ML20244B512

Determination of Potential Radiological Consequences from Fuel Handling Accident at Dresden Nuclear Power Station Unit 1

ML20244B512
Person / Time
Site:	Dresden
Issue date:	03/31/1989
From:	Laguardia T COMMONWEALTH EDISON CO.
To:
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ML20244B510	List: ML20244B508 ML20244B512
References
C04-22-002, C4-22-2, NUDOCS 8904190243
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Document C04-22-002 Page 1 of 15 JJ LL DETERMINATION OF THE POTENTIAL RADIOLOGICAL CONSEQUENCES FROM A FUEL HANDLING ACCIDENT AT THE DRESDEN NUCLEAR POWER STATION UNIT 1

'l f'

Prepared for the COMMONWEALTH EDISON COMPANY March, 1989 Approved by:

WnGm i

w Thomas S. LaGuardiaf P.E.

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Document C06-32-003 Page 2 of 15 TABLE OF CONTENTS PAGE 1.

EXECUTIVE

SUMMARY

3 2.

BASIS OF STUDY...........................................

4 2.1 Worst Case Exposure to Personnel in FB..................

4 2.2 Worst Case Exposure to Personnel Outside of FB..........

4 3.

CALCULATION METHODOLOGY.................................

5 3.1 Source Term.............................................

5 3.2 Environmental Concentrations.............

.............. 5 3.3 Dose Rates and Integrated Doses.........................

5 4.

SOURCE TERM.............................................

7 5.

RADIATION EXPOSURE

SUMMARY

9 5.1 Scenario 1..............................................

9 5.2 Scenario 2..............................................

9 6.

REFERENCES.............................................

14 Revision Log................................................

15

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Document C04-22-002 Page 3 of 15 1.

EXECUTIVE

SUMMARY

Dresden Unit 1 produced power commercially from July, 1960 to October 31, 1978, generating approximately 15,800,000 MW(e)hr of electricity.

The unit was taken off-1ine on October, 1978 to backfit it with equipment to meet new federal regulations and to perform a chemical decontamination of major pipi ng systems.

While it was out of service for retrofitting, additional regulations were issued following the March 1979 incident at Three Mile Island.

Commonwealth Edison concluded that the age of the unit and its rela-tively small size did not warrant the investment, and the unit was permanently shutdown.

The Maximum credible Accident (MCA) scenarios shifted from an operating reactor with active fuel to events in the Spent Fuel Pool (SFP) in which the spent fuel fission products have undergone significant decay.

In July of 1986, the Dresden-I license was amended to possess-but-not-operate status.

Currently, technical specification revisions, physical security plan revisions and an emergency plan revision are being prepared and will be submitted to the NRC for review and approval to obtain a possession-only licence.

This study provides projections of worst case skin and whole body radiation doses to personnel on-site and at the site boundary due to a spent fuel rupture accident of all 683 fuel assemblies in the Fuel Building fFB) pool with subsequent environmental releaseca)

This study shows that the maximum offsite whole body dose, at the nearest unrestricted area boundary, (475 meters northeast of the Fuel Building) from an accidental release of Krypton-85 would be 0.016 Rem within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and the corresponding skin dose would be 1.7 Rem within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

It can be concluded that the doses to members of the publ; c from an accidental release of Kr-85 would be sig-nificantl) less than the 10 CFR 100 whole body dose limit of 25 Rem in 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and the corresponding skin dose equivalent limit of 150 Rem in 2 Lours.

The study results also indicde that the whole body dose to members of the public would be significar.+ 1 v less than the U.S.

EPA-Protective Action Guide (PAG) limit of 1 Rem whole body.

EPA guidance indicates that no planned protecti'.m actions are recom-mended if the PAG is not exceeded, and that previously determined protective actions (such as alert notification, sheltering and/or evacuation) may be reconsidered or terminated (Reference 1).

This study was also conducted to determine the potential onsite radiological consequences of a postulated fuel handling accident in the Fuel Building, resulting in damage to the fuel cladding of 100%

of all pins in all 683 fuel assemblies with subsequent release of radioactive material into the Fuel Building atmosphere, with exposure to site personnel located in this building.

This accident was found to produce gamma whole body dose and acta skin dose rates of 10.3 rads / hour and 1085 rads / hour respectively.

(a) In addition to the 683 spent fuel assemblies contained in the Dresden-one SFP, three defective fuel pins are elso present. However, these pins havt been excluded from this analysis because Kr 85 pap activity was released at the time the pins becama defective.

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Document C04-22-002 Page 4 of 15 2.

BASIS OF STUDY A worst case accident analysis was performed consisting of a fuel handling accident resulting in a release of the available gap activity, or 30% of total inventory (Reference 2) of the fuel fis-sion product inventory to the FB with a corresponding breech of building integrity, and a release to the environment.

The radiation exposures resulting from this accident represent the upper bounds for any spent fuel accident at Dresden Unit 1.

However unlikely, this worst case accident scenario was analyzed to determine the potential onsite doses to personnel to determine the need for onsite notification, sheltering and/or evacuation.

The postulated accident which was analyzed has two exposure scenarios and are described below:

2.1 Worst Case Exposure to Personnel in FB This accident assumes cladding rupture of all 683 spent fuel assemblies located in the SFP.

The gap fission product activity is assumed to be immediately released into the building causing the maximum possible dose rates to personnel in the FB.

2.2 Worst Case Exposure to Personnel Outside of FB This accident, as above, assumes cladding rupture of all 683 spent fuel assemblies located in the SFP.

The gap fission pro-duct activity is assumed to be released to the environment, at ground level, through a postulated hole in the fuel building exterior wall.

The release of the fission products to the environment has a duration of two hours.

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Document C04-22-002 Page 5 of 15 3.

CALCULATION METHODOIDGY 3.1 Source Term The Kr-85 activity contained in the 683 Spent Fuel Assemblies in the Dresden-1 spent fuel pool, as of March 1989, was calcu-lated using the ORIGEN-2 computer code (Reference 3).

Actual historical data for each of the 683 fuel ac:emblies was used in performing this calculation (Reference 4).

3.2 Environmental Concentrations Meteorological parameters used to calculate X/Q's for the scenario 2 ground level release followed the Pasquill meth-odology.

The general equation used for ground level release is q

as follows:

X=

1 Q

rpa o,# (ground) y where:

X the short term average centerline value of the ground

=

3 level concentration (Ci/m )

amount of material released (Ci/sec)

Q

=

wind speed (m/sec) y

=

horizontal standard deviation of the plume spread (m) a

=

y o,

vertical standard deviation of the plume spread (m)

=

building wake correction factor

=

For ground level releases, the following conservative assump-tions of atmospheric dispersion were used in these calculations in accordance with guidance (References 2 and 5).

(a) wind speed - 1 m/sec in a uniform direction.

(b) Pasquill stability class F (c) release height = 0.0 m (grade)

(d) cross wind distance - 0.0 m (e) a building wake correction factor of 3 is assumed (Reference 2).

Ground level releases, in Reference 2 Figure 1, do not show X/Q values less than 200 meters from the point of release.

These values were calculated using the appropriate formula (Reference 6).

Results of these calculations are found in Table 5.1.

3.3 Dose Rates and Integrated Doses In the case of whole body doses (gamma) and skin doses (beta plus gamma), to personnel located in the fuel handling building (Scenario 1 accident), doses were calculated in accordance with Reference 2.

The beta dose rate due to a semi-infinite cloud TLG ENGINEERING,INC.

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Document C04-22-002 Page 6 of 15 exposure was calculated using:

1. D = 0.23 54 X where:

1. D = beta dose rate (rad /sec)

Ed = Average beta energy per disintegration (Mev/ dis) 3 X = Concentration of beta emitting isotope (curie /m )

The gamma dose rate due to semi-infinite cloud exposure was calculated using:

'D = 0.25 E' X where:

'D = gamma dose rate (rad /sec)

E' = Average energy per disintegration 3

X = Concentration of gamma emitting isotope (curie /m )

The integrated beta and gamma doses to individuals in the path of the activity released during a Scenario 2 environmental release were calculated in accordance with Reference 2.

The beta and gamma total doses, in rads, were calculated respec-tively using,

1. D = 0.25 58 9

'D = 0.25 5' 9 where:

3 9 = Concentration time integral for the cloud (sec/m )

1.e.,

sec x

curies x

sec 3

m sec The physical constants used in calculation of beta and gamma doses due to Kr-85 were:

54 = 2.505 E-1 Mev/ dis E' = 2.2 E-3 Mov/ dis 11G ENGINEERING,INC.

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Document C04-22-002 Page 7 of 15 4.

SOURCE TERM.

The principal fission gas remaining for any potential fuel damage accident, as of March 1989, is Krypton-85 (Kr-85).

The other Krypton and Xenon radioisotopes, which were produced during plant operation, have decayed to insignificant levels since plant shutdown in 1978.

Halogen radionuclides, such as radiobrominer and radiciodines, including I-131, have also decayed to stable elements since plant shutdown.

Some residual I-129 remains in the spent fuel assemblies.

This was computed to be less than 0.984 curies total.

Using a SFP water decontamination factor of 100, and a 30% release fraction (Reference 4), the amount of I-129 available for a release to the atmosphere would only be 2.95 millicuries.

This would correspond to a Fuel Building concentration of about 3.60 E-7 4Ci/cc.

With a release rate of 4.1 E-7 Ci/sec for a two hour release, this would result in an immeasurable thyroid and whole body dose to onsite per-sonnel and to a member of the public, beyond the unrestricted area boundary.

Other semi-volatile radionuclides within the spent fuel assemblies would be released to the SFP water in the event of a fuel damage accident, but would almost completely remain within the SFP water at ambient temperature, to be removed from solution by the SFP filter / demineralized system.

Pool water decontamination factors of up to a factor of 250 would occur.

Any small fractions of semi-volatile type radionuclides which would be released from the water would rapidly be removed from the FB atmosphere onto FB surfaces, or entrapped in FB ventilation duct work or in the FB ventilation fil-tration system.

The release fraction for semi-volatile type radionuclides from the fuel to the SFP Water could be as high as 0.01 with a subsequent release fraction from the SFP water to the FB atmosphere of about 0.004.

This would result in an effective dilution factor of about 25,000.

Therefore, more than 99.9% of the semi-volatile fission products would remain in the SFP Water.

Consequently, Krypton-85 is the only remaining fission product which could constitute a significant dose to onsite personnel and to the public beyond the unrestricted area boundary, due to atmospheric transport within a short time period (up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) after a fuel damage accident.

It is therefore the sole radionuclides considered in this analysis.

By computation, there are about 1.414 E5 curies of Kr-85 remaining within the 683 spent fuel assemblies currently stored in the Dres-den Unit 1 SFP.

The fraction of the Kr-85 fission products avail-able for immediate release into the water in the event of fuel clad-ding damage consists of the Kr-85 fission products which had migrated from the ceramic fuel pellet matrices ta the gap and plenum regions of the fuel pellet during normal operation followed by storage in the SFP.

The retention of noble gases, including Kr-85, TLG ENGINEERING,INC.

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Document C04-22-002 Page 8 of 15 by the SFP water is assumed to be negligible (i.e.,

a SFP water decontamination factor of 1) because the gases are chemically inert.

Using the criteria of Reference 2, it is assumed that 30% of the 1.414 E5 curies of Kr-85, or 4.241 E4 curies, is released to the FB atmosphere and could be subsequently released to the environment, if the FB walls were breached, l

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Document C04-22-002 Page 9 of 15 5.

RESULTS Calculations of whole body and skin dose rates were made for Scenario 1 and integrated whole body and skin doses were made for Scenario 2.

5.1 Scenario 1 - Worst Case Exposure to Personnel in Fuel Handling Building Following a worst case accident in which all the cladding of all the 683 spent fuel assemblies contained in the Dresden 1 spent fuel pool ruptured, 4.241 E4 curies of Kr-85 (gap activity) would be released into the FB atmosphere.

This activity represents 30 percent of the total 1.414 E5 curies of Kr-85 contained in the spent fuel.

The FD has an atmospheric volume of 8,186 m3 and therefore, would have a Kr-85 concentra-tion of 5.181 curies per m3 following the cupture of the fuel cladding.

The immediate dose rates that would be produced by this activity, from a semi-infinite immersion, would be 10.3 rads per hour gamma and 1075 rads per hour beta.

The total skin dose rate would therefore, be 1085 rads per hour.

5.2 Scenario 2 - Worst Case Exposure to Personnel outside of Fuel Building Integrated doses outside of the fuel handling building have been calculated using the methodologies previously discussed and the (X/Q) values found in Table 5.1.

All integrated doses are based upon a two (2) hour release period, during which all radioactivity would be released from the FB to the outside environment.

The calculation results are presented in Table 5.2.

The nearest unrestricted area boundary is approximately 475 meters to the northeast from the fuel building.

As such the worst case skin and total body doses would be 1.7 and 0.016 rads, respectively.

A site map, showing various radii from the Fuel Building are shown in Figures 5.1 and 5.2.

Site structures are depicted on this map, and dose rates at various distances from the Fuel Building can be determined from Table 5.2.

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Document C04-22-002 Page 10 of 15 TABLE 5.1 Atmospheric Dispersion Parameters calculation of kJ Method of Eisutis and Konicek Distance Sigma-Y sigma-Z X/0 X/QCB)

(Meters)

(Meters)

(Meters)

(Sec/ad)

(Sec/s[)

1.000E+01 6.016E-01 3.454E 01 1.532E-01 5.107E-01 2.000E+01 1.125E+00 6.072E-01 4.660E-01 1.553E-01 3.000E+01 1.622E+00 8.446E-01 2.323E 01 7.743E 02 4.000E+01 2.104E+00 1.067E+00 1.417E 01

l. 743E 02 5.000E+01 2.573E+00 1.280E+00 9.663E-02 3.221E-02 6.000E+01 3.034E+00 1.485E+00 7.065E 02 2.355E 02 7.000E+01 3.487E+00 1.683E+00 5.422E-02 1.807E-02 8.000E+01 3.394E+00 1.877E+00 4.311E-02 1.437E-02 9.000E+01 4.376E+00 2.066E+00 3.522E-02 1.174E-02 1.000E+02 4.813E+00 2.247E+00 2.943E-02 9.811E-03 1.500E+02 6.941E+00 3.156E+00 1.453E 02 4.844E-03 2.000E+02 9.000E+00 3.988E+00 8.869E-03 2.956E-03 2.500E+02 1.101D01 4.767E*00 6.066E 03 2.022E 03 3.000E+02 1.298E+01 5.506E*00 4.454E 03 1.495E-03 3.500E+02 1.492E+01 6.213E+00 3.434E-03 1.145E-03 4.000E+02 1.683E+01 6.895E+00 2.743E-03 9.143E-04 4.500E+02 1.872E+01 7.555E+00 2.251E-03 7.143E 04 5.000E+02 2.059E+01 8.195E+00 1.886E-03 6.288E 04 6.000E+02 2.427E+01 9.430E+00 1.391E-03 4.636E-04 7.000E+02 2.790E+01 1.061E+01 1.075E-03 3.584E 04 8.000E+02 3.147E+01 1.175E+01 8.607E-04 2.869E-04 9.000E+02 3.501E+01 1.285E+01 7.075E-04 2.358E-04 1.000E+03 3.850E+01 1.392E+01 5.938E 04 1.979E 04 Includes a building wake factor of 3 TLG ENGINEERING,INC.

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Document C04-22-002 Page 11 of 15 TABLE 5.2 ON-SITE INTEGRATED DOSES FOR SKIN AND TOTAL BODY Distance Skin Dose

• Total Body Dose *

(meters)

(rads)

(rads) 20 383.2 3.623 30 191.0 1.806 40 116.5 1.102 50 79.45 0.7513 60 58.09 0.5493 70 44.58 0.4216 80 35.45 0.3352 90 28.96 0.2738 100 24.20 0.2288 150 11.95 0.1130 200 7.293 0.06896 250 4.988 0.04716 300 3.663 0.03463 350 2.824 0.02670 400 2.255 0.02133 450 1.851 0.01750 500 1.551 0.01467 600 1.143 0.01081 700 0.8841 0.008360 800 0.7077 0.006692 900 0.5817 0.005501 1000 0.4883 0.004617 Based upon a release rate of 5.89 curies /srcond l

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REFERENCES 1.

Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-520/1-75-001, September 1985 (Revised June 1980).

2.

U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.25,

" Assumptions Used for Evaluation The Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactor", March 23, 1972.

3.

ORIGEN-2 Computer Code " Isotope Generation and Depletion Code -

Matrix Exponential Method", Oak Ridge National Laboratory 4.

DOE 1986 RW-859 Data, Dresden Unit i spent fuel assemblies 5.

NUREG-0324, XOQD0Q Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, (Draft, August 1977),

J.F.

Sagendorf and J.T.

Goll.

(Also NRC Regulatory Guide 1.145).

6.

Eimutus, E.C.

and Konicek, M.G.,

" Derivation of Continuous Functions For the Lateral and Vertical Dispersion Co-efficients" Atmospheric Environment, Vol.

6, pp 859-863 (1972)

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Document C04-22-002 Revision Log Page 15 of 15 REVISION IDC Rev.

Date Page Description Approval 0

3/29/119 Original I

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RADI0 ANALYTICAL REPORT prepared ~for Mr. Kevin Norberg Commonwealth Edison Company Dresden Nuclear power Station by the Staff of the l

Laboratory Services Division Science Applications International Corporation Utility Services Department 3 Choke Cherry Road Rockville, MD 20850 301-977-4480 Obilu )

AL Reviewed By ~ h%1 vin L. Wright 0

~

Radiochemistry Supervisor June 9, 1987 This report contains all requested data obtained from the sample shipment received at Rockville on May 26, 1987.

Data were produced and documented in accordance.with approved quality conttol and quality assurance procedures.

please l

direct any questions concerning these data to Kelvin Wright, I

Radiochemistry Supervisor, or Larry Coe, Department Manager, Utility Services Department.

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