB atmosphere, and subsequently into the atmosphere, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident.

Calculations of offsite doses were made for both accident scenarios at the Exclusion Area Boundary =and the Effluent Release Boundary. Calculations were made for both ground level and elevated releases Radiological consequences of Fuel Damaging Accidents at the Exclusion Area Boundary (EAB) had been previously addressed",5 A site map, including the Exclusion Area Boundary (EAB) and the Effluent Release Boundary (ERB) are shown in Figure 1. The EAB was established as part of the original 10 CFR 100 Siting Criteria for LACBWR. It is a radius of 338 meters (1108'8") around the Containment Building. Part of this EAB is within the Owner Controlled Area (OCA) and part of it is offsite. The entire EAB is located within the State of Wisconsin. The ERB was established as part of the 10 CFR 50 Appendix I Effluent Release criteria for LACbWR. It is essentially the OCA within the EAB. It is an irregular shaped area approximating a triangle. The North and South boundaries of the ERB are areas at the same distance as the EAB and are located within the OCA. To the. West, the ERB is bounded by the Mississippi River bank within the OCA. 'To the East, the ERB is bounded by the OCA fence. The shortest distance from the CB, stack sampling cubicle and the stack to the ERB is in the Easterly direction, at the OCA fence near the 10m meteorological tower. The distance is slightly over 100 meters (330').

Meteorological parameters used to calculate X/Q's for both ground leval and elevated releases follow the Pasquill methodology. The general equation used for ground level releases is:

X= 1 Q suoygaB where X = the short term average centerline value of the ground level concentration (Ci/m8 )

Q = amount of material released (Ci/sec) p = wind speed (m/sec) oy = the horizontal standard deviation of the plume spread (m) oz = the vertical standard deviation of the plume spread (m)

B = building wake effect = 3 for 100 m or A = 330 m2 which ever is more conservative. 2 l

For ground level releases, the following conservative assumptions for atmospheric dispersion were used in these calculations in accordance with l.

, guidance'.

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(a) a wind speed of 1 m/sec in a uniform direction (b) Pasquill stability class F (c) the release occurs at 0 m (grade)

(d) turbulent wake atmospheric diffusion reduction by the Reactor Building and other structures is taken into account.

The general equation used for elevated releases is:

y .e -h2/20 z Q Spo yzo where: X = the short term average centerline value of the ground level concentration (Ci/m3 )

Q = amount of material released (Ci/sec) y = wind speed (m/sec) oy = the horizontal standard deviation of the plume speed (m) cz = the vertical standard deviation of the plume speed (m) h = effective release height (m) 3 For elevated releases, a calculated worst case X/Q of 2.3 E-4 sec/m at 500m to the East (90*), was used in the calculations of offsite doses6 . This i value was determined using actual meteorological data from 1983 and 1984, and the EPA model called COMPLEX I. This model is used for situations, such as LACBWR, where the stack height is less than the surrounding height of the terrain. The results using this model have been documented to be extremely >

conservative. The worst case X/Q for elevated releases cannot possibly occur at the EAB (338m) nor at the shortest distance to the ERB (100m), since the plume ,

would still be elevated, and not have impacted the bluff. Even though the I frequence of wind distribution in the easterly direction is significantly less than in the ENE or SSE directions (as used in the ODCM), it is the shortest J distance to impact, and therefore would represent the theoretical worst case.

CALCULATIONS A. The worst case accident scenario is the Cask Drop Accident with a ground level release as previously defined. Extremely stable conditions at 1 the measured points of concern are assumed with a release duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> . An FESW water decontamination of 1.0 (no decontamination) for Kr-85 releases from the fuel rods is assumed . Using calculations *" the worst case ground 2 7 8

level X/Q at 100m, for the conditions previously defined, of 1.02 E-2 sec/m was determined.

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For 100 m, the oy and a zwere determined by: "

c y = aR b l

whcre: R = 100 m i For F stability class a = 0.0722 b = 0.9031 cy = (0.0722)(100)(0.9031) oy = 4.621 m b

C z = a'R ' + C where: R = 100 m For F stability class a' =.0.086 b' = 0.74 c = -0.35 oz = (0.086)(100)(0.74) + (-0.35)

C z = 2.247 m X= 1 = 1.02 E-2 sec/m3 Q w (1 m/sec)(4.621)(2.247)(1.0)3 Reference 1, Figure 1 does not show a X/Q value for ground level release at 100m from the release point l

. However, if this graph is extrapolated back to i the Y axis, a X/Q of about 3 E-2 sec/m swould be obtained without building wake effect taken into account.

The previously determined X/Q for 338m from the release point used for this calculation is 2.2 E-3 sec/m3 * ,

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1. Average Kr-85 Release Rater j u

1' 35,760 Ci = 4.967 Ci/sec 2 hrs x 3600 sec/hr

2. Worst Case X/0 at 100m (shortest distance to the ERB)

I 1.02 E-2 sec/m8  !

3. Average Concentration at 100m:

4.967 Ci/sec x 1.02 E-2 sec/m 3 = 0.051 Ci/m3 i s

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' 4. 'Wh' ole Body Dose at 100m:  !

0.051 Ci8 x 1839 mrem /hr(s) x 2 hrs = 186.34 mrem J i m Ci/ma l

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5. Skin Dose at 100m: ,

l 0.051 Ci'x 1.53 E5 mrem /hr(s) x 2 hrs = 15.606 Rem a m T- Ci/ma

6. Worst Case X/0 at 338m (EAB): q j

2.2 E-3 sec/ms

7. Whole Body Dose at 338m (EAB):

2.2 E-3'sec/m8 x 4.967 Ci/sec x 1839 mrem /hr x 2 hrs = 40.2 mrem Ci/ma

8. Skin Dose at 338m (EAB):

2.2 E-3 sec/ms x 4.967 Ci/sec x 1.53 E5 mrem /hr x 2' hrs = 3.34 Rem Ci/m8 B. The second (2nd) worst case scenario is the Fuel Handling Accident with a ground level release as previously defined.

1. Average Kr-85 Release Rate:

431.4 Ci 2 hrs x 3600 sec/hr = 6.00 E-2 Ci/sec

2. Whole Body Dose at 100m (shortest distance to ERB) 1839 mrem /hr x 1.02 E-2 see x 6.00 E-2 Ci/sec x 2 hrs = 2.25 mrem Ci/m3 ma

3. Skin Dose at 100m:

1.53 E5 mrem /hr x 1.02 E-2 see x 6.00 E-2 Ci x 2 hrs = 187.27 mrem Ci/ma m8 see i

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4. Whole Body Dose at 338m'(EAB):-

2.2 E-3 sec/m8 x 2.25 mrem = 0.49 mrem 1.02 E-2 sec/m8

5. Skin Dose at 338m:

-2.2 E-3 sec/m8 x 187.27 mrem = 40.39 mrem  !

1.02 E-2 sec/m s.

C.. The third (3rd) worst case scenario is in the Cask Drop Accident with an elevated release as previously defined:

1. Whole Body Dose at 500m (E):

2.3 E-4 sec x 4.967 Ci/sec x 2 hrs x 1839 mrem /hr = 4.20 mrem ma Ci/m 8 2; ' Skin Dose at 500m (E):

2.3 E-4 see x 4.967'Ci/sec x 2 hrs x 1.53 E5 mrem /hr = 350 mrem m8 Ci/m8 D. The fourth (4th) worst case scenario is'the Fuel Handling Accident with an elevated release as previously defined.

1. Whole Body Dose at 500m (E):

'2.3 E-4 sec/m3 x 2.25 mrem = 0.05 mrem 1.02 E-2 sec/m8

2. Skin Dose at 500m (E):

2.3 E-4 sec/m8 x 187.27 mrem = 4.22 mrem 1.02 E-2 sec/ms

SUMMARY

The following matrix summarizes the maximum doses to a Member of the Public from the release of Kr-85 for two different accident scenarios and two different release types:

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Accident l l l Whole Body l Skin Scenario. l Release Type  ! Location  ! Dose l Dose l l  ! I l .

l l 100mE (ERB)  ! 186.3 mrem ! 15.61 Rem Cask j Ground l 338m (EAB)  ! 40.2 mrem l 3.34 Rem Drop l l l l Accident l Elevated l 500mE l 4.2 mrem l 350 mrem l l l l l l 100mE (ERB) l 2.25 mrem ! 187.3 mrem Fuel l Ground  ! 338m (EAB)  ! 0.49 mrem l 40.4 mrem Handling j j j l i Accident j Elevated l 500mE l 0.05 mrem j 4.2 mrem l

!  !  ! I i i i i i

CONCLUSION  !

The doses to a Member of the Public from the accidental release of Kr-85 for two different accident scenarios have been very conservatively determined to be significantly less than the dose. limits specified by 10 CFR 100 and for the whole body dose, below the EPA Protective Action Guide which recommends seeking shelter, as a minimum, and the consideration for evacuation. 'If the whole body dose _is less than 1 Rem, EPA guidance indicates that no planned protective actions are recommended and that previously determined protective actions may be reconsidered or terminated 8 REFERENCES I

1 U. S. Nuclear Regulatory Commission, Regulatory Guide 1.25, " Assumptions  !

Used for Evaluating The Potential Radiological Consequences of a Fuel  ;'

Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 23, 1972.

I 2

Computer Program, FACT-1, DPC (Shafer), July, 1987.

a DPC Technical Memo, Shafer to Goodman, "LACBWR Spent Fuel Fission Product Inventory," August 5, 1987.

Safety Evaluation Report for the February 4, 1980 Amendment #18 to License DPR-45.

• SEP Topic KV-20, Radiological Consequences of Fuel Damaging Accidents -

La Crosse, October 14, 1982.

6 DPC Technical Memo, Shafer (Weiss) to Goodman, " Radiological Consequences of Fuel Damage during SAFSTOR," dated August 25, 1987.

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7 Computer Program, DOSECODE-1, DPC (Shafer), May, 1984.

l l a U.S. NRC Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Revision 1, October, 1977.

• Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-520/1-75-001, September, 1975 (Revised June, 1980).

1' NUREG-0324, X00D00 Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, (Draft, August, 1977), J. F.

Sagendorf and J. T. Goll.

• Also consulted were U. S. NRC Regulatory guides 1.3 and 1.145.

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