ML20199J731

Forwards Addl Analyses Which Support Emergency Plan Exemption Submitted 970919.Encl Analyses Address Resin Fires & Shine from Dry SFP

ML20199J731
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/20/1997
From:	Bourassa M CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20199J734	List: ML20199J731 ML20199J741
References
NUDOCS 9711280211
Download: ML20199J731 (6)

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.e 4-A CMS Energy Corrany B4 Rock Pont Nxiw nant Kennen F. Peeners 10269 t&31 North hant GerwalManaQer Charlevoix, All49120 November 20, 1997 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 - 0001 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - ADDITIONAL ANALYSES TO SUPPORT. EMERGENCY PLAN EXEMPTION SUBMITTED SEPTEMBER 19, 1997.

On October 29, 1997,-Big Rock Point staff met with Nuclear Regulatory Commission (NRC) staff in Rockville, Maryland to discuss the permanently defueled status of the Big Rock Point Plant. Following the dialogue with regard to the Emergency Plan Exemption submitted September 19, 1997, the Big Rock Point-staff agreed to:

1.

Consider forwarding additional information concerning an additional site specific analysis to determine if the spent fuel could support a rapid zircaloy oxidation resulting from an event which has caused the draining of the spent fuel pool.

2.

Provide an evaluation of the exposure to the public at the closest site boundary if the total volume of resin in one container, and in the worst case, all resin generated by chemical decontamination, were to burn and release a cloud of radioactive material.

3.

Provide an evaluation of the exposure from the spent fuel pool to the public at the closest site boundary for the hypothetical case of a loss of all spent fuel pool water.

Find attached evaluations /analyees tnat address resin fires and shine from a dry spent fuel pool. Both analyses conclude that all doses are well below the protective acticn guidelines of 1.0 rem Total' Effective Dose Equivalent (TEDE) and 5.0 rem Committed Dose Equivalent (CDE). With. regard to the r id zircaloy oxidation, Consumers Energy Company position is tat accidental draining of the spent fuel pool and zircaloy steam reaction in the spent fuel 9711290211 971120

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4 pool has already been dispositioned during the Spent Fuel Pool Expansion llearings in the ear 3y 1980's as not being credible.

Thiu position is based on (1) the licensee's and the Commission's conclusions provided in the ceptember 19, 1997 submittal, (2) the analysis presented for fuel handling accidents bounds the considerations of the August 1997 NRC report (NUREG/CR-6451, A l

Safety and Regulatory Assessment of Generric BWR and PWR Permanently Shutdown Nuclear Power Plants) and (3) there are no offsite dose consequences (doses greater than the PAGs) at the facility as of November 5,

1997,

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Michael D Bourassa Licensing Supervisor CC:

Administrator, Region III, UFNRC NRC Resident Inspector - Big Rock Point l

NRR Project Manager - OWFN, USNRC I

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CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT DOCKET 50 155 Recin Fire Analysis

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- NRC Requested Analysis: Rosin Fire PURPOSE This document evaluates the exposure to the public at the closest site boundwy if the total volume of resin in one container, and in the worst case, all resin generated by Chemical decontamination, were to burn and releases a cloud of radiot,tive material.

SYSTEM DESCRIPTION The Vectra Crosslinked polyethylene High Integnty Container (HIC) Model EL.142, is planned to be used for chemical decontamination resins. The HIC is designed fo* use as a burial container for class B and C waste as defined in 10 CFR 61. The HIC is a right circular cylinder with composstiori of g7% Ethylene hexene copolymer with melting point of 275 F (135 C). The HIC has i

dimensions of 64 inch inside diameter, height of 65 inches and thickness of 0.5 inch. During all i

phases of resin transfer, storage and transportation, the HlC resides inside a shield assembly composed of lead and steel.

The wall of the shield assembly contains 1.25 inches of lead encased in 0.38 inch thick inner steel shell and 0.88 inch thick outer steel shell. The top cover and assembly bottom are made up of two steel plates ranging in thickness from 2.0 to 3.0 inches.

DEWATERING PROCE83 i

The resin drying (dewatering) system processes powdered and bead type ton exchange resins by removing the excess water from the resins. This is accomplished in a three step process, i

The container is filit d from the plant's waste tank using excess water to keep the resin in a slurry so that a homogeneous nJxture is achiev6d in the container, During this transfer the container will be dewatered so that the available space in the container is filled with resin to the maximum estent practicable.

The excess water is pumped out of the container ucing a positive displacement diaphragm pump.

When all the pumpable water is removed, a blower is started to recircJlate air through the resin.

The blower warms the air and passes it through the resin until relative humidity of the air steam indicates that the resin bed is dry. Temperature of the resin and HIC may not exceed 170 F during this process. The system is then shut down, the fillhead rer euved and the container capped.

EXPOSURE EVALUATION Radionuclide concentration in resin were based on the analytical values provided in Appendix C.

Details of the calculation are provided in Appendix D. The resin analysis indicates that values of some of the nuclides are "Less than"(LT,) the reported values. For a conservative approach, these LT, values were used as actual concentrations except for some transuranic nuclides for which tM atM to Co40 were lower by this method than presented in Table 3.1 11 of the' Big Rock Point v..ammissioning plan To Maintain the conservative approach, the higher of either the L.T, values, or the transuranic concentrations based on Table 3,1 11, were used, l

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____.L A resin fire would be poss ble only for dry resins To the point of filling the High Integnty Container (HIC), resins are stored 'n a hold tank which provides a saturated water envir")nment with water covenng the resins The HIC has a maximum capacity of 3 43E400 cubic centimeters (121 cubic feet) of resin when completely full A full HIC is assumed to burn for the purposes of this analysis Rolease to air is in accordance with the release fractions listed in 10CFR30 7: 'or the individual radionuclides assumed present The entry in 10CFR30.72 for

• packaged mixeu waste" was not used since that value (0 001) is less conservative than release fractions for soroe of the ind'vidual radionuclides present Atmospheric dispersion to the site bounoary is taken as 6 4BE-4 Sec/ m*3 in accordance with short term ground level dispersion recommended by Regulatory Guide i M Figure 1. Dose conwucn factors for dose to the public at the site boundary were taken from Tabie 51 of EPA 400 it 92-001. These dose conversbn factors provide Total Effective Dose Equivalent (TEDE) from combined external, internal and deposition sources. Conversion factors from Table 5 2 of the same reference were used for Committed Dese Equivalent (CDE) to the thyroid.

CONCLUSION On the bases of the calculations, total HIC activity is 147 Ci Dose at the closest site boundary for the hypothetical fire is 95 5 mrem TEDE and 175 rnrem thyroid CDE to for the duration of the uccident Total reactor system radioactivity is calculated to be approximately 300 Ci As a worst case, if all this activity from chemical decontamination were to acct <nY te in one HIC although we expect it t

to be distributed in three or more), a fire could result in siw. E sundary '"se of 195 mrem TEDE, and 357 mrem thyroid CDE, based on these conservative calculations All doses are well below the PAG's of 1.0 rem TEDE and 5.0 rem CDE REFERENCES

1. Nuclear Regulatory Commission letter dated Sep 17,1997, Docket no 719159 2 Vectra Procedure OM 156 WS " User's Guide for the Vectra CL 200 Polyethylene High Integrity Containers",10-3-95 3 Teledyne Brown Engineenng Environmental Services, " report of Analysis" 7/21/95

4. Regulatory Guide 1.25 Assumptions used for evaluating the potential radiological consequence of fuel handling Accident US NRC 1992

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I Appendix A Certificate Or Compliance for Radioactive Materials Packages i

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.y NUCLEAR RE'_'ULATORY CTaMMISSION WASHINGTON, o.C. sones.4em Septader 17,1W7 Mr. Robert C. Hogg Molten Metal Technolofly, Inc.

1009 Comerce Park Dr' ve Oak Ridge, TN 37830

Dear Mr. Hogg:

As requested by your letter dated August 5,1997, enclosed is Certificate of Compliance No. 9159. Revision No. 8. for the Model No. NUPAC 14/190L, NUPAC 14/190M, NUPAC 14/190H, LN 14-170L, LN 14-170H, and LN 14-170H packages. This certificate supersedes, in its entirety, Certificate of Compliance No. 9159, Revision No. 7, dated A,til 4, 1996.

Changes made to the enclosed certificate are indicated by vertical lines in the margin.

Those on the attached list have been registered as users of the packages under the general license provisions of 10 CFR 571.12 or 49 CFR $173.471.

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The approval constitutes authority to use the packages for shipment of radioactive material and for the packages to be shipped in accordance with the provisions of 49 CFR 6173.471.

Sincerely,

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Cass R. Chappell, Chief Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No.: 71-9159

Enclosures:

1.

Certificate of Compliance No. 9159, Rev. No. 8 2.

Approval Record cc w/ enc 1: Mr. James K. O'Steen, Department of Transportation Mr. Jack D. Rollins, VECTRA Techr. ologies, Inc.

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.mixaW #egetatory agm ws, salatag tw goversaros of any ctsseuy dit99gh ce wo stub uw par bge ordi tw v.nspywd Il 1 Dal(1JTUDTIil L15ttb ON n(E B Astl Or A 1AJITY ANQT5tl uM*f or nE eACT. AGE MAIGN 08 ArttKAllON

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NUPAC application dated February 29, 1988, 1009 Comerce Park Dr ve as supple'nented, f

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Packaging 8

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Model Nos.: NUPAC 14/190L, NUPAC 14/190M, NUPAC 14/190H, LN 14-170L, 8

LN 14-170M, and LN 14-170H

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(2)

Description Steel encased lead shielded casks for radioactive material..The casks are W

right circular cylinders with a 75.5,nch 10 by 73.38-inch IH cavity. The h

p walls of the casks contain a lead thickness ranging from 1.25 to 2.63 inches I

encased in 0.38-Inch thick inner steel shell and 0.88-inch thick outer steel E,

shell. The top cover and cask botton are made up of two steel plates ranging N

in thickness from 2.0 to 3.0 inches. The primary cask lid is secured to the l

8 cylindrical cask body by eight,1-1/4-inch rachet binders. An optional F

b secondary lid is centered in the primary lid and is secured to the primary N

lid with eight, 3/4-inch studs and nuts.

Each lid is provided with a N

Neoprene gasket seal. The casks may be provided with an optional 12-gauge D

stainless steel liner (seal welded along ali edges), an optional lid vent B

line with pipe plug, and an optional 3/4-inch drain line and pipe plug. The b

< asks are provided with four equally spaced lifting / tie-down devices. The L

M primary lid is provided with three lif ting lugs and the optional setor.dary tL b

lid is provided with one lifting lug. The casks gross weights range from g

49,200 to 65,200 pounds, g;

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Model 00, Lead Tk, Top Tk, Bottos.Tk, Gross Wt, t

W Rudn inches inches inches inches counds

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NUPAC 14/190L, LN 14-170L 80.5 1.25 4.0 4.0 49,200 g!

b NUPAC 14/190M, LN 14-170M 81.5 1.75 4.0 4.0 53,500

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NUPAC 14/190H, LN 14-170H 83.25 2.63 5.0 5.0 65,200 M

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71-9159 5.

(a)

(3) Drawings D'

Model Hos. NVPAC 14/190l. HVPAC 14/190M. and NUPAC J 4/190H The packages are fabricated in accordance with Nuclear Packaging, Inc.

8 Drawing No. X-20-307-SNP, Sheets 1, 2 and 3. Revision No. A.

4 Model Hos. LN 14-170L. LN 14-170M. and LN 14-170H p.

p The packages are fabricated in accordance with LN Technologies p

Corporation Drawing No. 5025-M-2005: Sheets 1 and 2, Revision No. O.

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Contents p

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(1) Type and fonn of material Dewatered, solid, or solidified waste, or activated solid components, in p

secondary containers, and limited to the following:

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Materials in which the radioactivity is essentially uniformly distributed and in which the estimated average concentration per y

gram of contents does not exceed:

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0.0001 millicurie of radionuclides for which the A, quantity in 4

Appendix A of 10 CFR Part 71 is not more than 0.05 curle; 4

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'l 0.005 millicurie of radienuclides for which the A,ie, but notquantity in W

N Appendix A of 10 CFR Part 71 is more than 0.05 cur 4

y more than 1 curie; or h

0.3 millicurie of radionuclides for which the A quantity in E

Appendix A of 10 CFR Part 71 is more than 1 curie.

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Objects of nonradioactive material externally contaminated with M

radioactive material, provided that the radioactive material is W

not readily dispersible and the surface contamination, when E

averaged over an area of I square meter, does not exceed N

0.0001 millicurie (220,000 disintegrations per minute) per square W

centimeter of radionuclides for which the A, quantity in G

Appendix A of 10 CFR Part 71 is not more than 0.05 curie, or E

0.001 millicurie (2,200,000 disintegrations per minute) per N

l square centimeter for other radionuclides.

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8 (2) Maximum quantity of material per package g

8 Greater than Type A quantity of radioactive material which may contain N

fissile material provided the fissile material does not exceed the limits e

in 10 CFR 571.53. The decay heat load is limited to 7 watts for the N

8 Model Nos. NVPAC 14/190L, NVPAC 14/190M, LN 14-170L, and LN 14-170H; and W

8 25 watts for the Model Nos. NVPAC 14/190H and LN 14-170H casks.

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(a)

For any package containing water and/or organic substances which could t

radiolytically generate combustible gases, determination must be made by w

tests and measurements or by analysis of a representative package such that i

the following criteria are met over a period of time that is twice the 6

pj expected shipment Q n:

j (1) The hydrogen generated must be limited to a solar quantity that would be y

Dl no more than 5% by volume (or equivalent lialts for other inflammable i

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gases) of the secondary coptainer gas void if present at STP (i.e., no more than 0.063 g-moles /ft at 14.7 psia and 70'F); or (2) The secondary container and cosk cavity must be inerted with a diluent to i assure that oxygen must be limited to 5% by volume in those portions of c

the' package which could have hydrogen greater than 5%.

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For any package delivered to a carrier for transport, the secondary container 4

must be prepared for shipment in the same manner in which determination for d

p gas generation is made.

Shipment aeriod begins when the package is prepared d

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(sealed) and must be completed wit 11n twice the expected shipment time.

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(b)

For any package shipped within 10 days of preparation, or within 10 days b:

$i after venting of drums or other secondary containers, the determination in H

g (a) above need not be made, and the time restriction in (a) above does not J

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apply.

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Maximum gross weight of the contents, secondary containers, and shoring is limited i

to 20,000 pounds.

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Except for close fitting contents, shoring must be placed between secondary j

g containers and the cask cavity to minimize movement during normal conditions of j

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transport.

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The lid and the shield plug lifting lugs must not be used for lifting the cask, and 3

must be covered in transit, j

10.

The cask must be provided with either (or both) a drain line or a lid vent line as shown in the drawing in order to provide a method to leak test the package.

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,Page 4 - Certificate No. 9159 - Revision No. 8 - Docket No. 71-9159

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1. l11.

In addition to the requirements of Subpart G of 10 CFR Part 71:

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$j (a)

Prior to each sM ment, the packaging Neoprene lid seals if opened (or if 1

9 security sed is 6, ken), must be inspected.

The seals must be replaced with j

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r new seals '.f inspect:en shows any defects or every twelve (12) months, which D

ever occt' *s first. Cutty drain and vent lines must be sealed with D

approprute sealant app' led to the pipe plug threads.

(b)

Each pactaging must meet the Acceptance Tests and Maintenance Program of:

Model No dVPAC 14/190L. NUPAC 14/190M and NUPAC 14/190H 9

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Section 8.3 of O e application.

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tiodel Hos. LtLl4-170s. LN 14-170M and LN 14-170H W

LN Technologies Corpration Procedures W-036, Rev. A; 6026, Rev B; and

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Wi-013, Rev. F.

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The package shall be prepared for shipment and operated in accordance with p

the Operating Procedures of:

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tiodel Nos. NUPAC 14/190L. NUPAC 14/190M and NUPAC 14/190H Section 7 of the application.

f Model Nos. LN 14-170L. LN 14-170M and LN 14-170H LN Technologies Corporation Procedures W-025, Rev. C.

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12. The ratchet binders on the cask lid must be torqued to 100t10 ft-lb.

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13. The cask body and each cask lid must be marked in accordance with 10 CFR 571.85(c). (

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14. The rackages authorized by this certificate must be transported on a motor vehicle, N

railroad car, aircraft, inland watercraft, or hold or deck of a seagoing vessel assigned for the sole use of the licensee.

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15. The packages authorized by this certificate are hereby approved for use under the general license provisions of 10 CFR 571.12.

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16. Expiration date: April 1, 1999. This certificate is not renewable.

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(,Page 5 - Certificate No. 9159 - Revision No. 8 - Docket No. 71-9159 El l>

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e REFERENCES h

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pf Nuclear Packaging, Inc., application dated February 29, 1988.

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Supplements dated: April 19, 1988; February 16, 1993; and August 5, 1997.

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D NUS supplement dated:

November 22, 1985.

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LN Technologies Corporation supplement dated: February 16, 1988.

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Scientific Ecology Group, Inc., supplement dated: April 30, 1993 g

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l FOR THE b.S. NUCLEAR REGULATORY COMISSION g

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Cass R. Chappell, Chief I

Package Certification Section N

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t UNITED STATES

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y NUCLEAR REGULATORY COMMISSION wAmworow. o.c. seen.aoei APPROVAL RECORD Model Nos. NVPAC 14/190L, NUPAC 14/190M, NVPAC 14/190H, LN 14-170L, LN 14-170M, and LN 14-170H Certificate of Compliance No. 9159 i

Revision No. 3 By letter dated August 5,1997, VECTRA Technologies, Inc., and Molten Metal Technology, Inc., requested that the certificate holder for Certificate of Compliance No. 9159 for the Model No. NVPAC 14/190L, NUPAC 14/190M, NVPAC 14/190H, LN 14-170L, LN 14-170M, and LN 14-170H packages he changed from VECTRA Technologies, Inc., to Molten Metal Technology, Inc. Molten Metal Technology, Inc., has accepted responsibility for the completeness and accuracy of the statements and representations of the previous certificate holder. Molten Metal Technology, Inc., will be responsible for maintenance of the certificate, the safety analysis report for the package designs, and the i

quality assurance records in accordance with 10 CFR 571.91(c). Molten Metal Technology, Inc., stated that the records required by 10 CFR 571.91(c) for the package designs will be maintained at their document control center at 1556 Bear Creek Road, Oc Ridge, Tennessee. Molten Metal Technology Inc., has been issued Quality Assurance Program Approval for Radioactive Material Packages No. O.

O, under Subpart H of 10 CFR Part 71.

The Certificate' has been revised to show Molten Metal Technology, Inc., as certificate holder.

These changes do not affect the ability of the packages to meet the requirements of 10 CFR Part 71, fut f ml.l.

H Cass R, Chappell, Chief '

Packa.3 Certification Section Spent fuel Project Office Office of Nuclear Material Safety and Safeguards Date:

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