Text

Forwards Docketing of Licensee Info Provided Electronically for NRC Staff Review & Approval of BRP Emergency Plan & Associated Exemption Request Submitted by
	ML20216C943
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	03/11/1998
From:	Harris P; NRC (Affiliation Not Assigned)
To:	; NRC (Affiliation Not Assigned)
References
	TAC-M99688, TAC-M99689, NUDOCS 9803160245
	Download: ML20216C943 (102)
	v • d • e

{{#Wiki_filter:i I March 11, 1998 54EMORANDUM TO: File ORIGINAL SIGNED BY: . aOM: Paul W. Harris, Project Manager Non-Power reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation 3

SUBJECT:

DOCKETING OF LICENSEE INFORMATION PROVIDED ELECTRONICALLY FOR NRC STAFF REVIEW OF LICENSE SUBMITTALS The attached was received from Mr. Mike Bourassa, Big Rock Point (BRP) Nuclear Plant, as information for NRC staff review and approval of the BRP emergency plan and N associated exemption requests (TAC Nos. M99688 and M99689, respectively) as s submitted by letter dated September : 9,1997. This information shall be docketed, t. Docket No. 50-155 Attachments: 1. BRP Resin Fire Analysis (27 facsimile pages dated November 18,1997) 2. BRP Shine From Dry Pool (21 facsimile pages dated November 18,1997) 3. BRP General Information: Spem Fuel Pool and Racks (21 facsimila pages dated October 7,1997) 4. BRP Fuel Damage Decommissioning Accident Analysis (17 facsimilc pages dated October 7,1997) 5. BRP Fuel Damage Decommissier.iog Accident Analysis (14 facsimile pages dated October 7,1997) DISTRIBUTION (hard copy) sDocket Fils 50-155e Region 111 d {[ PUBLIC RLeemon, BRP Resident inspector [ PDND r/f (w/oenclosure) PHarris DISTRIBUTION (e-mail, cover memo only) MMasnik RBurrows RDudley JMinns MWebb MFairtile TFredrichs AMarkley LThonus LWheeler MMendonca DOCUMENT NAME: g:\\secy\\harrisidocketin.BRP Ts vocalvo e copu of this docuenent,inacate in the Irv " - Copy without enclosures *E' = Copy with enclosures "N* = No copy l OFFICE PDND: phi) lr E PDND/A ~ l E PDND:(A)SC PDND:(A)D l l l w ~ lNAME PHarris W @n MMasnik Md MMendonca P l DATE 31 9 /98' '3/ /f /98 3/ ll /98 3/ i /98 t OFFICIAL RECORD COPY y 9003160245 980311 ll-l h.. g l ll PDR ADOCK 05000155 FLgy= CENTE3 COPY l e eda c dg ^

f* **% 4 UNITED STATES NUCLEAR REGULATORY COMMISSION u o A f WASHINGTON. D.C. 20555 4 001 % *****/ March 11,1998 1 MEMORANDUM TO: File FROM: Paul W. Harris, Project Manager j Non-Power reactors and Decommissioning j Project Directorate i Division of Reactor Program Management Office of Nuclear Reactor Regulation q

SUBJECT:

DOCKETING OF LICENSEE INFORMATION PROVIDED ELECTRONICALLY FOR NRC STAFF REVIEW OF LICENSE SUBMITTALS l The attached was received from Mr. Mike Bourassa, Big Rock Point (BRP) Nuclear Plant, as information for NRC staff review and approval of the BRP emergency plan and associated exemption requests (TAC Nos. M99688 and M93689, respectively) as submitted by letter dated September 19,1997. This information shall be docketed. l Docket No. 50-155 Attachments: 1. BRP Resin Fire Analysis (27 facsimile pages dated November 18,1997) 2. BRP Shine From Dry Pool (21 facsimile pages dated November 18,1997) 3. BRP General information: Spent Fuel Pool and Racks (21 facsimile pages dated October 7,1997) 4. BRP Fuel Damage Decommissioning Accident Analysis (17 facsimile pages dated October 7,1997) 5. BRP Fuel Demage Decommissioning Accident Analysis (14 facsimile pages dated October 7,1997) I t

616 547 0340 'P4 0 0 18 '97 9:37 FROM BRP DECOM PLAT 4 PAGE.001 ^"#~"'* FAX TRANSMITTAL CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT \\ 10269 US-31 NORTH CHARLEVOIX MI 49720 FAX NUMBER [616] 547-8340 Internal Access: 1371-340 DATE: n is 47 TO: PA.UL-W HARRIS NAME: COMPANY: US NRC/NRR OFFICE PIIONE NUMBER: (3011 415-1169 (301) 415-3313 FAX NUMBER: FROM: NAME: Mike Bourassa OFFICE PIIONE NtafBER: (616) 547-8244 PAGES TO FOLLOW X D1 0'5# Fn%E h(L L L I M ' NA(2-Y E Sl'O US [ - C G E T & c-J TE3?cwSE To fi]Lcot.d

$0V 18 '97 9:38 FROM BRP DECOM PLAN PAGE.002 NRC-Requested Analysis: Resin Fire PURPOSE This document evaluates the exposure to the public at the closest srte 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 releases a cloud of radioactive 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 for use as a bunal container for class B and C waste as defined in 10 CFR 61. The HIC is a nght circular cylinder with composition of 97% Ethylene hexene copolymer v?,th melting point of 275 F (135 C) The HIC has dimensions of 64 inch inside diameter, height of 65 inches and thickness of 0.5 inch During all phases of resin transfer, storage and transportation, the HIC resides inside a shield assembly composed of lead and steel The wall of the shield assembly contains 125 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 PROCESS The resin drying (dewatenng) system processes powdered and bead type ion exchange resins by removing the excess water from the resins. This is accomplished in a three step process. The container is filled from the plant's waste tank using excess water to keep the resin in a slurry so that a homogeneous mixture is achieved in the container. Dunng this transfer the container will be dewatered so that the available space in the container is filled with resin to the maximum extent practicable The excess water is pumped out of the container using a positrve displacement diaphragm pump j When all tha pumpable water is remoced, a blower is started to recirculate 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 removed and the container capped. l EXPOSURE EVALUATION Radionuclide concentration in resin were based on the analytical values provided in AppendlX C. Details of the calculation are provided in Appendix D. The resin analysis indicates that values of j some of the nuclides are "Less than"(L T.) the reported values For a conservative approach. i these L T. values were used as actual concentrations except for some transuranic nuclides for l which the ratios to Co-60 were lower by this method than presented in Table 31 11 of the Big Rock Point Decommissioning plan To Maintain the conservative approach, the higher of either the L.T. values, or the transuranic concentrations based on Table 3.1-11, wre used l-

f. 140V 18 '97 9:38 FROM BRP DECOM PLAN PAGE.003 1 Appendix A 1 Certificate Of Compliance for Radioactive Materials Packages l l

J40V 18 ' 9,7 _9:38 FROM BRP DECOM PLAN PAGE.004 arste

e UNITED STATES 3

E NUCLEAR REGULATORY COMMISSION %,*...*/ WASHINGTON. o.C. Mets 40m 4** 17, IWI Mr. Robert C. Hogg Molten Metal Technology, Inc. 1009 Commerce Park Drive Oak Ridge, TN 37830

Dear Mr. Mogg:

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 14-170M, and LN 14-170H packages. This certificate supersedes, in its entirety, Certific?.te of Compliance No. 9159 Revision No. 7, dated April 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 171.12 or 49 CFR 6173.471. 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 5173.471. l Sincerely,- k // Cass R. Chappell, Chief Package Certification Section i Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No.: 71-9159

Enclosures:

1. Certificate of Complian:e 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 Technologies, Inc. Registered U:ers o

t40V 18 '37 9:39 FROM BRP DECOM F L Al4 PAGE.005 __ _g 3 CERTIFICATE OF COMPLIANCE d FOR RADIOACTIVE MATERIA 13 PACKAGES h I4Ci#TW A WMau 6 arvruoN W Man e racx4ct uaximcAnoM Wieta & PAGE WM8G e MAL WMSM PAGE3 9159 8 USA /9159/A 1 5 H i$ L FRLAMati

a. N ceruf cme 6s 6ssued to cerufy dus du pedagmg and corness erscribed as two $ tekre. moeu dw appiscable s,fery suedenh nas forth is rede 10.

Cade of Fedrent Regularians. N 71." Pack.agma and Trunnputasan of Radsoentwe Meenal * ( I t> h artafsme does aos reisew dw coesiseer frase - ':-~ wish any requirensras of the regaaepoes of the U1 Deperusset of Transpanaues or odwr apptasable regulatory agescws, aartedaag Ow goverur'ieal of any coestry through or isso stuch the package ed be transported I! 31'as Curt WED ow fME B Asts OF A SAFTTY AN ALYst REPOIrf OF ACKA I Molten Metal Technology, Inc. NUPAC application dated February 29, 1988, 1009 Cosmerce Park Drive as supplemented. l Oak Ridge, TN 37830 71-9159 ,pocx,7,u a a cowornons 5 h canarseme w coedasonal spas feardhug Ibe riquatermes of #0 cm Post 78 as appbutde, and the raad=-- spec (sed hetow. g s. 5 N (a) Packaging (1) Model Nos.: NUPAC 14/190L, NUPAC 14/190N, NUPAC 14/190H, LN 14-170L, LN 14-170N, and LN 14-170H E (2) Description Steel encased l'ead shielded casks for radioactive material..The casks are W I right circular cylinders with a 75.5-inch ID by 73.38-inch IH cavity. The k walls of the casks contain a lead thickness ranging from 1.25 to 2.63 inches F encased in 0.38-inch thick inner steel shell end 0.88-inch thick outer steel shell. The top cover and cask bottom are made up of two steel plates ranging 5 in thickness from 2.0 to 3.0 inches. The primary cask lid is secured to the E cylindrical cask body by eight, 1-1/4-inch rachet binders. An optional N secondary lid is centered in the primary lid and is secured to the primary lid with eight, 3/4-inch studs and nuts. Each lid is provided with a Neoprene gasket seal. The casks may be provided with an optional 12-gauge stainless steel liner (seal welded along all edges), an optional lid vent line with pipe plug, and an optional 3/4-inch drain line and pipe plug. The casks are provided with four equally spaced lifting / tie-down devices. The primary lid is provided with three lifting lugs and the optional secondary t lid is provided with one lifting lug. The casks gross weights range from n 49,200 to 65,200 pounds. g F Model 00, Lead Tk, Top Tk, Bottos,Tk, Gross Wt, g Number inches inches inches inches nounds k NUPAC 14/190L, LN 14-170L 80.5 1.25 4.0 4.0 49,200 g NUPAC 14/190M, LN 14-170M 81.5 1.75 4.0 4.0 53,500 k NUPAC 14/190H, LN 14-170H 83.25 2.63 5.0 5.0 65,200 g F r E t \\ g t E E F ....,..,,,, m,., _, _,..,,,,,,,,,,,,,. ,j i l

NOU le *97 9:39 FPOM BRP DECOM PLAN PAGE.C06 j ctmmoe-u.s.wucusan nesuwom em D Lac, m ess' b. I Page 2 - Certificate No. 9159 - Revision No. 8 - Docket No. 71-9159 I b I B 5. (a) (3) Drawings / B I tkdel Nos. NUPAC 14/190L. NUPAC 14/190M. and NUPAC 14/190N b b The packages are fabricated in accordance with Nuclear Packaging, Inc. I Drawing No. X-20-307-5Np, Sheets I, 2 and 3. Revision No. A. I Model Nos. LN 14-170L. LN 14-170N. and LN 14-170H I I The packages are fabricated in accordance with LN Technologies N Sheets I and 2, Revision No. O. I Corporation Drawing No. 5025-M-2005: I I (b) Contents 1 I (1) Type and form of material i Dewatered, solid, or solidified waste, or activated solid comporats, in g secondary containers, and limited to the following: p b' Materials in which the radioactivity is essentially unifonnly 3 (i) distributed and in which the estimated average concentration per y gram of contents does not exceed: y I 0.0001 millicurie of radionuclides for which the A, quantity in y Appendix A of 10 CFR Part 71 is not more than 0.05 curfe; y l 0.005 millicurie of radionucildes for which the A, quantity in Appendix A of 10 CFR Part 71 is more than 0.05 curie, but not g more than 1 curie; or y 0.3 millicurie of radionuclides for which the A quantity in l AppendixAof10CFRPart71ismorethan1 curie. g Objects of nonradioactive material externally contaminated with l (11) radioactive material, provided that the radioactive material is y not readily dispersible and the surface contamination, when l 8 averaged over an area of I square meter, does not exceed U 0.0001 millicurie (220,000 disintegrations per minute) per square 8 centimeter of radionuclides for which the A, quantity in 8 Appendix A of 10 CFR Part 71 is not more than 0.05 curie, or N 0.001 millicurie (2,200,000 disintegrations per minute) per square centimeter for other radionuclides. (2) Maximum quantity of material per package I Greater than Type A quantity cf radioactive material which may contain I fissile material provided the fissile material does not exceed the limits in 10 CFR 571.53. The decay heat load is limited to 7 watts for the 8 I Model Nos. NUPAC 14/190L, NUPAC 14/190M, LN 14-170L, and LN 14-170M; and l 25 watts for the Model Nos. NUPAC 14/190H and LN 14-170H casks. p p p B! b 1 1 y 3, 7 r y 7 ,, y y y y y 3 y y y y4 u

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DT *~ D F * ! I i T 't~f. I s

poV 18 '97 9: 40 FROM BRP DECOM__LAN _ _ _ _ _ _ _ _ _ _ _ _ _,PAGE,.0,07 P f Page 3 - Certificate No. 9159 - Revision No. 8 - Docket No. f[ g N 81** womoe.-- un wucuAn neouunmy== =- y 71-9159 E! E I 6. (a) for my package containing water and/or organic substances writch could 1 radiosytically generate combustible gases, determination must be made by gl 1 tests and measurements or by analysis of a representative package such that g I the following criteria are met over a period of time that is twice the g! I expected shipment time: g! I g, e (1) The hydrogen generated must be limited to a solar quantity that would be g: I no more than 5% by volume (or equivalent limits for other inflammable gl l gases) of the secondary coptainer gas void if present at STP (i.e., no gl s more than 0.063 g-moles /ft at 14.7 psia and 70'F); or g: f l (2) The secondary container and cask cavity suet be inerted with a diluent to assure that oxygen must be limited to 5% by volume in those portions of' f' i the' package which could have hydrogen greater than 55. For any package delivered to a carrier for transport, the secondary container I must be prepared for shipment in the same manner in which detensination for E. p gas generation is made. Shipment period begins when the package is prepared Il p (sealed) and sust be completed within twice the expected shipment time. l (b) For any package shipped within lo days of preparation, or within 10 days E j after venting of drums or other secondary containers, the determination in E (a) above need not be made, and the time restriction in (a) above does not E : apply. 7. Maximum gross weight of the contents, secondary containers, and shoring is limited to 20,000 pounds. 8. Except for close fitting contents, shoring must be placed between secondary E i containers and the cask cavity to minimize movement during normal conditions of E transport. E E 9. The lid and the shield plug lifting lugs must not be used for lifting the cask, and E must be covered in transit. E E 10. The cask must be provided with either (or both) a drain line or a lid vent line as E shown in the drawing in order to provide a method to leak test the package. E E E E 6 6 E E E E E E E E s I W v--- <,-,,.m _,6

240V 18 '97 9: 41 FROM BRP DECOM PLAN PAGE.008 g a sina co w. - us nocts u nacuLA70m cosemamon Page 4 - Certificate No. 9159 - Revision Ko. 8 - Docket No. 71-9159 b 11. In addition to the requirements of subpart G of 10 CFR Part 71: B (a) Prior to each shipment, the packaging Neoprene lid seals if opened (or if B security seal is broken), must be inspected. The seals must be replaced with b new seals if inspection shows any defects or every twelve (12) months, which ever occurs first. Cavity drain and vent lines must be sealed with b appropriate sealant applied to the pipe plug threads. W D (b) Each packaging must meet the Acceptance Tests and Maintenance Program of: I b Model Nos. NUPAC 14/190L. NUPAC 14/190M and NUPAC 14/190H I I Section 8.0 of the application. I I Model Nos. LN 14-170L. LN 14-170M and LN 14-170H E LN Technologies Corporation Procedures iM-036, Rev. A; W-026 Rev. 8; and p. W-013, Rev. F. p I g (c) The package shall be prepared for shipment and operated in accordance with I the Operating Procedures of: B Model Mos. NUPAC _14/190L. NUPAC 14/190M and NUPAC 14/190H g Section of the appilcation. I Model Nos. LN 14-170L. LN 14-170N and LN 14-170H g LN Technologies Corporation Procedures WM-025 Rev. C. l

12. The ratchet binders on the cask lid must be torqued to 100110 ft-lb.

13. The cask body and each cask lid must be marked in accordance with 10 CFR 571.ES(c).

[ l L

14. The packages authorized by this cerMficate must be transported on a motor vehicle, J

g railroad car, aircraf t, inland watercraft, or hold or deck of a seagoing vessel l assigned for the sole use of the licensee. y

15. The packages authorized by this certificate are hereby approved for use under the

{ f general license provisions of 10 CFR 571.12.

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

[ l b I B B' b I B Pi*s x s u O 1 z.1.12.n.1.1K1 zz.1xt.xxxxx1.2.xxxxu.sx=- s T3xx, t _

~~ l !40V 18 '97 9: 41 FROM BRP DECOM PL At4 PAGE.009 [ g Fo* staa com m w e u.s.wuctrAn ts:vLArony comanssion l E. l Page 5 - Certificate No. 9159 - Revision No. 8 - Docket No. 71-9159 gj i E: ) 1 El l E' REFERENCES u 4 E l Nuclear Packaging, Inc., application dated February 29, 1988. g y. Supplements dated: April 19, 1988; February 16, 1993; and August 5, 1997. l g i NUS supplement dated: November 22, 1985. g I LN Technologies Corporation supplement dated: February 16, 1988. gl E Scientific Ecology Group, Inc., supplement dated: April 30, 1993 g E' FOR THE U.S. NUCLEAR REGULATORY C00911SSION Cass R. Chappell, Chief l Peckage Certification Section Spent fuel Project Office I Office of Nuclear Material Safety E and Safeguards l. fl E Date: septader 17. 1997 E! E' E-E El E b E E E E 1 N [ r r,, . r,,,- r r,..,.,,, ...., 1 x 1 1 tx1.33,3232x2x

FOU 18 '97_ 9: 42 FROM BRP DECOM PLAN PAGE.010 no: \\ .p umrao sTATas g NUCLEAR REGULATORY COMMISSION j i j g, wasenwom, o.c. ame.en APPROVAL RECORD Model Nos. WPAC 14/190L, WPAC 14/190N, W PAC 14/190H, LN 14-170L, LN 14-170M, and LN 14-170H Certificate of Compliance No. 9159 Revision No. 5 sy letter dated August 5,1997, VECTAA Technologies, Inc., and Molten Metal Technology, Inc., requested that the certificate holder for Certificate of Compliance No. 9159 for the Model No. WPAC 14/190L, NUPAC 14/190M, WPAC 14/190H. LN 14-170L, LN 14-170M, and LN 14-170N packages be 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 quality assurance records in accordance with 10 CFR 571.91(c). Molten Metal Technology, Inc., stated that the records required by 10 CFR $71.91(c) for the package designs will be maintained at their document control center at 1556 8 ear Creek Road, Oak Ridge, Tennessee. Molten Metal Technology Inc.,!.as been issued Quality Assurance Program Approval for Radioactive Material Packages No. O, under Subpart H of 10 CFR Part 71. The Certificat7 has been revised to show Molten Metal Technology, Inc., as certificate holder. These changes do not affect the ability of the packages to meet the requirements of 10 CFR Part 71. f 1'=$ H Cass R. Chappell, Chief ' Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Date: s*

17. 1997 I

4 gm<n

- 140 U 18 '97-3:42,,_ FROM BRP DECOM PLAN .PAGE 011-5 i i i i Appendix B 1 Users Guide for the Vectora CL-200 Polyethylene i High Integrity Containers l L i ) l- ?

I NOV 1 8 ,',9 7 9: 43 FROM BRP DECOM PLAN PAGE.012 V VECTRA l USER'S GUIDE FOR THE VECTRA CL-200 POLYETHYLENE i HIGH INTEGRITY CONTAINERS CONTROLLED OM-16 WS 6,. 03>C03-NS Revision: 6 Effective Date: (0/3,.T Essential Related VECTRA Documents

nie following related VECTRA Document (s) contain operations or information essential to performance of instructions herein and must be issued in conjunction with this document:

1. DHEC-HIC-PL-012 C of C 2._ E 18-W S

3. H-19-WS-4.__

0.(& lOf&lQS Prepared By (i Date ' A.Jc-,L bJL-to It-l4s Engineering Date Ad $mmm x1_ _ s c l t ) c,s " Customer Service Date 5Y /b/W Quality Assurance Date/ ' m L fu l__ io/.e /r Other - Licensing Date 0. 00l> O h Document Control ( Date l GCTRA Technolos'en. Ne. on. Hortusen way. Sv. 209 Colomb o. SC 29212 3408 T.I. (So3) 7814426 Fcx !803) 781.9316 a

Nov 18_l97, 9: 43 FROM BRP DECOM PLAT 4 PAGE,013 RECORD OF REVISIONS rdure No. OM-16-WS Preparation Date 10/0v95 TITLE: User's Guide for the VECTRA CL-200 Polyethylene High Integrity Containers REV DESCRIPTION PAGE(S) DATE SIGNATURE AFFECTED 05 See Hhtory File 6 Revised to update to VECTRA format. All lOf4'qE Ikd, 1 3 _ r_. - ~ ^ i

NOU 18 '97 9: 43 FROM BRP DECOM PLAN PAGE.014 i OM.16.WS, Rw. 4 Ocukr 2,1995 i I I TABLE OF CONTENTS l 1.0 PURPOSE AND SCOPE l 2.0 RsFERENCEs l 3.0 DEFINITIONS l l 4.0 RESPONSIBILITIES 1 5.0 PRECAUTIONS / PREREQUISITES 6.0 PROCEDURE - CROSSLINKED POLYETHYLENE USE AND OPERATIONS l APPENDIX A ATTACHMFET 1 ATTACHMENT 2 t 1 ATTACHMENT 3 1 l l l il l

a NOU 18<'97 9: 44 FROM-BRP DECOM PLAN PAGE.015 OM.t6-W$, Rev. 6 Orth a, app

l 1.0 PURPOSE AND SCOPE This document delineates the proper use and operation of the VECTRA Crosslinked Polyethylene High Imegrity Container. It does not include a description of the operation of ar.y internally installed process equipment, but rather sets limits regr.rding their usc.

2.0 REFERENCES

2.1 Eneineerina Pronetties of Marlex Resi_ns. TSM-243, Technical Memoraudum, Phillips Chemical Company. \\ 2.2 G. E. Carrow Cross Linkable Polvethylene. The Proven Placie for Handil ng Corrosive Chemicals Paper 252. Corrosion '82. 2.3-Title 10, Part 61, Code of Federal Regulations. 2.4 Title 49, Part 173, Code of Federal Regulations. i 3.0 DEFINITIONS "Not Applicable" 4.0 RESPONSIBILITIES "Not Applicable" 5.0 PRECAUTIONS / PREREQUISITES "Not Applicable" 6.0 - PROCEDURE CROSSLINKED POLYETHYLENE USE AND OPERATIONS 4 6.1 The VECTRA Crosslinked Polyethylene High Integrity Container is designed for use as a burial container for class B and C wastes (as defined in 10 CFR 61 the Barnwell, South Carolina disposal site. The container may be allowed for use at an alternate dispcsal site wNch may implement additional quiremer.ts that must be met before use is granted. Contact the alternate disposal site for specific requirements. 6.1.1 The user must be certified as a user per Artachment I, which must be on file with the State of South Carolina prior to use. The requirements of this step must still be roet if the container is destined for en alternate disW site. 1 of 7

- t40V.- 10 '97 _.S: 44 .FROM BRP DtICOM PLAN-PAGE.OlG I oM.36.ws, a,v. 6 osober 2,1994 1 -NO'TE: De user is required to ensure that operational requirements are in effect to control the storage and use of this container. Bis procedure may be used direct!) or as a reference to :he end user's procedure. 6.2 Storage 6.2.1 VECTRA Crosslinked Polyethylene Containers must be stored upright with the black plastic wrapping kept on the cantainers until just prior to use. j The wrapping may be removed if the containers are to be stored indoors prior to use. In no case shall a container be used : hat has received more than 1 year of ultraviolet exposure. 6.2.2 The empty containers may be handled using the lifting sling or lift ring _ j assembly cables, as appropriate, or by forklift, provided the containers are placed on a suitable pallet Once the contameris loaded, the package must be handled only by the lifting sling or lift ring assembly cables, as appropriate. 6.2.3 De storage area should be kept free from' gravel, trash or other debris capable of damaging the package. De chemicals listed in Appendix A shall not be allowed to contact the Polyethylene while the conta ner is being handic.1 or is in storage. 6.14 De empty containers may be stacked 3_high. Appropriate measures should be taken to prevent toppling of the stacks for safety considerations. In proportion to their size, the empty containers are light weight and winds could cause toppling. 6.3 Operation 6.3.1 ne black plastic wrapping shall be removed prior to any loading ) operaions. 6.3.2 The package shall be carefully examined prior to loading to verify that no i googes or other damage has occurred during handling or storage of the container. Scaling surfaces shall be examined to be free from scratches, gouges or other surface blemishes. 6.33 Prior to loading, the lifting harness and sling shall be carefully examined for proper placement cf strtps. If the container is equipped with the lift ring assembly, verify the placement cf the steel straps and ensure that the lift cables are not frayed. /~ 2 of 7 l

r-9 ,Nov 187'y7, 3: 45 FROM BRP DECOM PLAN PAGE.017 - OM.16.Ws, Rev. 6 October 2,1995 l-6.3.4. If the container is to be loaded within a process shield or shipping cask, l care shall be taken to ensure sinooth entry and removal from.'he shield or i cask. I 6.3.5. Tb containermay be loaded per the requirementsof applicable regulations ifor on site safety as well as burial of low level radioactive waste. The j i temperature of the container may not exceed 170*F during or following l loading. i 6.3.6 The chemicals listed in Appendix A are not permitted in the container. Small amounts of these chemicals present in the waste to be disposed of shall be repor1ed to VECTRA prior to use to determine compatibility per the requirements of the state of South Carolina and 10 CFR 61. 1 6.3.7 Care shall be taken to ensure that the total gross weight of the container is below the max gross weight indicated on the side of the container. 1 i 6.3.8 - Ifliquids are introduced into the container, the contents shall be dewatered i to less than 1% free liquid by volume following loading, t 6.3.9 Install the gasket, seal lid and the closure tid. Tighten lid to a minimum of 50 ft.lbs. 6.3.10 After completion of Steps 6.3.1 through 6.3.9, the container will be ready for shiprnent to the dispond site. Prior to shipping, the ' shipper shall certify that tne VECTRA Polyethylene HIC has been used according to the State of South Carolina certificate of compliance. Complete appropriate Attachment (2 or 3), depending upon disposal site chosen. i 6.3.11 At the disposal site the container may be lifted by the lifting sling and plaud into the burial trench. The sling msy be rel:ased after the container has been placed in its final position for burial. 5 3 of 7 4

,N O U. 18, 9 7, 9: 45 FROM BRP DECOM PLAN PAGE.018 OM 16.WS, Rev. 6 O m 2,1995 APPENDIX A MATERIALS NOT COMPATIBLE WITH CLPE Aliphatic Hydrocarbons (hexane, octane, Gasoline hexene, octene, etc.) Acetone Iodine Amyl Acetate Amyl Chloride Methyl Bromide Anihne Methyl Chloride Aqua Regia Methyl Ethyl Ketone.(MEK) Methylene Chloride Benzene Moist Chloriru Gas Bromine Liquid Butane Nitric Acid (50 % weight cor..entration) Camphor Oil Carbon Disu15de Organic Peroxides Carbon Tetrachloride Octyi Cresol Chlorine Liquid Oleic Acid Moist Chlorine Gas Oleum Chlorobenzene Chloroforrn Pentane Chlorosulfonic Acid Petroleum Ether Chromic / Sulfuric Acid Phenol Cyclohexanone Propane Propylene Dichloride Dibutyl Phthalate Dimethylamine Sulfuric Mid (60 % weight Diesel Fuel concentration) Ethyl Acetate Tetrahydrofuran Ethyl Butyrate Tetralia Ethyl Chloride Titanium Tetrachloride Ethyl Ether Toluene Ethylene Chloride Trichloroethylene Ethylene Chlorohydrin Torpentine Ethylene Dichloride Xylene Fluorine Furfural Furfuryl Alcohol Fuel Oil g 4 of 7

-N O U 18 '97 9: 45' FROM BRP DECOM PLAN PAGE.019 OM 16.WS, Rev. 6 October 1,1995 ATTACHMENT 1 USER CERTIFICATE I VECTRA hereby authorizes --- herein after referred to as USER, to utilize the VECTRA Polyethylene High Integrity Container for disposal for low level radioactive waste. USER agrees to comply with the South Carolina Department of Health and Environmental Controls Bureau of Radiological Health. Certificate of Compliance No. DHEC-HIC-PL-012 and all amendments thereto. USER also agrees to comply with all requirements delineated in l VECTRA Procedure OM-16-WS, as well as all revisions and amendments thereto. i USER hereby certifies that its use of such containers as referenced above has complied with all applicable regulations and procedures as referenced above. COMPANY: - TTILE: DATE: l 5 of 7

r -- l 440V 18 '97 9: 46 FROM BRP DECOM PLAN PAGE.020 O M.16 W S,Rev.6 October 2,1995 i . ATTACHMENT 2 ( BARNWELL DISPOSAL Certification Statement for Disposal of VECTRA Inc. Polyethylene Container. For the VECIRA Polyethylene High Integrity Container to be disposed of at Chem Nuclear's Earnwell, South Cvolina Low Level Radioactive Waste Burial Facility and identified by Serial j No. ( ,J. Company - . hereby certirm that its use of such containers has complied with the certain South Carolina Department of Health and Environmental Controls bereau of Radiological Health Certificase of Compliance No. DHEC-HIC-PL-012 and all amendments thereto. Company: By: l

Title:

Dated: 1 l l 1 6 of 7

I .NOV-18 '9J,Jt46 FROM BRP DECOM PLAN PAGE.021 l ^ ' OM 16.WS, Rev. 6 october 2,1995 ATTACHMENT 3 l HANFORD DISPOSAL Certification Statement for Disposal of VECTRA Inc. Polyethylene Container, j i For the VECTRA Polyethylene High Integrity Container to be disposed of at U.S. Ecology's Richland (Hanford), Washington Low Level Radioactive Waste Burial Facility and identified byl Serial No. f 1, Company . hereby certifies that its use of such containers has complied with the certain South Carolina Department of Health and Environmental Controls Bureau of Radiological Health Certificate of Compliance No. i DHEC-HIC-PIA)12 and all amendments thereto. i Company: By: I l

Title:

Dated: l l l 1 t l 7of7 } l l

i j .NOV 18 '97 9: 46,'FROM BRP DECOM PLAN PAGE.022 l L Appendix C r l l l 1 l Teledyne Brown Engineering Environmental Services " Report of Analysis" 7/21/95 l l l-I l l l l t t

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616 547 8340 I40V 1897 9:23 FROM BRP DECOM PLAN PAGE.001 i FAX TRANSMITTAL CONSUMERS ENERGY COMPANY BIG ROCK PoldT PLANT IO269 US-31 ^ NORTH CHARLEVOIX MI 49720 FAX SMiBER [616] 547-8340 Internal Access: 1371-340 DATE: _n te 97 TO: NAME: PAUL W HARRIS courrNv: US NRC/NRR (301) 415-1169 OFFICE PHONE NUMBER: (301) 415-3313 FAX NUMBER: FROM: M N6_ Bourassa NAME: (616) 547-8244 omce FirosE Nuunsa: PAGES TO FOLLOW - 3I 7aea-~wy Res/bosc - [bc scTG @ES?owSE To NL Lo ud '9

noU 18 '97, 9:23 FROM BRP DECOM PLAN PAGE.002 l NRC-Requested Analysis: Shine from Dry Pool PURPOSE This document evaluates the exposure from the spent fuel pool at Big Rock Point closest site boundary (US Highway 31 at 805 meters distance), for the hypothetical case of loss of spent fuel pool coolant. SYSTEM DESCRIPTION There are a total of 441 bundles of fuel in the spent fuel pool. The dimensions of a bundle are 1 15 24 Cm (6) by 15.24 Cm (6") with an active fuellength of 177.8 Cm (70"). Weight is approximately 2 2SES gram (500 lb.) per bundle. The bundles are approximately evenly distnbuted within the storage area Decay time for 84 bundles (latest group removed from the reactor)is taken 93 days. which is the time af;er shutdown at which an emergency planning exemption from cer1ain offsite response requirements has been requested. The remaining 357 bundles were conservatively assumed to have decayed 458 days (one year plus 93 days). All of the 357 bundles actually have decayed longer than 458 days - some as long as 13 years at the time of this calculation. The spent fuel pool liner is made of stainless steet, with a dimension of 7 925E2 Cm (26 feet) long, by 6 096E2 Cm (20 feet) wide by 9 449E2 Cm (31 feet) deeo, with thickness of 1.905 Cm (0187 inch). The bottom of the !'ner is 2.54 Cm (one inch) of lead followed by 15 24 Cm (6 mehes) of concrete The liner is su Munded by concrete, With a minimum thickness of 1.067E2 Cm (3 5 feet) on the north side of the pool The spent fuel and all the reactor component are located inside of a steel sphere with a radius of 2000 Cm (65.6 feet), with the thsckness of 19 Cm (0.75 inch) on the bottom haff, and 1.27 Cm (0.5 in h) on the 100 half The spent fuel pool is centrally :ocated within the spisere. EXPOSURE EVALUATION Evaluation was performed for the exposure of the public at the site boundary due to shine from the fuelin the unhkely event that water shielding were lost. For purposes of the calculation using the program MicroSkyshine (Grove Engineenng), radionuclides present in the fuel were uniformly destnbuted six feet above the floor of the fuel poolin a layer of uranium matrix 0118 meters thick. This layer accounts for the total weight of the fuel bundles stored in the pool, and Conservabely spreads the radioactivity (1 1E+08 Ci represented by 26 major nuclides) in the thinnest layer possible so as to minimizo self shielding Vanous other geometries for the fuel mass were modeled This thin slab model was chosen because it provided the most conservative (highest) offsite oose results input and output data for MicroSkyshine are provided in Appendix A Direct exposure through the thinnest wall of the pool also was calculated, and is provided in AppeWx B in addition, scatter from the containment sphere itselt was evaluated and found to be negligible compared to air scatter due to the long path length through steel which a 90 degree scatte'ed photon must take to reach the sit) boundary l

l NOU 18 '97 9:24 FROM 3RP DECOM PLAN PAGE.003 ) CONCLUSION Calculations indicate that dose rate at the closest site boundary (US Highway 31) is 0.046 mrom/hr Technical Specifications proposed for the period in which fuel resides in the spent fuel pool require that fuel pool makeup be available within 24 hours of makeup loss. The integral dose over this interval would be 1 10 mrem (0.0011 rem) at the closest site boundary. This dose is well below the PAG of 1 rem. REFERENCES-1-Bechtel Drawing 0740G20142 Rev A l 2 Bechtel Drawing 0740G20143 Rev A 3-Bechtel Drawing 0740G40103 Rev E

4. Bechtet Drawing 0740G20142 Rev A 5 Bechtel Drawing 0740G40102 Rev D

6. Bechtel Drawing 0740G40102 Rev D

7. MicroSkyshine computer code. Grove Engineenng Inc. Rockville. MD l

i

f40V 18 '97 9:23, FROM BRP DECOM PLAN PAGE.004 j Appendix A ) Shine Exposure Due to the Lost of Spent Fuel Pool Coolant j l i l

NOV 18 '97 9:24 FROM BRP DECOM PLAN PAGE.005 Pega 2 CASE: Exposure due to spent fuel shine for even distrubution SOURCE NUCLIDES: Nuclide Curies Nuclide Curies Ba-137m 4.1100e+06 Ce-141 1.6300e+06 Ce-144 1.8500e+07 Cs-134 1.7400e+06 Cs-137 4.3900e+06 Eu-155 1.4400e+06 I-131 2.3500e+03 Kr-85 3.5300e+05 La-140 5.8400e+04 Nb-95 7.4900e+06 No-95m 9.5600e+04 Pm-147 6.5900e+06 Pr-143 9.7000e+04 Pr-144 1.8500e+07 Rh-103m 1.9000e+06 Rh-106 1.2000e+07 Ru-103 1.9400e+06 Ru-106 1.2000e+07 Sb-125 2.4800e+05 Sr-89 1.5600e+06 Sr-90 3.0300e+06 Te-127 9.1700e+04 Te-127m 9.2100e+04 Y-90 3.0300e+06 Y-91 2.6400e+06 Zr-95 4.5000e+06 RESULTS OF SENSITIVITY STUDY ON DIMENSION X - REFERENCE CASE: Reference case value: 20 Group Energy Activity Dose point Dose rate (mov) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 4.658e-21 1.036e+02 2 1.46 9.346e+15 4.337e-21 1.672e+02 3 1.08 1.168e+16 4.141e-21 1.995e+02 4 .77 5.049e+17 2.353e-21 4.898e+03 5 .64 2.797e+17 1.865e-21 2.150e+03 6 .50 1.616e+17 9.223e-22 6.145e+02 7 .32 7.747e+14 3.717e-23 1.187e-01 8 .24 9.637e+14 2.408e-25 9.568e-04 9 .15 2.985e+16 4.671e-35 5.750e-12 10 .13 8.4999+16 2.632e-34 9.225e-11 11 12 13 i 14 15 16 ('2' 17 18 y 19 Q* 20 p TOTALS: 1.0890+18 8.133e+03 i

, NOU 18 '9_7,,9:25 FROM BRP DECOM PLAN PAGE 006 \\. Pcge 3 RESULTS FOR SENSITIVITY ITERATION 1 OF 10 (DIMENSION X = 20): Group Energy Activity Dose point Dose rate (nev) (photons /sec) rads / photon (mr/hr) l l 1 2.19 5.391e+15 4.658e-21 1.036e+02 l 2 1.46 9.346e+15 4.337e-21 1.672e+02 l 3 1.08 1.168e+16 4.141e-21 1.995e+02 4 .77 5.048e+17 2.353e-21 4.898e+03 5 .64 2.797e+17 1.865e-21 2.150e+03 l 6 .50 1.616e+17 9.223e-22 6.145e+02 l 7 .32 7.747e+14 3.717e-23 1.187e-01 8 .24 9.637e+14 2.408e-25 9.568e-04 l 9 .15 2.985e+16 4.671e-35 5.750e-12 10 .13 8.499e+16 2.632e-34 9.225e-11 11 12 13 14 15 16 17 18 19 20 TOTALS: 1.089e+18 8.133e+03 J RESULTS FOR SENSITIVITY ITERATION 2 OF 10 (DIMENSION X = 106.667): Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (ar/hr) 1 2.19 5.391e+15 4.876e 1.084e+01 2 1.46 9.346e+15 4.486e-22 1.7290+01 3 1.08 1.168e+16 4.258e-22 2.051e+01 4 .77 5.048e+17 2.386e-22 4.968e+02 5 .64 2.797e+17 1.825e-22 2.104e+02 6 .50 1.616e+17 8.728e-23 5.815e+01 7 .32 7.747e+14 3.313e-24 1.058e-02 8 .24 9.637e+14 1.729e-26 6.871e-05 9 .15 2.985e+16 2.926e-36 3.602e-13 10 .13 8.499e+16 1.629e-35 5.707e-12 11 12 13 14 15 16 17 b;f 18 f 19 TOTALS: 1.089e+18 8.140e+02

NOV 18 '97 9:25 FROM BRP DECOM PLAN PAGE.007 paga 4 RESULTS FOR SENSITIVITY ITERATION 3 OF 10 (DIMENSION X = 193.333): Group Energy Activity Dose point Dose rate (mov) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 1.236e-22 2.748e+00 2 1.46 9.346e+15 1.112e-22 4.287e+00 3 1.08 1.168e+16 1.024e-22 4.934e+00 4 .77 5.048e+17 5.677e-23 1.182e+02 5 .64 2.797e+17 4.085e-23 4.711e+01 6 .50 1.616e+17 1.875e-23 1.249e+01 7 .32 7.747e+14 6.893e-25 2.202e-03 8 .24 9.637e+14 2.964e-27 1.178e-05 9 .15 2.985e+16 4.357e-37 5.364e-14 10 .13 8.4990+16 2.357e-36 8.260e-13 11 12 13 14 15 16 17 18 19 20 TOTALS: 1.089e+18 1.897e+02 RESULTS FOR SENSITIVITY ITERATION 4 OF 10 (DIMENSION X = 280): i Group Energy Activity Dose point Dose rate (mov) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 3.736e-23 8.306e-01 2 1.46 9.346e+15 3.258e-23 1.256e+00 3 1.08 1.168e+16 2.881e-23 1.388e+00 4 .77 5.048e+17 1.575e-23 3.278e+01 5 .64 2.797e+17 1.071e-23 1.235e+01 6 .50 1.616e+17 4.716e-24 3.142e+00 7 .32 7.747e+14 1.681e-25 5.371e-04 8 .24 9.637e+14 6.229e-28 2.475e-06 9 .15 2.985e+16 7.855e-38 9.669e-15 10 .13 8.499e+16 4.095e-37 1.435e-13 11 12 13 14 i 15 t 16 17 18 19 20 TOTALS: i 589e+i8 s!i7se+5i L l

NOV 18_'97_ 9:25 FROM BRP DECOh PLAN PAGE.008 Page 9 RF.SULTS FOR. SENSITIVITY ITERATION 5 OF 10 (DIMLNSION X = 366.667): Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 1.214s-23 2.699e-01 2 1.46 9.346e+15 1.022e-23 3.937e-01 3 1.08 1.168e+16 8.638e-24 4.161e-01 4 .77 5.048e+17 4.643e-24 9.666e+00 5 .64 2.797e+17 2.996e-24 3.455e+00 J 6 .50 1.616e+17 1.265e-24 8.429e-01 7 .32 7.747e+14 4.363e-26 1.394e-04 8 .24 9.637e+14 1.445e-28 5.742e-07 9 .15 2.985e+16 1.560e-38 1.920e-15 10 .13 8.499e+16 7.764e-38 2.721e-14 11 12. 13 14 15 16 17 18 19 4 20 TOTALS: 1.089e+18 1.504e+01 RESULTS FOR SENSITIVITY ITERATION 6 OF 10 (DIMENSION X = 453.333): Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391o+15 4.112c-24 9.142e-02 2 1.46 9.346e+15 3.331e-24 1.284e-01 3 1.08 1.168e+16 2.687e-24 1.294e-01 4 .77 5.048e+17 1.417e-24 2.950e+00 5 .64 2.797e+17 8.716e-25 1.005e+00 6 .50 1.616e+17 3.529e-25 2.351e-01 7 .32 7.747e+14 1.173e-26 3.746e-05 8 .24 9.637e+14 3.579e-29 1.422e-07 9 .15 2.985e+16 3.333e-39 4.103e-16 10 .13 8.499e+16 1.570e-38 5.502e-15 11 12 13 14 15 16 17 18 f 19 20 l TOTALS: 1.089e+18 4.540e+00

l NOV 18,l97 9:26 FROM BRP DECOM PLAN PAGE.009 l Pcg3 6 l RESULTS FOR SENSITIVITY ITERATION 7 OF 10 (DIMENSION X = 540): Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 1.431e-24 3.182e-02 2 1.46 9.346e+15 1.114e-24 4.294e-02 3 1.08 1.168e+16 8.566e-25' 4.126e-02 1 4 .77 5.048e+17 4.425e-25 9.212e-01 5 .64 2.797e+17 2.605e-25 3.004e-01 6 .50 1.616e+17 1.011e-25 6.735e-02 7 .32 7.747e+14 3.225e-27 1.030e-05 8 .24 9.637e+14 9.290e-30 3.692e-08 9 .15 2.985e+16 7.577e-40 9.328e-17 10 .13 8.499e+16 3.357e-39 1.176e-15 11 12 13 14 15 16 1 17 18 19 20 TOTALS: 1.089e&18 1.405e+00 RESULTS FOR SENSITIVITY ITERATION 8 OF 10 (DIMENSION X = 626.667): Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 5.078e-25 1.129e-02 2 1.46 9.3460+15 3.796e-25 1.463e-02 l 3 1.08 1.168e+16 2.780e-25 1.339e-02 4 .77 5.048e+17 1.404e-25 2.923e-01 5 .64 2.797e+17 7.940e-26 9.157e-02 6 .50 1.616e+17 2.955e-26 1.969e-02 7 .32 7.747e+14 9.013e-28 2.879e-06 8 .24 9.637e+14 2.495e-30 9.914e-09 9 .15 2.985e+16 1.821e-40 2.241e-17 10 .13 8.499e+16 7.568e-40 2.652e-15 11 12 13 14 15 16 l 17 18 f 19 l 20 TOTALS: 1.089e+18 4.428e-01

140U 18 '97 9?26 FROM BRP DECOM PLAN PAGE.010 Paga 7 RESULTS FOR SENSITIVITY ITERATION 9 OF 10 (DIMENSION X = 713.333): l l Group Energy Activity Dose point Dose rate (nev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 1.828e-25 4.064e-03 l 2 1.46 9.346e+15 1.311e-25 5.053e-03 l 3 1.08 1.168e+16 9.149e-26 4.407e-03 4 .77 5.048e+17 4.509e-26 9.386e-02 5 .64 2.797e+17 2.459e-26 2.836e-02 6 .50 1.616e+17 8.783e-27 5.852e-03 7 .32 7.747e+14 2.549e-28 8.144e-07 8 .24 9.637e+14 6.871e-31 2.730e-09 9 .15 2.985e+16 4.591e-41 5.651e-18 l 10 .13 8.499e+16 1.794e-40 6.288e-17 l 11 12 13 14 15 l 16 17 18 19 20 1 TOTALS: 1.089e+18 1.416e-01 l RESULTS FOR SENSITIV'.TY ITERATION 10 OF 10 (DIMENSION X e 800): 1 Group Energy Activity Dose point Dose rate (mev) (photons /sec) rads / photon (mr/hr) 1 2.19 5.391e+15 6.665e-26 1.482e-03 2 1.46 9.346e+15 4.583e-26 1.766e-03 i 3 1.08 1.168e+16 3.048e-26 1.469e-03 4 .77 5.048e+17 1.463e-26 3.046e-02 5 .64 2.797e+17 7.721e-27 8.905e-03 6 .50 1.616e+17 2.648e-27 1.765e-03 l 7 .32 7.747e+14 7.280e-29 2.325e-07 l 8 .24 9.637e+14 1.929e-31 7.663e-10 9 .15 2.985e+16 1.207e-41 1.485e-18 10 .13 8.499e+16 4.460e-41 1.563e-17 11 12 i 13 14 15 16 17 18 19 20 TOTALS: 1.089e+18 4.585e-02

i l NOU 18 '97 9:27 FROM BRP DECOM PLAN PAGE.011 i \\ l i 1 i Appendix B 1 l I Direct Exposure Due to the { Lost of Spent Fuel Pool Coolant t

'N O U 1 8 _' 9,7,,9:27 FROM BRP DECOM. PLAN PAGE.012 MicroShiold v5.01 (5.01-00152) Consumers Energy aga

1 File Ref:

, DS File: NEDO2.MS5 Date: l un Date: November 17, 1997 By: un Time: 5:36:45 PM Checked: uration: 00:07: 11 Case

Title:

Exposure De2cription: Direct Exposure due to lost of spent fuel coolant at parking Geometry: 13 - Rectangular Volume Source Dimensions Length 764.4 cm 25 ft 0.9 in Width 588.0 cm 19 ft 3.5 in Height 11.82 cm 4.7 in Dose Points Y Y A

1 2000 cm 5.5 cm o cm 65 ft 7.4 in 2.2 in 0.0 in Z

Shields Shield Name Dimension gaterial Density Source 5.313 m3 Uraninm 18.7 Shield 1 .005 m Iron 7.86 shield 2 2.057 m Concrete 2.35 i Shield 3 .019 m Iron 7.86 Air Gap Air 0.00122 Source Input Grouping Method : Standard Indices Number of Groupa : 25 Lower Energy Cutoff : 0.015 Photons < 0.015 : Excluded Library : ICRP-38 Nuclidg cutigs beccuerels gi/cm3 Ba/cm3 Ba-137m 4.1100e+006 1.5207e+017 7.7362e+005 2.8624e+010 Ce 14? 1.6300e+006 6.0310e+016 3.0681e+005 1 1352e+010 Ce-144 1.8500e+007 6.8450e+017 3.4822e+006 1.2884e+011 Cs-134 1.7400e+006 6.4380e+016 3.2752e+005 1.2118e+010 Cs-137 4.3900e+006 1.6243e+017 8.2632e+005 3.0574e+010 Eu-155 1.4400e+006 5.3280e+016 2.7105e+005 1.0029e+010 I-131 2.3500e+003 8.6950e+013 4.4234e+002 1.6366e+007 Kr-85 3.5300e+005 1.3061e+016 c.6445e+004 2.4584e+009 La-140 8.4800e+004 3.1376e+015 1.5962e+004 5.9058e+008 Nb-95 7.4900e+006 2.7713e+017 1.4098e+006 5.2164e+010 Nb-95m 9.5600e+004 3.5372e+015 1.7995e+004 6.6580e4008 Pm-147 6.5900e+006 2.4383e+017 1.2404e+006 4.5896e+010 Pr-143 9.7000e+00; 3.5890e+015 1.8258e+004 6.7555e+008 Pr-144 1.8500e+00i 6.8450e+017 3.4822e+006 1.2884e+011 Rh-103m 1.9000e+006 7.0300e+016 3.5763e+005 1.3232e+010 l Rh-106 1.2000e+007 4.4400e+017 2.2587e+006 8.3573e+010 f Ru-103 1.9400e+006 7.1780e+016 3.6516e+005 1.3511e+010 l Ru-106 1.2000e+007 4.4400e+017 2.2587e+006 8.3573e+010 Sb-125 2.4800e+005 9.1760e+015 4.6681e+004 1.7272e+009 Sr-89 1.5600e+006 5.7720e+016 2.9364e+005 1.0865e+010

1 i40V 18j,*97 9:27 FROM BRP DECOM PLAN PAGE.013 l)OS Filo: NEDO2.MS5 ltun Date: November 17, 1997 [Mn Time: 5:36:45 PM l>uration: 00:07:11 l Nuclide curies becquerels uCi/cm3 Bc/cm3 l Sr-90 3.0300e+006 1.1211e+017 5.7033e+005 2.1102e+010 Te-127 9.1700e+004 3.3929e+015 1.7261e+004 6.3864e+008 Te-127m 9.2100e+004 3.4077e+015' 1.7336e+004 6.4142e+008 Y 3.0300e+006 1.1211e+017 5.7033e+005 2.1102e+010 Y-91 2.6400e+006 9.7680e+016 4.9692e+005 1.8386e+010 q Zr-95 4.5000e+006 1.6650e+017 8.4703e+005 3.1340e+010 l Buildup The material reference is : Source Integration Parameters X Direction 10 l Y Direction 20 Z Direction 20 Results Enerav Activity Eluence Rate Fluence Rate Exnosure Rate Exposure Rate l EgN chotons/ces MeV/cm8/sec ReV/cm2/sgg mR/hr mR/hr l No Buildup With_BuilduD No Builduo With Builduo l 0.015 1.686e+15 0.000e+00 6.803e-21 0.000e+00 5.835e-22 0.02 7.004e+15 0.000e+00 3.768e-20 0.000e+00 1.305e-21 0.03 1.718e+16 0.000e+00 1.396e-19 0.000e+00 1.384e-21 0.04 9.760e+16 0.000e+00 1.080e-18 0.000e+00 4.775e-21 0.05 4.177e+15 0.000e+00 5.911e-20 0.000e+00 1.575e-22 l 0.06 1.036e+15 0.000e+00 1.829e-20 0.000e+00 3.632e-23 0.08 3.015e+16 2.775e-289 7.603e-19 4.392e-292 1.203e-21 t 0.1 1.365e+16 2.571e-173 4.707e-19 3.933e-176 7.202e-22 j 0.15 1.029e+17 9.276e-226 2.647e-06 1.528e-228 4.360e-09 0.2 1.626e+15 7.871e-123 1.645e-19 1.389e-125 2.904e-22 0.3 9.456e+14 1.340e-58 1.264e-19 2.542e-61 2.397e-22 { 0.4 4.103e+15 6.621e-38 9 338e-19 1.290e-40 1.819e-21 0.5 1.569e+17 3.446e-27 5.687e-17 6.765e-30 1.116e-19 0.6 2.794e+17 9.867e-22 1.438e-16 1.926e-24 2.806e-19 l 0.8 5.066e+17 1.300e-15 6.214e-15 2.472e-18 1.182e-17 1.0 1.116e+16 7.014e-14 3.888e-13 1.293e-16 7.167e-16 1.5 8.056e+15 1.709e-09 1.624e-08 2.876e-12 2.732e-11 2.0 6.203e+15 2.246e-07 2.615e-06 3.474e-10 4.043e-09 3.0 1.900e+14 1.410e-06 1.982e-05 1.914e-09 2.689e-08 i ! TOTALS: 1.251e+18 1.637e-06 2.510e-05 2.264e-09 3.532e-08 Sensitivity Variable X Dose Point 1 (1 of 14) (20 m) 0.015 1.686e+15 0.000e+00 6.803e-21 0.000e+00 5.835e-22 0.02 7.004e+15 0.000e+00 3.768e-20 0.000e+00 1.305e-21 0.03 1.718e+16 0.000e+00 1.396e-19 0.000e+00 1.384e-21 0.04 9.760e+16 0.000e+07 1.060e-18 0.000e+00 4.775e-21 0.05 4.177e+15 0.000e+00 5.911e-20 0.000e+00 1.575e-22 0.06 1.036e+15 0.000e+00 1.829e-20 0.000e+00 3.632e-23 0.08 3.015e+16 2.775e-289 7.603e-19 4.392e-292 1.203e-21

i NOU 18 ] Sn', 9:28 FROM BRP DECOM PLAN PAGE.014 >OS File: NEDO2.MS5 tun Date: November 17, 1997 tun Time: 5:36:45 PM >urotion: 00:07:11 Energy Activity Fluence Rate Fluence Rate Exoosure Rate Exoosure Rate May photons /sec MeV/cm /sec MeV/cm2/sec mR/hr mR/hr a No Buildup With Builduo No Bulldup With Suilduo 0.1 1.365e+16 2.571e-173 4.707e-19 3.933e-176 7.202e-22 0.15 1.029e+17 9.276e-226 2.647e-06 1.528e-228 4.360e-09 0.2 1.626e+15 7.871e-123 1.645e-19 1.389e-125 2.904e-22 0.3 9.456e+14 1.340e-58 1.264e-19 2.542e-61 2.397e-22 0.4 4.103e+15 6.621e-38 9.338e-19 1.290e-40 1.819e-21 l 0.5 1.569e+17 3.446e-27 5.687e-17 6.765e-30 1.116e-19 l 0.6 2.794e+17 9.867e-22 1.438e-16 1.926e-24 2.806e-19 I 0.8 5.066e+17 1.300e-15 6.214e-15 2.472e-18 1.182e-17 L 1.0 1.116e+16 7.014e-14 3.888e-13 1.293e-16 7.167e-16 l 1.5 8.056e+15 1.709e-09 1.624e-08 2.876e-12 2.732e-11 2.0 6.203e+15 2.246e-07 2.615e-06 3.474e-10 4.043e-09 3.0 1.900e+14 1.410e-06 1.982e-05 1.914e-09 2.689e-08 j 1CTALS: 1.251e+18 1.637e-06 2.510e-05 2.264e-09 3.532e-08 Sensitivity variable X Dose Point 1 (2 of 14) (80 m) 0.015 1.686e+15 0.000e+00 3.038e-22 0.000e+00 2.606e-23 1 0.02 7.004e+15 0.000e+00 1.683e-21 0.000e+00 5.829e-23 ! 0.03 1.718e+16 0.000e+00 6.235e-21 0.000e+00 6.179e-23 0.04 9.760e+16 0.000e+00 4.821e-20 0.000e+00 2.1329-22 l 0.05 4.177e+15 0.000e+00 2.640e-21 0.000e+00 7.032e-24 1 0.06 1.036e+15 0.000e+00 8.167e-22 0.000e+00 1.6220-24 0.08 3.015e+16 1.416e-290 3.396e-20 2.240e-293 5.373e-23 0.1 1.365e+16 1.319e-174 2.102e-20 2.017e-177 3.216e-23 ! 0.15 1.029e+17 5.619e-227 1.182e-07 9.253e-230 1.947e-10 0.2 1.626e+15 4.612e-l'4 7.348e-21 8.141e-127 1.297e-23 0.3 9.456e+14 7.118e-E9 5.644e-21 1.350e-62 1.071e-23 0.4 4.103e+15 3.306e-39 4.170e-20 6.441e-42 8.125e-23 L 0.5 1.569e+17 1.664e-28 2.540e-18 3.265e-31 4.985e-21 0.6 2.794e+17 4.670e-23 6.421e-18 9.116e-26 1.253e-20 0.8 5.066e+17 6.021e-17 2.875e-16 1.145e-19 5.469e-19 1.0 1.116e+16 3.209e-15 1.779e-14 5.915e-18 3.280e-17 1.5 8.056e+15 7.688e-11 7.297e-10 1.293e-13 1.228e-12 L 2.0 6.203e+15 1.004e-08 1.165e-07 1.552e-11 1.802e-10 3.0 1.900e*14 6.272e-08 8.757e-07 8.510e-11 1.188e-09 TOTALS: 1.251e+18 7.284e-08 1.111e-06 1.007e-10 1.564e-09 l i sensitivity variable X Dose Point 1 (3 of 14) (140 m) 0.015 1.686e+15 0.000e+00 9.503e-23 0.000e+00 8.151e-24 0.02 7.004e+15 0.000e*00 5.264e-22 0.000e+00 1.823e-23 0.03 1.718e+16 0.000e+00 1.950e-21 0.000e+00 1.933e-23 0.04 9.760e+16 0.000e+00 1.508e-20 0.000e+00 6.670e-23 0.05 4.177e+15 0.000e+00 8.257e-22 0.000e+00 2.200e-24 0.06 1.036e+15 0.000e+00 2.555e-22 0.000e+00 5.074e-25 0.08 3.015e+16 1.852e-291 1.062e-20 2.931e-294 1.681e-23 'O.1 1.365e+16 1.679e-175 6.576e-21 2.568e-178 1.006e-23 l l

NOU,18 '97-9:29 FROM BRP DECOM PLAN PAGE.015 -y> 0,5 File: NEDO2.MS5-un Date: November 17, 1997 un Time: 5:36:45 PM uration: 00:07:11 Enerav Activity Fluence Rate Fluence Rate Exoosure Rate Exoosure Rate f liey photons /sec }igy/cm 2 /sec MeV/cm /sec mRlbr mR/hr a No Buildup Mith Buildun No Buildun With Builduo j 0.15 1.029e+17 8.525e-228 , 3.698e-08 1.404e-230 6.090e-11 l 0.2 1.626e+15 6.816e-125 2.299e-21 1.203e-127 4.057e-24 0.3 9.456e+14 1.085e-60 1.766e-21 2.059e-63 3.349e-24 0.4 4.103e+15 5.302e-40 1.305e-20 1.033e-42 1.542e-23 l 0.5 1.569e+17 2.795e-29 7.945e-19 5.487e-32 1.559e-21 i 0.6 2.794e+17 8.163e-24 2.009e-18 1.593e-26 3.920e-21 0.8 5.066e+17 1.120e-17 5.628e-17 2.130e-20 1.070e-19 1.0 1.116e+16 6.254e-16 3.481e-15 1.153e-18 6.416e-18 1.5 8.056e+15 1.622e-11 1.550e-10 2.729e-14 2.608e-13 2.0 6.203e+15 2.226e-09 2.607e-08 3.442e-12 4.032e-11 l 3.0' 1.900e+14 1.477e-08 2.084e-07 2.003e-11 2.828e-10 30TALS: 1.251e+18 1.701e-08 2.716e-07 2.350e-11 3.d42e-10 l Sensitivity Variable X Dose Point 1 (4 of 14) (200 m) ! 0.015 1.686e+15 0.000e+00 4.579e-23 0.000e+00 3.927e-24 0.02 7.004e+15 0.000e+00 2.536e-22 0.000e+00 8.786e-24 0.03 1.718e+16 0.000e+00 9.397e-22 0.000e+00 9.313e-24 0.04 9.760e+16 0.000e+00 7.267e-21 0.000e+00 3.214e-23 0.05 4.177e+15 0.000e+00 3.978e-22 0.000e+00 1.060e-24 1 0.06 1.036e+15 0.000e+00 1.231e-22 0.000e+00-2.445e-25 0.08 3.015e+16 2.984e-292 5.118e-21 4.722e-295 8.099e-24 j 0.1 1.365e+16 2.814e-176 3.168e-21 4.3 5e-179 4.847e-24 0.15 1.029e+17 1.645e-228 1.782e-08 2.709e-231 2.934e-11 O.2 1.626e+15 1.38?e-125 1.107e-21 2.439e-128 1.955e-24 J 0.3 9.456e+14 2.420e-61 8.507e-22 4.591e-64 1.614e-24 0.4 4.103e+15 1.272e-40 6.285e-21 2.479e-43 1.225e-23 0.5 1.569e+17 7.104e-30 3.828e-19 1.394e-32 7.513e-22 06 2.794e+17 2.172e-24 9.677e-19 4.239e-27 1.889e-21 i 0.8 5.066e+17 3.193e-18 1.739e-17 6.073e-21 3.307e-20 1.0 1.116e+16 1.877e-16 1.048e-15 3.460e-19 1.933e-18 ( 1.5 8.056e+15 5.298e-12 5.101e-11 8 914e-15 8.582e-14 f 2.0 6.203e+15 7.662e-10 9.060e-09 1.185e-12 1.401e-11 3.0 1.900e+14 5.412e-09 7.727e-08 7.342e-12 1.048e-10 ,.90 t A L S : 1.251e+18 6.183e-09 1.042e-07 8.536e-12 1.483e-10 Sensitivity Variable X Dose Point 1 (5 of 14) (260 m) 0.015 1.686e+15 0.000e+00 2.685e-23 0.000e+00 2.303e-24 0.02 7.004e+15 0.000e+00 1.487e-22 0.000e+00 5.152e-24 0.03 1.718e+16 0.000e+00 5.511e-22 0.000e+00 5.462e-24 0.04 9.760e+16 0.000e+00 4.261e-21 0.000e+00 1.885e-23 0.05 4.177e+15 0.000e+00 2.333e-22 0.000e+00 6.215e-25 0.06 1.036e+15 0.000e+00 7.218e-23 0.000e+00 1.434e-25 0.08 3.015e+16 5.546e-293 3.001e-21 8.776e-296 4.749e-24 0.1 1.365e+16 5.561e-177 1.858e-21 8.508e-180 2.843e-24 0.15 1.029e+17 3.702e-229 1.045e-08 6.097e-232 1.721e-11

NOU._I,G.'. 9 7.. 9:29 'FROM 3RP DECOM PLAN PAGE.016 K:!S File: NEDO2. MSS

un Date: November 17, 1997
un Time: 5:36:45 PM lPuration: 00:07:11 Engrgy Activity Fluence Rat,g Eluence Rate Exoosure Rate Excesure Rate May ohotons/see MeV/cm2/sec MeV/cm /sec mR/hr mR/hr a

No Builduc With__ Buildup No Buildup With Buildup 0.2 1.626e+15 3.338e-126 6.494e-22 5.892e-129 1.146e-24 0.3 9.456e+14 6.522e-62 4.989e-22 1.237e-64 9.463e-25 i 0.4 4.103e+15 3.706e-41 3.686e-21 7.221e-44 7.181e-24 l 0.5 1.569e&l7 2.196e-30 2.245e-19 4.311e-33 4.406e-22 0.6 2.794e+17 7.038e-25 5.675e-19 1.374e-27 1.108e-21 l 0.8 5.066e+17 1.110e-18 6.822e-18 2.112e-21 1.298e-20 1.0 1.116e+16 6.875e-17 3.855e-16 1.267e-19 7.105e-19 1.5 8.056e+15 2.114e-12 2.051e-11 3.557e-15 3.450e-14 2.0 6.203e+15 3.224e-10 3.848e-09 4.986e-13 5.951e-12 3.0 1.900e+14 2.426e-09 3.504e-08 3.291e-12 4.754e-11 ! TOTALS: 1.251e+18 2.750e-09 4.936e-08 3.793e-12 7.073e-11 Sensitivity Variable X Dose Point 1 (6 of 14) (320 m) 0.015 1.686e+15 0.000e+00 1.763e-23 0.000e+00 1.512e-24 j 0.02 7.004e+15 0.000e+00 9.764e-23 0.000e+00 3.382e-24 1 0.03 1.718e+16 0.000e+00 3.618e-22 0.000e+00 3.585e-24 0.04 9.760e+16 0.000e+00 2.797e-21 0.000e+00 1.237e-23 l 0.05 4.177e+15 0.000e+00 1.532e-22 0.000e+00 4.080e-25 l G.06 1.036e+15 0.000e+00 4.739e-23 0.000e+00 9.412e-26 j 0.08 3.015e+16 1.134e-293 1.970e-21 1.794e-296 3.118e-24 0.1 1.365e+16 1.218e-177 1.220e-21 1.864e-180 1.866e-24 l 0.15 1.029e+17 9.206e-230 6.860e-09 1.516e-232 1.130e-11 0.2 1.626e+15 8.971e-127 4.263e-22 1.583e-129 7.525e-25 i 0.3 9.456e+14 1.964e-62 3.275e-22 3.726e-65 6.212e-25 l 0.4 4.103e+15 1.209e-41 2.420e-21 2.355e-44 4.714e-24' O.5 1.569e+17 7.607e-31 1.474e-19 1.493e-33 2.892e-22 0.6 2.794e+17 2.556e-25 3.725e-19 4.989e-28 7.272e-22 0.8 5.066e+17 4.329e-19 3.165e-18 8.235e-22 6.020e-21 1.0 1.116e+16 2.824e-17 1.589e-16 5.206e-20 2.929e-19 1.5-8.056e+15 9.467e-13 9.248e-12 1.593e-15 1.556e-14 2.0 6.203e+15 1.523e-10 1.834e-09 2.354e-13 2.836e-12 3.0 1.900e+14 1.220e-09 1.783e-08 1.655e-12 2.419e-11 l l TOTALS: 1.251e+18 1.373e-09 2.653e-08 1.893e-12 3.834e-11 Sensitivity variable X Dose Point 1 (7 of 14) (380 m) 0.015 1.686e+15 0.000e+00 1.245e-23 0.000e+00 1.068e-24 0.02 7.004e+15 0.000e+00 6.898e-23 0.000e+00 2.389e-24 0.03 1.718e+16 0.000e+00 2.556e-22 0.000e+00 2.533e-24 0.06 9.760e+16 0.000e+00 1.976e-21 0.000e+00 8.740e-24 0.05 4.177e+15 0.000e+00 1.082e-22 0.000e+00 2.882e-25 0.06 1.036e+15 0.000e+00 3.347e-23 0.000e+00 6.649e-26 0.08 3.015e+16 2.477e-294 1.392e-21 3.920e-297 2.202e-24 0.1 1.365e+16 2.861e-178 8.616e-22 4.377e-181 1.318e-24 0.15 1.029e+17 2.450e-230 4.846e-09 4.035e-233 7.980e-12 0.2 1.626e+15 2.589e-127 3.012e-22 4.569e-130 5.316e-25 l t

L l ~

N O V 18 '97 9:30 FROM BRP DECOM PLAN PAGE.Ol7 20$ File: NEDO2.MS5 i

16n Date: Novcmber'17, 1997 itun Time: 5:36:45 PM [>urotion: 00:07:11 Engr 2y. Activity Elvence Rate Fluence Rate Exnosure Rate Exoosure Rate i May chotons/see MeV/cm2/sec Mev/cm2/see mR/hz mR/hr Hg Builduo Nith Buildup No Builduo Nith Builduo l 0.3 9.456e+14 6.366e-63 2.313e-22 1.208e-65 4.388e-25 l 0.4 4.103e+15 4.243e-42 1.709e-21 8.268e-45 3.330e-24 l 0.5 1.569e+17 2.837e-31 1.041e-19 5.569e-34 2.043e-22 ) i 0.6 2.794e+17 9.998e-26 2.632e-19 1.952e-28 5.137e-22. I i 0.8 5.066e+17 1.818e-19 1.684e-18 3.459e-22 3.204e-21 l 1.0 1.116e+16 1.250e-17 7.059e-17 2.304e-20 1.301e-19 l 1.5 8.056e+15 4 567e-13 4.494e-12 7.684e-16 7.561e-15 2.0 6.203e+15 7.746e-11 9.418e-10 1.198e-13 1.456e-12 j 3.0 1.900e+14 6.615e-10 9.776e-09 8.974e-13 1.326e-11 ' TOTALS: 1.251e+18 7.394e-10 3.557e-08 1.010e-12 2.271e-11 Sensitivity Variable X Dose Point 1 (8 of 14 (440 m)

0.015 1.686e+15 0.000e+00 9.262e-24 0.000e+00 7.944e-25 0.02 7.004e+15 0.000e+00 5.130e-23 0.000e+00 1.777e-24 0.03 1.718e+16 0.000e+00 1.901e-22 0.000e+00 1.884e-24 O.04 9.'r 'o+16 0.000e+00 1.470e-21 0.000e+00 6.501e-24 0.05 4.1 a 15 0.000e+00 8.047e-23 0.000e+00 2.144e-25 0.06 1.030e+15 0.000e+00 2.490e-23 0.000e+00 4.945e-26 l

0.08 3.015e+16 5.679e-295 1.035e-21 8.987e-298 1.638e-24 I 0.1 1.365e+16 7.064e-179 6.409e-22 1.001e-181 9.805e-25 0.15 1.029e+17 6.849e-231 3.604e-09 1.128e-233 5.936e-12 0.2 1.626e+15 7.857e-128 2.240e-22 1.387e-130 3.954e-25 0.3 9.456e+14 2.172e-63 1.721e-22 4.120e-66 3.264e-25 0.4 4.103e+15 1.569e-42 1.271e-21 3.057e-45 2.477e-24 O.5 1.569e+17 1.115e-31 7.743e-20 2.188e-34 1.520e-22 0.6 2.794e+17 4 120e-26 1.958e-19 8.042e-29 3.821e-22 0.8 5.066e+17 8.047e-20 1.009e-18 1.531e-22 1.919e-21 1.0 1.116e+16 5.828e-18 3.303e-17 1.074e-20 6.089e-20 1.5 8.056e+15 2.3220-13 2.301e-12 3.906e-16 3.871e-15 { 2.0 6.203e+15 4.153e-11 5.096e-10 6.422e-14 7.880e-13 1 3.0 1.900e+14 3.779e-10 5.648e-09 5.126e-13 7.663e-12 00TALS: 1.251e+18 4.196e-10 9.765e-09 5.772e-13 1.439e-11 Sensitivity variable X Dose Point 1 (9 of 14) (500 r.) ! 0.015 1.686e+15 0.000e+00 7.157e-24 0.000e+00 6.139e-25

0.02 7.004e+15 0.000e+00 3.965e-23 0.000e+00 1.373e-24 l 0.03 1.718e+16 0.000e+00 1.469e-22 0.000e+00 1.456e-24

'10.04 9.760e+16 0.000e+00 1.136e-21 0.000e+00 5.024e-24 0.05 4.177e+15 0.000e+00 6.219e-23 0.000e+00 1.657e-25 0.06 1.036e+15 0.000e+00 1.924e-23 0.000e+00 3.822e-26 -0.08 3.015e+16 1.350e-295 8.000e-22 2.137e-298 1.266e-24 i 0.1 1.365e+16 1.810e-179 4.953e-22 2.769e-182 7.577e-25 I 0.15 1.029e+17 1.986e-231 2.785e-09 3.271e-234 4.587e-12 0.2 1.626e+15 2.477e-128 1.731e-22 4.371e-131 3.055e-25 0.3 9.456e+14 7.701e-64 1.330e-22 1.461e-66 2.522e-25 1

i 'HOU 18 '97 9:31 FROM DRP DECOM PLAN PAGE.018 OS File: NEDO2. MSS i un Dato: Novembar 17, 1997 iun Time: 5:36:45 PM !urction: 00:07:11 i Enercy Activity Fluence Rate Fluence Rate Exoosure Rate Exoosure Rate HAV ohotons/sec MeV/cm2/sec MeV/cm2/sec mR/hr mR/hr No Buildup With Duildtm ?]p_Builduo Nith Builduo 0.4 4.103e+15 6.030e-43 9.825e-22 1.175e-45 1.914e-24 I 0.5 1.569e+17 4.551e-32 5.983e-20 8.934e-35 1.174e-22 0.6 2.794e+17 1.765e-26 1.513e-19 3.444e-29 2.953e-22 0.8 5.066e+17 3.701e-20 6.671e-19 7.040e-23 1.269e-21 1.0 1.116e+16 2.825e-18 1.607e-17 5.207e-21 2.962e-20 1.5 8.056e+15 1.227e-13 1.225e-12 2.064e-16 2.060e-15 2.0 6.203e+13 2.315e-11 2.866e-1C 3.579e-14 4.431e-13 3.0 1.900e+14 2.244e-10 3.392e-09 3.044e-13 4.602e-12 ?OTALS: 1.251e+18 2.476e-10 6.465e-09 3.40le-13 9.634e-12 Sensitivity Variable X Dose Point 1 (10 of 14) (560 m) 0.015 1.686e+15 0.000e+00 5.696e-24 0.0000+00 4.856e-25 0.02 7.004e+15 0.000e+00 3.155e-23 0.000e+00 1.093e-24 O.03 1.718e+16 0.000e+00 1.169e-22 0.000e+00 1.159e-24 O.04 9.760e+16 0.000e+00 9.040e-22 0.000e+00 3.998e-24 0.05 4.177e+15 0.000e+00 4.949e-23 0.000e+00 1.318e-25 0.06 1.036e+15 0.000e+00 1.531e-23 0.000e+00 3.042e-26 0.08 3.015e+16 3.303e-296 6.367e-22 5.226e-299 1.008e-24 0.1 1.365e+16 4.775e-180 3.942e-22 7.306e-183 6.030e-25 0.15 1.029e+17 5.930e-232 2.217e-09 9.765e-235 3.651e-12 0.2 1.626e+15 8.'040e-129 1.378e-22 1.419e-131 2.432e-25 0.3 9.456e+14 2.813e-64 1.058e-22 5.335e-67 2.008e-25 0.4 4.103e+15 2.387e-43 7.819e-22 4.651e-46 1.524e-24 0.5 1.569e+17 1.915e-32 4.762e-20 3.758e-35 9.347e-23 O.6 2.794e+17 7.786e-27 1.204e-19 1.520e-29 2.350e-22 0.8 5.066e+17 1.754e-20 4.776e-19 3.336e-23 9.084e-22

7. 0 1.116e+16 1.411e-18 8.056e-18 2.600e-21 1.485e-20 L.5 8.056e+15 6.679e-14 6.715e-13 1.124e-16 1.130e-15 2.0 6.203e+15 1.329e-11 1.660e-10 2.055e-14 2.567e-13 3.0 1.900e+14 1.373e-10 2.099e-09 1.862e-13 2.847e-12

'OTALS: 1.251e+18 1.506e-10 4.482e-09 2.069e-13 6.756e-12 Sensitivity Variable X Dose Point 1 (11 of 14) (620 m) i0.015 1.686e+15 0.000e+00 4.641e-24 0.000e+00 3.981e-25 l 0.02 7.004e+15 0.000e+00 2.571e-23 0.000e+00 8.905e-25 O.03 1.718e+16 0.000e+00 9.525e-23 0.000e+00 9.440e-25 l 0.04 9.760e+16 0.000e+00 7.365e-22 0.000e+00 3.257e-24 0.05 4.177e+15 0.000e+00 4.032e-23 0.000e+00 1.074e-25 l 0.06 1.036e+15 0.000e+00 1.248e-23 0.000e+00 2.478e-26 0.08 3.015e+16 8.265e-297 5.187e-22 1.308e-299 8.209e-25 j 0.1 1.365e+16 1.289e-180 3.211e-22 1.973e-183 4.913e-25 0.15 1.029e+17 1.811e-232 1.806e-09 2.983e-235 2.974e-12 0.2 1.626e+15 2 671e-129 1.123e-22 4.714e-132 1.981e-25 0.3 9.456e+14 1.052e-64 8.622e-23 1.995e-67 1.636e-25 0.4 4.103e+15 9.676e-44 6.371e-22 1.885e-46 1.241e-24 t i ~

r i j"NOV 1897 9:31 'FROM BRP DECOM PLAN PAGE.019 3QS Files NEDO2.MS5 Run Date: November 17, 1997 Run. Time: 5:36:45 PM Jurotion: 00:07:11

j. EDerav' Activity Fluence Rate Fluence RatJt Exnosure Rate ExDosure Rate l

May ohotons/sec MeV/cm2/see MeV/cm3/see mR/hr mR/hr l No Builden With,Buildun No Buildun With Buildup { i 0.5 '1.569e+17 8.246e-33 3.880e-20 1.619e-35 7.615e-23 0.6 2.794e+17. 3.517e-27 9.809e-20 6.866e-30 1.915e-22 i 0.8 5.066e+17 8.510e-21 3.632e-19 1.619e-23 6.908e-22 p l 1.0 1.116e+16 7.212e-19 4.137e-18 1.329e-21 7.625e-21 115 9.056e+15 3.723e-14 3.770e-13 6.264e-17 6.343e-16 2.0. 6.203e+15 7.814e-12 9.848e-11 1.208e-14 1.523e-13 3.0 .1.900e+14 8.600e-11 1~.329e-09 1.167e-13 1.804e-12 ! TOTALS: 1.251e+18 9.385e-11 '3.234e-09 1.288e-13 4.931e-12 Sensitivity Variable X Dose Point 1 (12 of 14) (680 m) I 0.015 .1.686e+15 0.000e+00 3.854e-24 0.000e+00 3.306e-25 0.02 7.004e+15 0.000e+00 2.135e-23 0.000e+00 7.395e-25

- 0.03 1.718e+16 0.000e+00 7.910e-23 0.000e+00 7.839e-25 0.04 9.760e+16 0.000e+00 6.116e-22 0.000e+00 2.705e-24 i

i 0.05 4.177e+15 0.000e+00 3.349e-23 0.000e+00 8.920e-26 0.06 1.036e+15 0.000e+00 1.036e-23 0.000e+00 2.058e-26 0.06 3.015e+16 2.107e-297 4.308e-22 3.335e-300 6.817e-25 0.1 1.365e+16 3.548e-181 2.667e-22 5.428e-184 4.080e-25 l 0.15 1.029e+17 5.638e-233 1.500e-09 9.284e-236 2.470e-12 0.2 1.626e+15 9.045e-130 9.322e-23 1.596e-132 1.645e-25 l 0.3 9.456e+14 4.000e-65 7.160e-23 7.602e-68 1.358e-25 0.4 4.103e+15 3.998e-44 5.290e-22 7.789e-47 1.031e-24 0.5 1.569e+17 3.621e.?3 3.222e-20 7.107e-36 6.324e-23 0.6 2.794e+17 1.620e-27 8.145e-20 3.162e-30 1.590e-22 0.8 5.066e+17 4.210e-21 2.888e-19 8.007e-24 5.493e-22 1.0 1.116e+16 3.759e-19 2.166e-18 6.930e-22 3.993e-21 1.5 0.056e+15 2.116e-14 2.157e-13 3.560e-17 3.629e-16 2.0 6.203e+15 4.684e-12 5.955e-11 7.243e-15 9.208e-14' 3.0 1.900e+14 5.492e-11 8.584e-10 7.452e-14 1.165e-12 < TOTALS: 1.251e+18 5.963e-11 2.418e-09 8.179e-14 3.727e-12 Sensitivity Variable X Dose Point 1 (13 of 14) (740 m) l 0.015 1.686e+15 0.000e+00 3.251e-24 0.000e+00 2.789e-25 0.02 7.004e+15 0.000e+00 1.801e-23 0.000e+00 6.238e-25 0.03 1.718e+16 0.000e+00 6.673e-23 0.000e+rd 6.G13e-25 0.04 9.760e+16 0.000e+00 5.160e-22 0.000e400 2.282e-24 0.05 4.177e+15 0.000e+00 2.825e-23 0.000e-00 7.526e-26 0.06 1.036e+15 0.000e+00 8.740e-24 0.000e+00 1.736e-26 O.08 3.015e+16 5.458e-298 3.634e-22 8.636e-301 5.751e-25 0.1 1.365e+16 9.917e-182 2.250e-22 1.517e-184 3.442e-25 0.15-1.029e+17 1.783e-233 1.265e-09 2.935e-236 2.084e-12 0.2 1.626e+15 3.112e-130 7.864e-23 5.492e-133 1.388e-25 l 0.3 9.456e+14 1.552e-65 6.040e-23 2.944e-68 1.146e-25 0.4 4.103e+15 1.678e-44 4.463e-22 3.270e-47 8.696e-25 0.5 1.569e+17 1.615e-33 2.718e-20 3.171e-36 5.335e-23 l l l

'*NOU'18 '97 9:32 FROM BRP DECOM PLAN PAGE.020 Oh Fileh NEDO2.MS5 uh Date: November 17, 1997 ,un. Time: 5:36:45 PM ' uration: 00:07:11 1 EnpyJ;04 ' Activity Fluence Rate Fluence Rate Excosure Rate Exoosure Rate Mgy .ohotons/see MeV/cm /sec MeV/cm8/see mR/hr mR/hr i a No Builduo' With Buildup No Builduo With Buildun 0.6 2.794e+17-7.581e-28 6.872e-20 1.480e-30 1.341e-22 0.8 5.066e+17 2.116e-21 2.372e-19 4.024e-24 4.511e-22 1.0 .1.116e+16 1.991e-19 1.154e-18 3.670e-22 2.127e-21 i 1.5 8.056e+15 1.222e-14 1.254e-13 2.056e-17 2.109e-16 2.0 6.203e+15 2.853e-12 3.658e-11 4.411e-15 5.657e-14 3.0 1.900e+14 3.564e-11 5.632e-10 4.836e-14 7.641e-13 ?OTALS: 1.251e+18 3.851e-11 1.865e-09 5.279e-14 2.?O5e-12 i Sensitivity Variable X Dose Point 1 (14 of 14) (800 m) 0.015 1.686e+15 0.000e+00 2.780e-24 0.000e+00 2.384e-25 i 0.02 7.004e+15 0.000e+00 1.540e-23 'O.000e+00 5.334e-25 t 0.03 1.718e+16 0.000e+00 5.705e-23 0.000e+00 5.654e-25 0.04 9.760e+16 0.000e+00 4.412e-22 0.000e+00. 1.951e-24 0.05 4.177e+15 0.000e+00 2.415e-23 0.000e+00 6.434e-26 0.06 1.036e+15 0.000e+00 7.473e-24 0.000e+00 1.484e-26 0.08 3.015e+16 1.432e-298 3.107e-22 2.266e-301 4.917e-25 0.1 1.365e+16 2.809e-182 1.923e-22 4.297e-185 2.943e-25 0.15 1.029e+17 5.711e-234 1.082e-09 9.404e-237 1.781e-12 i 0.2 1.626e+15 1.085e-130 6.723e-23 1.915e-133 1.187e-25 i 0.3 9.456e+14 6.090e-66 5.164e-23 1.155e-68 9.796e-26 0.4 4.103e+15 7.141e-45 3.816e-22 1.391e-47 7.435e-25 i ,0.5 1.569e+17 7.304e-34 2.324e-20 1.434e-36 4.561e-23 0.6 2.794e+17 3.595e-28 5.875e-2C 7.018e-31 1.147e-22 .0.8 5.066e+17 1.078e-21 2.014e-19 2.050e-24 3.830e-22 1.0 1.116e+16 1.069e-19 6.2380-19 1.970e-22 1.150e-21 1.5 8.056e+15 7.151e-15 7.385e-14 1.203e-17 1.242e-16 2.0 6.203e+15 1.761e-12 2.277e-11 2.723e-15 3.522e-14 3- 0 -1.900e+14 2.344e-11 3.745e-10 3.181e-14 5.080e-13 JOTALS: 1.251e*18 2.521e-11 1.479e-Oo 3.454e-14 2.325e-12

NOV 18,,'97, 9:32 FROM BRP DECOM PLAN-PAGE.021 Nuclida Dec;y CElcs for Shins Dos 3 ~~ ~84 Dundles13$7 Bundis 441 B'u'n~dif 1.0 Mwt 1.0 Mwt ~ -~~~ Nuclide - Ci @ 100d Ci @ 365d Lam.de/d_ Ci @ 93d Ci @ 458d Total Cl l' Ba-131Tn~ 3.32E+03 3.27E+03 0.000057 7.57E+05 3.32E+06 4.11 E+06 . 1.63E+06 C3-141 5.84E+03 2.03E+01 0.02136 1.63E+06 ' 2.84E+03 Co-144 2.75E+04 1.44E+04 ,,0.002431 6.71'E+06 1,.17E+07 1.85E+07, Cs-134 1.78E+03 1.40E+03 - 0.000906 4.30E+05 1.31 E+06 1.74E+06 Co-137 3.555'+03 3.49E+03 6~660064 8.52E+05 3.54E+06 4.39Eid6 ~ [Eu-155 1.53E+03 1.16E+03', 0.001058 3.70E+65 1.07E+06 1445 M l-131 5.36E+00 6.84E-10 0.085969 2.35E+03 2.35E-10 2.35E+03 Kr-85 2.925'+~0'2 2.81E+02. 0.dDD145 7.02E+04 E.~83E+05 3.53E+05 ~ La-140 2.42E+02 1.455-04 0.054067 838E+04 9.69E-04 ~8.48E+04 Nb-95 2.73E+04 2.03E+02 0.018496 7.46E+06 3.71 E+04 7 49E+06 Nt> 95m 3.40E+02 2.03E+T1 ' O.010647 8.80E+64 " 7.67E+03 9.56Eidd' IPm-147 .,,6.40E+03_ 5.29E+63 j 0.000722 1.54E+66 5.04E+06 6.59E+d6 lPr-143 2.83E+02 3.94E-041 6.~650587' 9.70E+04. 3.545-03 9 70E+04 ~ jPr-144 2.75E+04 1.44E+04 i 0.002431_6.71E+06 } 1.17E+07 _1.85E+07 Rh-103m 6.95E+03..6.78E+01 0.017471 ! 1,88E+06 i 1.36E+04 1.90E+06 t 'Ru-103 7.09E 03 6.91E+01 0.017475 ' 1.92E+06 1.39E+04 1.94E+06 'Ru-106 1.58EJd4 9.58E+03 0.001888 ~^3.84E+06 8.20E+66 ~ 1.20E+07 Rh-106 1.58E+04 l 9.58E+03 0.001888 3.84E+06 : 8.20E+06 1.20E+07 ISb-126' 2.40E+02 1.99E+02' ~ ~0 Oli6701 i 5.78E+04 l 1.96E+05 2.48E+05i 0.00006f[! 5.88E+05l 2.44Ei 6 ~ 0.0136'75 1.52E+06 i 4.37$404 1.56E+06i ~ lSr-89 ~5.}4E+03 1.53E+02 iSr-90 2.45E+03 2.41E+03 '3.03E+06 i . (To-127 - 2.58E+02' I.79E+01 0.006348 ! 6.46E+04j, 2_71E+04 s.T75f04) ITe-127m 2.59E+02 4.81E+01i 0.006349 l 6.49E+04 2.72E+04 9.21E+04 Y-90 2.45E+03 2.41 E+03 0.000062 i 5.88E+05 2.44E+06 3.03E+06 Y-91 9.57E+03 4.28E+02~D.011726 ~ 2.49E+06 1.41E405 2.64E+06 ~ "~ I Jr-95 1.60E+04 9.54E+02 0.01064 4.14E+06 3.62E+05 4.50E+06 'I RAE 11/11/97 03:10 PM ++ TOTAL PAGE.021 ++

c.. : ;~.. ~.

OCf 7 '97 12:35 F R Ot1 BRP DE C Ot1 PLAN PAGE.001 l FAX TRANSMITTAL l i CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT 10269 US-31 NORTH CHARLEVOIX MI 49720 FAX NUMBER [616] 547-8340 l Internal Access: 1371-340 DATE: lOM/Q7 I l TO: PAUL W HARRIS NAME: COMPANY: US XRC/3RR omCe erioxe xuusER: (301) 415-1169 (301) 415--3313 FAX NUMBER: FROM: Mike Bourassa NAME: OFFICE PHOST NUMBER: (616) 547-8244 do PAGES TO FOLLOW _ %u 2 ame mneue2 % e Geaswu m ro fquy Tuc 5/'eMr FJEt 7 e L M'b W E 3ACd5

I OCT 7 '97 12:35 'FROM BRP DECOM PLAN PAGE.002 m. 1.2 General Description i.2.1 Present Design j l The spent fuel storage pool is located in the reactor building at the refueling level. l The pool currently contains two type "A" racks with 48 storage cells each, one type l "B" rack with 72 storage cells, and one failed fuel storage rack with 25 cells. Thus, the total existing capacity is 193 storage cells. A channel storage rack with 90 cells is } also in the pool. The existing racks are aluminum with a minimum center-to-center spacing of 12 inches. The design is such that the maximum k,gg is approximately 0.80, as stated in the Final Hazards Summary Report (FHSR), and is much less than' the required k,gg of less than 0.95. The spent fuel pool cooling system, which is described in Section 2.2.1, is adequate to remove the decay heat from the proposed expanded storage, plus the heat from a full [ core offload. 1.2.2 Proposed Modification I l-l The proposed modification involves removal of the failed fuel storage rack and the addition of three new racks with a center-to-center spacing of 9 inches between storage locations. The new storage racks are designed to maintain the k,ff at less than 0.95. The new racks consist of storage cans in arrays of 8 x 11, 8 x 13, and 9 x 9. These arrays were chosen to optimir,e use of the pool space. The expanded fuel storage capacity is 441 fuel assemblies. The pool layout with the new racks is l described and illustrated in Section 2.2.2, while a typical rack is described more fully and shown in Section 2.3.1. The racks are free-standing and designed to preclude imparting loads to the pool wa!!s or other racks or equipment in the pool during seismic events. I tt*i J i-3 l.

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7. '97 12:38 FROM BRP DECOM PLAN PAGE.006

2.0 DESCRIPTION

AND SAFETY ANALYSIS 2.1 Introduction Because of uncertainties in government policy on fuel reprocessing, and on the potential unavailability of government spent fuel storage facilities, CPCo plans to increase the storage capacity of the spent fuel pool at the Big Rock Point Plant to allow continued plant operation. The proposed method of accomplishing this increase is, as stated before, to add three high-density spent fuel storage racks to those racks already existing in the spent fuel pool. The original plant design assumed a viable fuel reprocessing industry in the United 4 States by the time the plant commenced operations. Therefore, the original spent fuel N pool was provided with racks to accommodate approximately 2 cores, the assumption being that the one-quarter core discharged each year would be transferred to a reprocessing facility prior to the next year's refueling, and that the pool would always ~ have the capability to accept a f ull-core of fload. However, there is not now and will not for the near term be a capability for reprocessing in the U.S. Therefore, additional spent fuel generated as a result of reactor operation cannot be disposed of and must be stored. l. As of April I 1979, 86 assemblies were being stored in the spent fuel pool, leaving 107 of j the 193 spaces to maintain (at that time) a full-core offload capability until 1981. CPCo deems it necessary to increase the capacity of its spent fuel pool and requests the approval of the NRC to increase the capacity of its spent fuel pool to 441 4 elemen ts. This increase allows the storage of reormally discharged spent fuel until 1990, while retaining the capability to offioad a full core up to that time. This section discusses in detail the various design features incorporated in this modification and demonstrates that they will have no detrimental effect on the health and safety of the

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I l OCT 7 ' 9J J 2 : 38 FROM BRP DECOM PLAN PAGE.007 ~ w 1 2.2 General Description l 2.t1 Present Design l l The spent fuel torage pool is located in the reactor building with a floor elevation of i 602 feet,1-3/16 inctes. The pool currently conti ns two type "A" racts with 96 storage cells, one type "B" rack with 72 cells, and one failed fuel storage rack with 25 l cells. The total existing capacity is 193 cells. There is also a channel storage rack with 90 celis. The present fuei storage racks are aluminum with a minimum center-to-center spacing of 12 inches. At design conditions the maximum K,gg is apodmately 0.80. 1 Provisions for fuel sipping for fuel rod R&D have been included in the fuel pool arrangement. There are two locations allocated for sipping as well as a fuel elevator in which fuel assemblies can be disassembled. Irradiated fuel rods from these assemblies are stored in shipping liners and storage cans, the number of cans being determiwd by the number of rods to be stored, so that in addition to the fuel bundles I there will also be a group of irradiated fuel rods stored in the pool. The fuel pool cooling system is a closed-loop system consisting of two half-capacity l pumps, two half-capacity heat exchangers in parallel, a bypass filter, piping, valves, and instrumentation. The existing heat removal system is adequate to handle the additional decay heat due to pool expansion including a f ull-core offload. 6 The spent fuel pool cooling system has a heat removal capability of 6x10 BTU /hr. The spent fuel pool cooling system is conservatively designed to maintain pool average temperature at less than 95 F with a one-quarter core of fuel with full cycle exposure i in the pool, 48 hours af ter reactor shutdown. The entire fuel pool cooling system is protected against tornado-induced damage and is located in a Seismic Category I structure, although the cooling system itself is nonseismic. Fuel pool makeup water is supplied from the treated radwaste system. A secondary backup supply of water is j available from the fire protection systems. This would be utilized to replenish the fuel i pool water inventory in the event of a loss of pool water up to 200 gpm. The clarity and purity of the:fatercin.the sNnt fusl: pool n're maintained by passing a portion of N the flow through the bypass filter. i 2-2 1

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OCT 7 '97 12:39 FROM BRP DECOM PLAN PAGE.009 ) l l 2.3 Mechanical Design i 2.3.1 Spent Fuel Storage Racks The spent f uel storage racks, shown in Figure 2.3.1, consist of square stainless steel i l cans made of quarter-inch-thick type 304 austenitic stainless steel, and are approxi-mately seven feet long. The active fue! length of 70 inches is entirely enclosed by the stainless steel can. Cans are spaced at 9 inches center-to-center. A half-inch-thick fuel support plate is welded at the bottom of the can to provide support for the fuel. An opening in the plate allows cooling water to flow upward through the fuel assembly to provide for removal of the decay heat from the fuel element. The storage cans are wetoed to a gridded base comprised of a lattice of stainless steel beams 1-1/4 inches wide by 4 inches deep and are structurally tied at the top with an over-under stainless steel grid system. Adjustable legs are provided to accommodate unevenness in the pool floor and to raise the racks off the floor to allow natural circulation flow. All postulated forces will be transmitted through the can and grid system to the leveling legs and into the floor of the fuel pool. The square cans and base plates wi!! support the deadweight of the fuel assemblies. The downward load from the deadweight of the fuel assembly will be transmitted from the baseplate to the grid 2.9.1.4 beams to the gussets and bosses into which the leveling legs are screwed. Details of a typical fuel rack and can, including the supporting grids, fuel seating surface, leveling legs, and connections between various elements are shown on NUS drawing 5148-M-2001.3 Three different rack ce!! arrays maximize use of the available fuel storage space in the pool. For safe handling of the fuel racks, each rack is provided with four lif ting lugs. Lif ting lugs will be welded to the upper grid beam system that is comprised of a lattice of stainless steel beams 1-1/4 inches wide by 2 inches deep that will be welded to the cans. Lif ting forces will be transmitted through the lif ting lugs, to the upper grid beams, to the cans and into the lower grid beam system and base plates.,

2. 9.1..

f O O i..t. t i ,e r J :t. / I Seismic loads will act horizontally on the cans and vertically on the base plates. These loads will be transmitted through the cans and grid systems to the floor of the fuel 2-7 i i

OCT 7 '97 12: 40 FROM BRP DECOM PLAN PAGE.010 pool via the leveling Nys. Impact loads from a dropped fuel assembly will be transmitted either frwn the lead-in guides or a baseplate (depending on impact location) to the remainder of the structure via the grid beam and can system. In the case of a stuck fuel assembly, a vertical force would be applied to a can in the event 2.9.1.4 of an attempted withdrawal. This force would be distributed through the can to the grid beam system and base,ilates to the remainder of the rack. This force would be limited by the 2000 lb load limit cutoff switch of the fuel transfer cask winch. All parts of the new racks are fabricated from type 304 stainless steel, except for the leveling legs which are made from ASTM 276-UNSS 21800 bars. This material is an austenitic stainless steel of the 18-8 type with additives such as manganese and silicon to improve the yield strength to almost twice that of type 304 55. Corrosion tests 2.9.3.2 performed and reported by manufacturers on the UNSS 21-800 material indicate even better corrosion resistance properties than type 304 SS when subjected to similar j tests. (

Reference:

ARMCO Product Bulletin S-MA.) 3 2.3.2 Codes, Standards, and Practices for Fuel Rack Design, Construction, and Assembly The following are the codes, standards, and practices to which the fuel storage racks will be designed, constructed, and assembled. In addition, all provisions of the NRC guidance on spent fuel pool modifications, entitled " Review and Acceptance of Spent ? Fuel Storage and Handling App!! cations" (including errata), are met, as applicable, in o the design of the new spent fuel storage racks. Structural, thermal-hydraulic, and e: criticality analyses have been performed using methodology, criteria, and NRC 2.9.1.2 guidance in effect as of October 1,1979. Although the. analysis methodology may g differ from earlier FHSR analysis methodology, the present design conforms to all FHSR acceptance criteria. 1. Design Codes a. AISC Manual of Steel Construction, 7th Edition,1970, including supplements 1,2, and 3 to the AISC Specification. 2-8

OCT 7 '97 12: 40 FROM BRP DECOM 'LAN PAGE.Oll b. USNRC Regulatory Guides, Division 1 Regulatory Guide Nos. 1.13,

1. 25, 1. 28, 1. 29, 1. 37, 1. 38, 1. 60, 1. 61, 1. 64, 1. 71, 1. 8 5, 1.88, 1. 92, and 1.123. (Revisions effective as of March 1978.)

c. General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos. 1,2,61,62, and 63). d. ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations. Nuclear Regulatory Commission Standard Review Plan (SRP) 3.8.3 e. and 3.8.4. f. NRC Standard Review Plan 9.1.2 (As applicable to spent fuel racks). g. NRC position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14,1978. 2. Material Codes 1 1 American Society for Testing and Materials (ASTM) Standards. a. b. American Society of Mechanical Engineers (ASME) Standards. c. ACI 318-77 Building Code Requirements for Reinforced Concrete. 3. Welding Code ASME Boiler and Pressure Vessel Code, Section IX-1977, Welding and a. Brazing Qualifications. i l' ,8 0 e l 2-9 l

r i OCT 7 '97 12: 41 FROM BPP DECOM PLAN PAGE.012 l i 4. Quality Assurance, Cleanliness, Packaging, Shipping, Receiving, Storage, i and Handling Requirements a. Quality Assurance Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 30, Appendix B. b. "Packa6 ng, Shipping, Receiving, Storage and Handling of items for i Nuclear Power Plants During Construction Phase," ANSI N45.2.2, 1973. " Qualifications of Inspection, Examination, and Testing Personnel for c. the Construction Phase of Nuclear Power Plants,"' ANSI N43.2.6, 1973. d. " Requirements for the Collection, Storage and Maintenance of Qual. ity Assurance Records for Nuclear Power Plants," ANSI N45.2.9, 1974. " Quality Assurance Terms and Definitions," ANSI N45.2.10,1973, e. f. " Quality Assurance Requirements for the Design of Nuclear Power Plants," ANSI N45.2.ll,1974. g. " Requirements for Auditing of Quality Assurance Programs of Nuclear Power Plants," ANSI N45.2.12,1977. J h. " Quality Assurance Requirements for Control of Procurement of Equipment, Materials, and Services," ANSI N45.2.13,1973. 2.10

OCT 7 '97 12: 41 FROM BRP DECOM PLRt1 PAGE.013 ~

* '. LEAD.lN GUIDE t,/ i w40WN ON OTHER FUEL FUEL l

BOXES FOR CLARITY) BOX S 00 g / l GRIOS N s'/j 95 50 BOTTOM FUEL SEATIN' GRID SURFACE 1 l l[ l lkl% /l/p I LEVELING LEG i g 7" l 81.00 IN. %/ IGNd2.3[1. Bid b'Ck POINTTYPICAL FUEL RACK l 2-11

r 1 OCT 7 '97 12:,2 FROM BPP DECOM PLAN PAGE.014 4 l VOLUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 j' CHAPTER 9 - FUEL POOL SYSTEM Page 1 of 1 TABLE OF CONTENTS ) 1.0 FUEL POOL SYSTEM 1.1 PURPOSE

2.0 DESCRIPTION

2.1 FUEL POOL Z.2 SURGE TANK 23 FUEL PIT FILTER 2.4 UNDERWATER FILTERING AND VACUUMING UNIT 2.5 FUEL PIT PUMPS 2.6 FUEL PIT HEAT EXCHANGERS 3.0 PRINCIPLES OF OPERATION

4.0 REFERENCES

23CH9/nrk 08/09/95

-ocT 7..'.9. 7 12_: 42 FRoM BRP DECOM PLAN PRGE.015-VOLUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 1 of 7 1.0 FUEL POOL SYSTEM 1.1 PURPOSE l -The function of the Fuel Pool System is to provide a storage space for spent fuel elements and other highly radioactive materials, and to cool and filter the demineralized water in the pool. i l

2.0 DESCRIPTION

2.1 .The fuel o001 is a concrete structure lined with slainic5s steel and is located in the sphere at.the 632'6" level'5 It is 28' long, 20' wide and 29' deep with a capacity of 1.2 x 10 gallons of water. Fuel pool water normally overflows a weir on the east end r and spills into a surge tank. The area between the stainless steel liner and concrete is segmented into eight sections each with a telltale drain that terminates ut a sink in the Fuel Pit Pump Room (Room 418) to detect any leakage The fuel pool has no bottom penetrations or drain valves. The pool width is spanned by a 480 V ac powered bridge which is ' fed by Panel 2P. This bridge is equipped with a trolly /holst for handling equipment in the pool. Storage racks are located on the floor of the pool to store spent fuel elements, control 51ades, i incore ion chambers and fuel channels. l -Fuel pool level instrumentation is provided by Level Transmitter LT-5803. A Power Supply Unit ES-5803 is fed by Breaker 3Y-8 and provides a signa > to LT-5803 which sends-a signal to Level Indicator LI-5803 on Section A of the cont ol console. I.T-5803 reads from 620'6" to 630'6". The normal poA level is 630'0" A 15-section aluminum cover is provided for the pool to l-prevent dust and foreign elements from entering the pool when it 1s not being used However, the cover is not normally used. The cover is capable of supporting a concentrated load of 200 pounds at any location L 23CH9/nrk 08/09/95 l

ocT 7 ' 9 7,,,12 : 42 FRoM BRP DECoM PLAN PAGE.01E VOLUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM ' Page 2 of 7 2.2 The suroe tank is located east of the pool and accepts the pool overflow. It has a capacity of 4,750 gallons of water; however, I an overflow to the enclosure clean Sump is provided et a level equal to - 2.630 gallons. An outlet pipe leaves the bottom of the' tank and goes to the filter tank or directly to the pump suction via a filter bypass. A sight glass is provided ore 17e south side' of the. surge tank (LG-3625) and covers 48' range which is from the 1 outlet pipe to the overflow hne. Remote and local level instrumentation is provided by Level Element LE-3819 to Level 3 Transmitter LT-3174. which is powered by Panel 2-Y. A local Level i Indicator (LI-3332) 1s located on the east wall of the ' drum enclosure. The meter reads 0 to 100% and surge. tank overflow occurs at 55%. Remote Indication LI-3388 1s located on the Radwaste Panel CO-6 and its signal is provided through a converter (E/P 7804). This gauge also reads out in percent. 2.3 The fuel oit filter is located at elevat1on 600*6" and contains 26 dispcsable socks which are changed monthly or when the pump suction pressure reaches -8" Hg. The filter socks are precoated' with diatomaceous earth to provide better filtering. The water inlet to the filter is provided by the surge tank outlet and the driving force is provided by clevation difference. The flow to the filter tank is regulated by Level Controller LC 3009 which i operates.CV-4128 on the inlet to the tank. The tank is provided with manual 4' inlet, outlet and bypass valves and an overflow to L the enclosure clean sump. A Pressure Gauge (PI-476) is-installed-l on the outlet of the filter tank to' indicate pump suction pressure. PI 426 is an aid to the operators to prevent tripping the fuel p1t pump on low suction pressure during valving operations. The filter tank has a lid and flow distribution baffle which is a single unit. A lifting dev1ce is provided so the lid-baffle unit can be removed to change socks. i i 23CH9/nrk 08/09/95 l

oct 7 '97 12: 43 rRoM SRP DECoM PLAN PAGE.017 VOLUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 3 of 7 2.4 The underwater filter 1no and vacuumino unit (5-43) is located on the south side of the spent fuel pool hanging at a depth of - 14 feet from the pipe hanger rail. This is a portable unit which can be moved in the pool to accommodate fuel pool operations. The filter unit contains two filter cartridges operated in parallel which are changed either once every two months. when the flow decreases to one half the flow with clean filters in the unit, or when the dose rate on the side of one filter housing reaches 20 R/hr whichever comes first. There are { three sizes of filter cartridges available* 10 micron (recommended 1 for initial vacuuming operations). 5 micron and 1 micron (recommended for continuous filtering when no fuel pool operations are in progress).. Flow through the filtering unit is driven by a Submersible Pump (P-97) with a capacity of 250 gpm. Power to the ) pump is supplied by Panel 11L. The starter box is located along i the east railing / wall on the fuel pool deck. A self-powered Flow i Meter (FI-2339) is provided to indicate flew through the discharge of the filter unit. The unit is supplied with a 3 inch discharge hose and two 2 inch intake hoses which can be used to vacuum the bottom of the fuel pool or the surge tank. 2.5 The fuel n1t Dumos are located in Room 418. Elevation 585'6" The pumps are Allis-Chalmers centrifugal type each with a capacity of 250 gpm and powered by General Electric 15 hp. 480 volt motors. Pump 1 is fed by Bus lA and Pump 2 by Bus 2A. In the event that the Fuel Pool System is to be used as a part of the Alterrate Shutdown System. the #2 fuel pool pump motor. M-138, will be fed by the emergency diesel generator v1a a four (4) conductor cable from Panel P29 located within Containment. Refer to Chapter 36. Alternate Shutdown System. for a detailed description. Each pump has a mechanical seal that is cooled and lubricated by the pumped fluid. The pumps are equipped with a 4" gate suction valve. 3" gate discharge valve. 1" drain valve and a suction wye strainer to prevent large foreign material from entering the pumps. The pumps are also protected against low suction pressure by Pressure i Switches PS-671 and PS-672 on Pump 1 and 2. respectively. The switches trip the pump at -2 psig falling pressure. A local Pressure Gauge (PI-401) is located between the pumps to show suction pressure. The discharges of the pumps flow into a common l header which has a Pressure Indicator (PI-340) and lemperature Indicator (T1-1330). The discharge header can route water to the two fuel pit heat exchangers. radwaste, shield tank or the cleanup demin On the discharge line to the heat exchangers there is a flow element (RE-2810) and flow indicator (RE-2328) which reads in percent of 60 gpm Number 2 fuel pit pump differs from Number 1 in that it has a 2" suction and discharge line ins 1de of the two main isolation valves. These 1ines are provided to use Number 2 pump for vacuuming the fuel pool floor while Number 1 pump is still in service to provide cooling. 23CH9/nrk 08/09/95

DCT 7 '97 12: 44 FROM BRP DECoM'PLRN P AGE '. 018 V'0LUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION. MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 4 of 7 2' 6 The fuel oit heat exchancers are located in Room 418. Elevation 589'. and are made by Struthers Wells. 1They are single pass shell side and four-pass tube side with a design pressure of 75.psig at 150 F..The shell side cooling medium is reactor cooling water. Each heat exchanger has a vent; drain and 3 Shell-Side Relief Valve RV-5030 on Number 1 and RV-5031 on z Number 2 which are piped to the enclosure dirty sump. The relief valves are set to relieve at 75 psig. Both shell and tube sides oftheheatexchangersareratgdat 125.000 lb/hr mass flow and are capable of removing 3 x 10 Btu /hr. Each heat exchanger has a 4" gate inlet valve and 4" globe outlet valve on both the tube and shell sides. One 4" gate bypass is provided for the heat exchangers. The discharge of each heat exchanger enters a common header which has a Temperature Indicator-(TI-1308) that reads 0 to 200*F The discharge header then returns the water to the -northwest corner of the fuel pool. The below surface discharge was designed to minimize surface disturbance thus improving l visibility. There is a hole in the discharge-pipe - 1 foot below the surface to serve as a syphon break and prevent syphoning the water out of the pool. l l o 23CH9/nrk 08/09/95 I

ocT 7 "97d 2: 44 FRot1-BRP DE Cofi PLAN PAGE.019 l VOLUME 23-BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 5 of 7 L 3.0 PRINCIPLES OF OPERATION The Fuel. Pool Cooling System is a manually operated system and is normally in service with the fuel pit filter valved in and one pump and heat exchanger in operation. The fuel pool temperature will ordinarily be maintained below 80 F. Due to the large heat sink capacity of the pool, cooling can be stopped for a long period of time (days) with no ill effects. A graph is maintained I in the Tech Data Book (Volume 15). Chart 15.5.G.I. which shows the. number of days before cooling must be initiated. The chart is based on initial temperature being at 100'F and varies with the number of days since the last shutdown. 150'F is the maximum temperature that the pool can reach without degrading the structural strength of the concrete. A makeup water supply has been provided via PI-8 from the 4" post incident recycle line to provide cooling when the sphere is unaccessible such as during a. " Loss of Coolant Accident." Whenever the Post Incident-System'is in the recycle mode.' 28 gpm is continuously supplied to the pool. If cooling is required prior to reaching the recycle mode. MO-7072 .can be opened remotely from the Control Room or by hand locally and the 28 gpm makeup will be initiated. If the fuel pool water conductivity or tadioactivity increases, the fuel pool can be placed on recycle through the radwaste demin per Operating Procedure SOP-11. The fuel pool floor will be kept clean by using the underwater filters and vacuuming unit or by using the No 2 fuel pit pum) to vacuum. The Number 2 fuel pit pump can be lined up to taqe a suction on a vacuuming tank-located in the southwest corner of the pool. Ihis tank has four filter socks to collect any objects larger than 25 mircon. The discharge from Number 2 fuel p1t pump 1s routed to the filter tank during vacuuming procedures. A hose is connected to the inlet of the vacuuming tank and the other end is fastened to a floor vacuuming head which is moved over the floor surface by Operators. PI-410 is located on the pool level to indicate vacuum line pressure or vacuum. The Fuel' Pool vacuum System can also be used to vacuum the inside of the reactor vessel during refueling operations. During this evolution. reactor water makeup is provided by the Control Rod Drive System and the excess water to the Fuel Pool System goes o'11 the surge tank overflow line to the clean sumps. Specific fuel pool operations are outlined in Operating Procedure $0P-9. The fuel pool liner leak detection is check:d at the start of each shift and logged on the Control Room yellrA log sheet. L 1 23CH9/nrk 08/09/95

oCT 7 '97 12: 45 FRoM BRP DECoM PLAN PAGE.020 VOLUME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 6 of 7 Normal water makeup to the Fuel Pool System is from the treated waste header from radwaste: however, demin water is also available. Spent fuel assemblies shall be stored in designated storage racks located in the fuel pool. Six racks are provided, one with a 8 x 13 array (104) on a square patch. one with a 6 x 12 array (72) on a square pitch. two with a 6 x 8 array (96) on a square pitch, one with a 9 x 9 array (81) on a square pitch and one with an 8 x 11 array on a square pitch for a total of 441 bundles. There also exists a 9 x 10 array (90) storage rack for fuel channels. Movement of fuel into or out of.the storage racks shall be restricted to one bundle at a time.

4.0 REFERENCES

a. Prints 0740G20137 Reactor Building R & D Attachment and Spent Fuel Storage Rack 0740G20138 Reactor Building Fuel Pool Plan and Miscellaneous Detail 0740F20329 Rack l.eveling Tool Spent Fuel Poci Storage Rack 0740G20330 Pool Arrangement 0740G30104 Schematic Diagram In: stentation 0740G30105 Schematic Diagram Station Power & 480 Volt Transformers 0740G30113 Schematic Diagram Service & Cooling Water Systems 0740G40102 Equipment Location Section A-A 0740G40103 Equipment Location Sections B-B & C C 0740G40111 Piping 8 Instrument Diagram Circulating. Cooling & Service Water Systems 0740G40152 Spent Fuel Pit Cooling Water Reactor Building Plan l 0740G40153 Spent Fuel Pit Cooling Water Reactor i Building Section 0740G44017 Spent Fuel Pool System Valve Line-Up Diagram 23CH9/nrk 08/09/95

OC7 7,,' 9 7 12:45 FROM BRP DECOM PLAN PAGE.021 l V01.UME 23 BIG ROCK POINT SYSTEMS DESCRIPTION MANUAL Revision 11 CHAPTER 9 - FUEL POOL SYSTEM Page 7 of 7 i h. Volume 2. Technical Specifications (4.2.11.b) (License) (3.0) (7.3.2) (7 4) c. Volume 15. Tech Data. (15.5.G.1) d. Volume 4. Instrument Data e. FHSR (5.12.4) (5.12.5) (1.2.2.4) (2.1.1.3) (2.2.2.6) (3.2) (3.5) (6.2) (6.3.1.2.2) (6.3) (Chapter 9) (10.4.5.3) (11.4) (11.5) (12 1.2) (12.1) (12.2.1) (12.3) (13.5) (15.7) f. Volume 3. Operating Procedures. (50P-9) (SOP-ll) (ONP 2.38) (ONP 2.100) (ONP 2.2) (ONP 2.105) l 1 I i l l i 23CH9/nrk 08/09/95 l

- TOTAL PAGE.021 ++

616 54'7 8340 OCT 7 ' '3 7 G:28 F R Of1 BPP DECOM PLAN PAGE.001 l FAX TRANSMITTAL CONSUMER $ ENERGY COMPANY BIG ROCK POINT PLANT i 10269 US-31 NORTH CHARLEVOIX MI 49720 FAX NUMBER [616] 547-8340 Internal Access: 1371-340 DATE: I o 7; cr7 TO: PAUL W HARRIS urus: cOuvixv: US XRC/NRR (301) 415-1169 OFFICE PliONE NUMBER: (301) 415-3313 FAX NUMBER: 4 FROM: NAME: Mike Bourassa (616) 547-8244 OFFICE PIiONE NUMBER: PAGES TO FOLLOW I$ n a w ~.c w - se,c~.~s occ, w n8nuYsis ig

OCT 7 '97 8:28 FROM BRP DECOM F _ Ata PAGE.002 l 4Yn rt'N ejr t f RECORD INDEXING FORM ~ = v Cartndge/ Frame No_ / ~ Be Completed by ORIGINATOR / Completed by N N b1 i Location 3/TN Dale Completed 3 '30-Y I SUBJECT INDEX NUMBERS Area and a Systems and/or Administrative Number MUST Be Coded ~ Area Systems Administrative Equipment Equipment Administratwe Area Unit System C;assification NumW Number (AAFA) (UNIT) tSYSCODE) ! (EOUIPCL) (EQUIPNO) ,jADMiN) yqo p '/c, o s - ) l RETENTION (RE TE N) _. [b (Life of ;ne Plant (L) Lite of the Company (LC) or fiumcu of Years To Be Retained) is This Aecord Safeguards,10CFR 2 790(d) or Confidential Information? Yes No Following Descriptors To Be Indexed Where Applicable animammesu-t i Protec1 N?me f.O!!1rn $$jo r1 o'itC (PROJNAM) { Architect / Engineer Number Personal File Number (BECHNO) (PERFIL) EAf'8R7'*DPQdS# Document Name l?t4 ln e2/ o A 4 k ne / sis..__ Sequence Number I (DOCNAM) / / (SEONO) Rct;ronce L C0/71 YAn 0 A sshf r.$ (REF) / ~ Department (DEPD Analin's Subiect SMl' .. _ 2 tr1 M *! SeLo nrMg.$3l0Miofc $1* l /1 (F RETEXT) / [ / %,MP i ~ To Be Completed by Documen Dste(s) f Y /2s_ I (DATE) Addresse s).. Organaation n (TONAM) (TOORG) 1 i ~ i Originator (s), _. Organization. (FROMNAM) (FROMORG) ~ a g

oCT 7 '97 8:29 FROM BRP DECoM PLAN PAGE.003 I peer EA-BRP-DP-CH5-1 l BIG ROCK POINT NUCLEAR PLANT REVISION O l ENGINEERING ANALYSIS SHEET Sheet 1 of 11 1 TITLE: Fuel Damage Decommissioning Accident Analysis for BRP { j j IN]TIATION AM) R[ VIEW fnitiated Deview Mett od Check / Terhmically Reviewed Rev Att Det Qual By Date By Date j / Cotc Revaew Test 8 Descrippon [d O /[ /HfW / / + o orioWt tuve / G j OBJECTJXE: This engineering analysis is performed in order to establish a safety basis for decommissioning of Big Rock Point which is consistent with the current updated FHSR. While the FHSR applies specifically to the operating phase of 819 Rock Point, the Decommissioning Plen is expected to serve as the plant's safety basis document for the decommissioning phase. Radiation doses to the public (beyond the site boundary), and to onsite personnel have been calculated to provide a basis for protective action determinations. Dose consequences of several potential fuel-related events l hue been evaluated i ANALYSIS INPUTS: Effort has been made u maintain consistency with established FHSR anelyses of Section 15 7.1 (Fuel Driaging Accidents) which include a single bundle fuei ? handling event, a loss of fuel transfer cask cooling event. and a fuel trarofer cask drop event in addition., this analysis includes effects of gap activity loss from all bundles of the latest core, and from all 500 bundles (maximum expected fur.1 pooi capacity post shutdown) stored in the fuel pool Dose calculational techniques have changed with regard to internal dose since the original FHSR analyses were performed Since January 1. 1993. internal and external doses bwe been summed in accordance with 10 CFR Par t 20.1202 " Compliance with requirements for summation of external and internal doses" A similar summation requirement fee emergency planning purposes on was implemented on January 1. 1994 fc_r EPA-400 guidance [ Reference 11 The new emergency planning criteria also address o.in dose. Thus, skin dose is 7-calculated and internal and external doses are summed in this analysis. both of which which differ from the FHSR analyses. In addition. revised dose conversion factors for lodines are used in this analysis as presented in the EPA guidance n 330 n,u,n o, n. fb l t, a i i b ' ~bS 5 Il @) El l

I oct 7 ',9_7 _ 8 : 3 0 FRoM BRP DECoM PLAN PAGE.004 ,,g, EA-BRP-DP-CH5-1 l .cewe inun comunes k 8M BIG ROCK POINT NUCLEAR PLANT REVISION 0 g c -- ,% M ENGINEERING ANALYSIS 51EET Sheet 2 of 11 TITLE: Fuel Damage Decommissioning Accident Analysis for BRP l ASSUMPTIONS:

1) Fission product inventories are for a 240 Mwt core, based on values from NE00 24782 [ Reference 21

2) Meteorology is in accordance with Regulatory Guide 1.25.

Dispersion 3 factors at the site boundary for short duration releases are 1.8E-4 sec/m (Reg Guide 1.25. Figure 4 at 805m site boundary distance) for an elevated release from the 75 meter stack. and 6.48E-4 for a ground level release (Reg 2 Guide 125 Figure 1 at 805m divided by the value for 500m reactor building cross-sectional area /2 from Reg Guide 1.25 Figure 2).

3) For the sake of conservatism. no containment isolation is assumed for icalculation of offsite dose.

4) No fuel may be present in the reactor vessel under the possession only license (PDL) for which this calculation applies.

The PDL will be requested to take effect 7 days post-shutdown to allow sufficient time to fully offload i the reactor. Thus minimum decay time for spent fuel in this analysis is 168 hours (7 days). 5) It is assumed that fuel will not be loaded into a dry transportable canister prior to one year of decay. This is a conservative interval to assume for decay prior to manipulating fuel in a dry configuration. Thus, an accident involving dry fuel (no water scrubbing of iodines) is not analyzed for decay times shorter than 1 year l

6) Rad 1onuclide release is 10% of the noble gas and,odine inventory of the l

fuel except for Kr-85 and 1 129 of which 30% 15 released. In the dry case it is assumed that all this activity escapes to the environment. In the wet case. 99% of the lodines released from fuel are retamed in the water per Regulatory Guide 1.25 which assumes a minimum water depth above the fuel of 23 feet. Big Rock's fuel pool weir overflow structure. 28' 5" above the floor of the pool (Drawing M-153, sheet 1) maintains 23 feet of water above the fuel l active region (Drawings 0740G20137 for racks and ANF-304.643 fcr fuel assemblies). 8'~l ! I t, 4 4 I e '! 'l ii

n oCT 7,' 9 7 8:30 FRoM BPP DECoM PLAr4 PAGE.005 l EA-BRP-DP-CII5-1 g mru ( w casames i k N BIG ROCK POINT NUCLEAR PLANT REVISION 0 1 E ENGTNF.ERING ANALYSIS SHEET Sheet 3 of 11 i TITLE: Fuel Damage Decommissioning Accident Analysis for BRP FUEL TRANSFER CASK DROP _ CALCULATIONS 1 5_i_te floundary Doses Dose in rem TEDE = (X/0)(Rel Rate)(DCF,) Equation 1 Dose in rem CDE - (X/0)(Re1 Rate)(DCF )(T) Equation 2 t Dose in rem to Skin - (x/0)(Re1 Rate)(DCF,)(T) Equation 3 i Where-DCF = combined (inhalation. ground shine and submersion) dose e 3 conversion factor. (rem /br)/(C1/m ), per EPA-400. Table b-1 i DCF, - skin dose conversion factor (units as above. after converting 3 l froin Sv per Bq s m by a factor of 1.333E+19), per EPA-402 l [ Reference 3]. Table Ill.1 i DCF - thyroid dose conversion factor. (units as above). per EPA-400. TabIe5-2 3 3 l X/0, - 1.8E-4 s/m /3600s/hr - 5.0E-8 hr/m for elevated (stack) releases, fumigation condition. Pasquill F stability, wind speed 1 m/sec 3 X/0, - 6.48E-4 s/m'/3600s/hr - 1.8E 7 hr/m for ground level release Re1 Rate - (gp)(.22)(240 %)(Ci/Mwt),(rf),F/.T. C1/Mwt per NEDO-24782 gp = gap fraction per R G 1.25:.10 except 1-129 & Kr-85 .30 F - factor of 2 0 for iodines only. to correct for the 50% lodines assumed released in NEDO-24782. rf - release fraction, iodines through water .01, all else = 1.0 l T = hours of exposure = hrs of release (I's cancel) { subscript i - notation for dose type (thyroid. skin, and TEDE) 4 subscript n - notation that the parameter is nuclide-specific l 1 t i e i i e il .1 1 '. O J

ocT 7 Jy_,e : 31 FRoM BRP DECoM PLAN PAGE.00G ( ,,g, EA-BRP-DP-CHS.1 l meum W PLMT Cpggggg i 8 8N DIG ROCK POINT NUCLEAR PLANT REVISION O h** ENGINEERING ANALYSIS SHEET Sheet 4 of I] HTLE: Fuel Damage Decommissioning Accident Analysis for BRP I l Applying the appropriate constants to Equations 1. 2 and 3 gives: l Dose (elevated release) - (5.0E-8)(.22)(240)(Ci/Mwt),(DCF),,(rf),F 3 i units: (hr/m )(C1/Nt)(rem *m'/C1*hr) = rem - (2.64E-6)(C1/Nt),(DCF}n,(rf),F l Dose (ground level) ( 1. 7 E - 7 ) (. 22))( 240) (C1/ht ),(DCF )h,r()rf),F 3 units: (hr/m (C1/Nt)(rem *m /C1* = rem - (8.98E-6)(C1/ht ),(DCF),( r f ),F Ground level dose at the site boundary is a constant of 3.4 times the elevated dose. The ratio is that of the X/Q's (1.7E-7/5 OE-8) - 3.40. The 1.otus spreadsheet CSKDRP3.WK4 (Attachment I) was utilized to provide site boundary doses as a functicn of time after plant shutdown, based on Equations

1. 2 and 3 above.

For nuclide quantitles (C1/Nt) at decay times not provided by the NEDO-24782 document. the values at closest earlier time tvere decayed by l their halflives [ Reference 4]. All I-132 decay was performed using its parent's halflife (Te-132. halflife 3.26 days). Since the parent halflife is l much longer than the daughter. this is conservative. 1 l Doses for the ground level release case are summarized on page 2 of the spreadsheet CSKDRP3.WK4 output. Dry case values at 365 days utilize wet case values for noble gas plus 100 times the lodine values to account for the lack of water scrubbing l Elevated case doses (a constant factor of 3.4 lower than the around level doses as discussed previously). are prcsented on page 3 of Attachment 1 l II. O_nsite Dost Onsite doses were calculated using the same source term values as used for l dose to offsite populations. Onsite concentrations were taken as containment l concentration (assuming all activity is released into the containment free volume of 2.66E+10 cc) l l The computer code MicroShield Version 4 (Grove Engineering. Inc) has been utilized to provide dose rates at i.he edge of a " wall-less" containment and through the wall (1.91 cm steel) at distances of 45 meters and 90 meters from the reactor building center Doses were obtained for fission product I inventories with the same decay times as used in offsite dose calculations. Micro $hield data is provided in Attachment II for the cask drop case of 18.5 bundics One bundle values are obtained by using the ratio 1 5/18.5 (the value of 1.51s to account for peaking factor of 1.5). Full core 7s 1/.22. t l ("I fl t 17 e's if ' 4 fI

oCT 7 '97 -8131 FROM BRP DECOM PLAN PAGE.007 ,,i, EA-BRP-DP-CHS-1 l nun cemens: mesa (- BIG ROCK POINT NUCLEAR PLANT REVISION 0 ) gp E ENGINEERING ANALYSIS SHEET Sheet 5 of 11 TITLE: Fuct Damage Decommissioning Accident Analysis for B'RP and 500 bundles 15 (1/.22)+((500-84)/84))/.22 times the microshield outputs for the 18.5 bundle case The dry case is the same as the wet case for one full core because Kr-85 is the only significant contributor. and is no different with or without water scrubbing. The " wall-less" (no steel shielding) results were multiplied by a factor of two to provide sphere center dose rates [ Reference 5]. The factor of two is accurate to better than 10% provided that the product of the linear energy i absorption coef f1cient and the containment radius is less than or equal to 0.2 (in this case it equals ~0.06 for gamma energies of 0.08 MeV and above) (Reference 6]. The 45 meter distance represents approximate distance to the entry to the Service Building and the roadway around containment within the protected area. The distance of 90 meters represents the Security Building and the Annex Building, and is conservative for the parking area where time would be spent (especially in winter) getting cars under way in an evacuation scenario. Table 1 provides a summary of dose rates inside containment and at 45 and 90 meters as a function of time after plant shutdown. For shine dose, the dry and wet cases at 365 days are equal because iodines provide insignificant dose contributions even when multiplied by 100 (see MicroShield runs in Attachment II). The program 0FFSITE [ Reference 7]. was utilized to determine plume shine dose for onsite personnel prior to evacuation from the site for the accident case in which all evolved activity is released in a 2-hr period. OFFSITE outputs are provided in Attachment Ill. Stability G. one mile per hour wind speed values were used at 0.05 miles for input to table 5.2-1 of Decommissioning plan Section 5, This table is provided in Attachment IV. Direct containment shine is added for intervals an individual would not be shielded by the 4'6" control room shield wall. Dose rate behind the shield wall )s a factor of approximately 1.000 less than the unshielded dose rate. Thus, dose during intervals behind the well (control room, technical support center and operations support center). 15 considered negligible in comparison wtth the intervals an individual would not be protected by the wall. Shine dose at 365 days is no different for the dry drop than the wet drop case. Essentially all the dose is due to Kr-85 which is the same for wet and dry cases. Release rate is the same in these calcualtions as in the calculation of offsite population dose in Section I. Exposure times to shinc from the , containment and the elevated plume during evacuation are assumed as follow: 1 minute at an equivaient of 45 meters from s)here e traveling from an unshielded work area to a slielded assembly area oas.. O t o i

1 oCT 7 '97 8:32 FRoM BRP DECoM PLAN PAGE.008 ,,l, EA-BRP-DP-CH5-1 i ;wn i rum tumme ( BIG ROCK POINT NUCLEAR PLANT REVISION O .,_,,M ENGINEERTNG ANALYSIS SilEET Sheet 6 of 11 ) i Tmf; Fuel Damage Decommissioning Accident Analysis for BRP 9 minutes at guardhouse or equivalent unshielded e distance (90 meters) preparing for and performing evacuation 10 minutes total exposure to plume shine (continuing through the above activities) 5 minutes of submersion in the plume at the site e boundary dose rate (as calculated for offsite dose) is assumed during evacuation. Total dose to non-essential onsite personnel evacuating the site is summarized in Table 5.2-1 of Attachment IV. The dose includes containment shine, plume shine and immersion in the plume at the site boundary from an elevated plume. lhe elevated plume is utilized for evacuees because the most realistic expectation for rapid enough release of radioactivity from containment to affect these individuals would be via ventilation exhaust through the stack. The site boundary submersion is conservative because under the worst (highest dose) stability conditions (Pasquill F and G). the plume does not intersect the ground until beyond the site boundary. External dose rates at containment center. and inhalation committed dose rates from breathing containment c1r concentrations of radioactivity have been calculated for operations personnel who might be required to enter containment or otherwise be exposed to concentrations approaching those of containment from leakage pathways Containment concentrations are conservatively utilized for these calculations given that true pressure boundary 1 solation of containment is not planned for decommissioning and that manual means of release path 1 solation may require exposure to these concentrations. Concentrations are derived from the total activity released from the accident as described for offsite release, divided by containment free volume of 2.66E+10 cc. The spreadsheet SITEDOSE.WK4 (Attachment V) was used to calculate dose rates using maximum internal and external dose contributions. lhe same dose conversion factors describe previously for offsite dose were used here. However, the MicroShield external (unshielded. containment center) dose is used rather than external dose from a plume. For times not run on MtcroShield. the ratio of Microshield to plume dose from the previous time interval was used Use of the previous time interval ratio gives a conservatively high adjustment ratio. since adjustment ratios decrease with time (see "ADJ RATIO" entries on page 2 of SITEDOSE output)- I l Doses with and without use of a self contained breathing apparatus (SCBA) in f pressure demand mode (assigned a protection factor of 10.000 in 10 CFR Part l

20. Appendix A) are provided on pages 2 and 3 of the SITEDOSE.WK4 output.

The SCBA's currently are maintained for uSe by s1te emergency personnel. O s i o i. e ,1 1'.- O ;. a />

oCT 7,'97 8:33 F R ot1 BRP DECofi PLAT 4 PAGE.009 EA-BRP-DP-CH5-1 i f n,un l urma casamos r meer BIG ROCK POINT NUCLEAR PLANT REVISION 0 mE ENGINEERING ANALYSIS SliEET Sheet 7 of 11 l l TITLE: Fuel Damage Decommissioning Accident Analysis for BRP LOSS OF FUEL TRANSFER CASK COOLING i This event is described briefly in the Updated FHSR. Section 15.7.1.3.3. No doses are calculated. but it is stated that the incident would result in release of 20% of the bundle's inventory of radioiodines and noble gas, or l 61.000 Ci noble gas and 37.000 C1 halogens over a period not exceedir; 2 l hours. Movement of freshly irradiated fuel does not apply to the Possession Only License since all fuel must be loaded into the pool prior to the POL becoming effcctive. Consequently. this accident would be possible only upon movement of fuel from the pool to dry transportable canisters in preparation for shipment or storage of dry fuel after reasonable decay. An analysis will be required prior to this evolution to determine fuel clad temperatures and resultant radionuclide release potential during transfer incidents specific to the transfer process chosen No analysis is provided here because the l transfer process is not defined at this time. i SJNGLE BUNDLE DROP The single bundle drop is described in FHSR Section 15.7.1.2.2. The accident also 1s included in this analysis as the lower end of an accident s)ectrum l which ranges from this single bundle event to the upper end event w1ich would involve damage to the maximum of 500 bundle pool capacity expected to reside in the pool during decommissioning (an additional rack would be required for 500 bundle capacity). A peaking factor of 1.5 is applied to the single bundle event. Dose calculations cre performed as for the fuel transfer task drop. with source term equal to 1 5/18 5 - 0.0811 times the task drop source term. This factor accounts for the peaking factor of 1.5 for a worst single bundle. and for the 18.5 bundles (22% of the core) damaged in the cask drop event. 1 FMLL CORE AND FULL POOL EVLNTS l In order to allow for ultimate use of larger casks than the fuel transfer cask and to account for decommissioning activities which could generate questions as to potential for fuel damage greater than that analyzed for the fpel i j transfer cask. analysis has been extended to include damage to a full core offload and to the full 500 bundles in the pool. These analyses have been performed by use of CSKDRP3.WK4. SITEDOSE.WK4. and 1 MicroShield with the same assumptions as described for the spent fuel cask analysis. Source term for the full core case is 1/.22 times the cask drop. s' I !! I f f f 8 j ti vt t

oCT, 7 '97 8:33 FPOM BPP DECoM PLAN PAGE.010 1 l,,, EA-BRP-DP-CHS-1 Nur.Ws PLMI Cgggggig mer BIG ROCK POINT NUCLEAR PLANT REVISION 0 j ENGINEERING ANALYSIS SHEET Sheet 8 of 11 TTTLE: Fuel Damage Decommissioning Accident Analysis for BRP since the cask drop damaged 22% of t'e core. Source term for the full pool n equals the full core plus the source term of 500-84 (=416) bundles decayed for 365 days (this conservatively assumes that all bundles other than the final core offload have decayed only one year prior to the accident). CONCLUSIONS The results of this analysis indicate that without containment isolation, an accident involving a single bundle could occur 21 days or more atMr plant shutdown without exceeding EPA PAG's. An acc1 dent of suft1cient size to release gap activity of all fuel bundles in the spent fuel pool could occur 68 days or more after plant shutdown without exceeding PAC's. Figures 1 and 2 present plots of number of bundles damaged to reach PAG's as a function of time. Figure 1 includes the full range of damaged bundles (1 to 600), while Figure 2 is the same curve steep portton of the curve. plotted to 120 bundles for added clarity in the Without use of respiratory protective devices, doses to essential personnel in the performance of mitigating actions could be large, particularly with regard to thyroid dose (and its contribution to TEDE) prior to 150 days post-shutdown. Consequently. a commitment will be made in the Decommissioning Plan to retain the current Emergency Plan and Emergency Implementing Procedures (which call for use of SCBA's) for 150 days post-shutdown to provide that conservatively low doses will be maintained in a worst case event. The high calculated skin dose from Kr-85 would be mitigated by protective clothing thickness in the range of 60 to 110 mg/cm (mean range to maximum 2 range for Kr-85 betas) [ References 6 and 8]. This range of absorber thickness is met by a cloth garment covered by a rubber or plastic rainsuit (which is standard wear for extreme contamination conditions). and with respect to the face, by the facemask of a respiratory 2rotective device. Commitment will be made in the Decommissioning Plan that tie Defueled Emergency Plan address the the availability and use of such clothing in the event of a severe fuel handling event. ~ <, u

oCT 7 '97 8:34 FROM BRP DECoM PLRN PRGE.011 l i '3 l C m EA-BRP-DP-CHS y peuw BIG ROCK POINT NUCLEAR PLANT REVISION 0 _,= ENGINEERING ANALYSIS SHEET Sheet 9 of 11 l TITLE: Fuel Damage Decommissioning Accident Analysis for BRP REFERENCES 1. EPA-400-R-93-081. " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents". May, 1992. 2. NED0-24782 "BWR Owner's Group NUREG-0578 Implementation: Analysis and Positions for Plant-Unique Submittals". August. 1980. 3. FPA-402-R-93-081. " External Exposure to Radionuclides in Air Water and Soil". Federal Guidance Report No. 12. Table III 4 General Electric Company. "Nuclides and Isotopes. Chart of the Nuclides" Edition 14. 1989. 5. G.J. Hine and G L. Brownell. " Radiation Dosimetry". 1956 pp 853-855. 6. Radiological Health Handbook, 1970 edition, page 135 (linear energy absorption factors) and page 156 (beta particle range energy curve). 7. Big Rock Point Emergency Implenienting Procedure EPIP SA-

6. " Automated Dose Assessment Program". Revision 149 8.

L.T. Dillman and F.C. Von der Lage. "Radionuclide Decay Schemes and Nuclear Parameters for use in Radiation-Dose Estimation". MIRD Phamphlet No. 10. Society of Nuclear Medicine. 1975 i '

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OCT 7 '97 ~'P:36 F rot 1 BRP DECOM PLAN PAGE.014 6/7 -8RP-PP-CHS - 1 4?w. o Anselmed.E \\ twp1W3.Wk4 odyd g g g g, l

OCT 7 '97 Oj,36 FROM BRP DECOM PLAN 'PAGE 015 ,i g

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l DECOM CA0 OFFSITE DOSE FOR CASK DROP (SPREADSHEET CSKORP3 WK4) - GROUND L.EVEL RELEASE i I I .NUCLID RF & DCF Ci/ Mw and Dose m Rem (All Activity released) g Days After Shutdown j 05 50 70 10 0 30.0' 60 0 100 0 150 0 365 0 HFLFE i 1-129 6OE03 5.1E 04 51 E-04 51 E-04 5.1 E-04 51E-04 51E-04 5.1E-04 51E 04 51E-04 5 8E+09 I THY 6 9E+06 2 OE-04 2.0E-04 2 CE-04 2 0E-04 2 OE-04 2 OE-04 2.0E-04 2 OE-04 2.0E 04 ' EXT 4BE+00 14E-10 14E-10 14E-10 14E 10 14E-10 1.4E-10 1.4E-10 14E-10 1.4E-10 TOT 2.1E + 05 61 E-06 61E 06 61E 06 61E-06 61E-06 61E-06 6.1E-06 61E-06 61E-06 SKIN 1.5E + 01 4 3E 10 4.3E 10 4 3E 10 4 3E-10 4.3E-10 4 3E-10 4 3E 10 4 3E-10 4.3E-10 uCi/cc 61E-09 61E-09 61E-09 61 E-09 6.1 E-09 61E-09 61E-09 6.1E-09 61E-09 l131 2 OE-03 14E+04 9 6E+03 01E+03 6.2E+03

1. iE+03 8.4E*01 2.7E400 ' 3 6E-02 3 4E-10 8 OE+00 THY 1.3E+ 06 3 SE+02 2 4E+02 2 OE+02 1 SE+02 2.7E+01 21E*00 6 7E 02 9 OE-04 8 4E-12 EXT 2 2E+02 5 9E-02 4 0E-02 3 4E-02 2.6E 02 4.6E-03 3 5E-04 1.1E-05.1.5E-07 1.4E 15 TOT' S3E+04 14E+ 01 9 7E+00 81E+00 6 2E+00 1 1E+00 8.5E-02 2.7E-03 3 7E-05 3 4E 13 SKIN -

4 OE+02 ' 1.1E-01 7.2E-02 6.1E-02 4 7E 02 8 3E-03 6.3E 04 2.0E-05 2.7E-07 2.6E-15 uCs/cc - 5.6E-02 3 8E-02 3 2E 02 2 SE-02 4 4E-03 3 3E-04 1.1E-05 14E-07 1.JE 15 l-132-2 GE-03 92E+02 1 SE+02 9 BE+01 51E+01 7.1E-01 ' 1.2E 03 0 OE+00 0 OE+00 0.0E+00 3 3E+00 THY 7.7E+03 13E-01 2 2E-02 14E 02 7 SE-03 1.0E-04 17E-07 0 CE+00 0.0E+00 0 OE+00 EXT 14E403 2 4E-02 4 OE-03 2 SE-03 1.4E 03 1 9E-05 3 2E-08 0 OE+00 0 OE+00 0.0E+00 TOT 49E+03 8 SE-02 14E-02 91E-03 4.7E-03 6 6E-05 1.1E-07 0.0E+00 0 OE+00 0 DE+00 SKIN 21E+03 3.7E 02 6 OE-03 3 GE-03 2 OE-03 2.8E-05 4.8E-08 0 OE+00 0.0E+00 0.0E+00 uCi/cc 3 6E-03 6 OE-04 3 9E 04 2.0E-04 2 8E-06 4.7E-09 0 OE+00 0.0E+00 0 OE+00 1133 2 OE-03 1,6E

04 4.6E+02 9 3E+01 8 7E+00 1.2E-06 0.0E+00 0.0E+00 0.0E+00 0.0E+00 8 7E-01 TNY 2 2E+05 6 8E+01 5 3E-01 1 1 E-01 1.0E-02 13E-09 0.0E+00 0 OE+00 0 OE+00 0 OE+00 EXT 3 SE+02 3 OE-02 8 SE-04 1 7E-04 16E 05 2.1 E-12 0.0E+00 0 DE+00 0 OE+00 0.0E+00 TOT 1SE 04 13E*00 3.6E-02 7 4E-03 6 9E-04 91E-11 0 0E+00 0 OE+00 0 OE+00 0.0E+00 SKIN 7 8E+02 6 6E-02 19E-03 3 8E-04 3 6E-05 4.7E-12 0 OE+00 0.0E+00 0 OE+00 uCdcc 6 4E-02 18E-03 3 7E-04 3 5E-05 4 6E-12 0.0E+00 0 OE+00 0 OE+00 0.0E+00 1135 2 OE-03 7,7E+ 03 21E 01 13E-03 1.8E-06 0DE+00 0.0E+00 0 OE+00 0 OE*00 0.0E+00 2.7E-01 THY 3 8E+04 5.5E+00 1 SE-04 9 6E-07 13E-09 0DE+00 0.0E+00 0 OE+00 0.0E+00 0 OE+00 EXT

- 9 SE+02 14E-01 3 BE-06 2 4E-08 3 3E 11 0 DE+00 0.0E+00 0 OE+00 0.0E+00 0 OE+00 TOT 81E+03 12E+00 3 2E-05 21E-07 2 BE-10 0 OE+00 0 OE+00 0 OE+00 0 OE+00 0.0E+00 SKIN 1 SE+03 2.2E-01 5 9E-06 3 8E-08 5.1E-11 0.0E+ 00 0 OE+00 0 OE+00 0 OE+00 0.0E+00 uCt/cc 3 OE-02 8 3E 07 5 3E-09 7 3E 12 0OE+00 0.0E400 0 OE+00 0 OE+00 0.0E+00 ' Xe 131m.1 OE-01 17E+02 1 SE+02 13E+02 12E+02 4 9E+01 9 9E+001 OE+00. 5 4E-02 2 2E-07 12E+01 THY 0OE+00 0OE+00 0 OE+00 0 OE+00 0 OE+00 0.0E*00 0 DE+00 0 OE+00 0.0E+00 0 OE+00 EXT 4 9E+00 81E-0A 7 OE-04 6 2E-04 5 6E-04 2.3E-04 4.6E-05 4.7E-06 2.5E-07 1 OE 12 TOT 4 9E+00 81E-04 7 OE-04 6 2E 04 5 6E-04 2 3E-04 4 6E 05 4 7E 06 2 SE-07.1.0E-12 SKIN 64E+01. 1 1 E-02 9 2E-03 61E-03 7.3E 03 3 DE-03 6 OE-04 61E-05 3.3E 06 1.3E-11 vCi/cc 3 5E-02 3 OE-02 2 7E-02 2 4E-02 9.7E 03 2.0E-03 2 DE-04 1.1E-05 4 4E-11 X3133m 1 OE-01 18E+03 5 OE+02 2 7E+02 1 1E+02 2 7E-01 0 OE+00 0 OE+00 0 OE+00 0.0E+00 2 2E+00 THY: 0OE+00 0DE+00 0 CE+00 0 OE+00 0 OE+00 0 OE*00 0 OE+00 0.0E+00 0 OE+00 0 OE+00 ~ EXT

1. 7 E + 01 2.8E-02 81E-03 4 3E-03 18E-03 4 4E-06 0 OE+00 0 OE+00 0 OE+00 0.0E+00 TOT 17E+01 2 8E-02 81E 03 4 3E 03 1 BE-03 4 4E-06 0 DE400 0 OE+00 0 DE+00 0 OE+00 SKIN 14E+02 2 3E-01 6 6E 02 3 5E 02 1.4E-02 3 6E 05 0.0E+00 0.0E+00 0 OE+00 0 OE+00 uCi/cc -

3 SE-01 9 9E 02 5 3E-02 2 2E-02 5 4E-05 0 OE+00 0.0E+00 0.0E+00 0 OE+00 Xe-133 1 OE 01 '4E+04 3 2E+04 2 SE404 17E+04 12E+03 24E+01 13E-01 1.7E-04 0 OE*00 5 2E+00 THY 0OE+00 0CE+00 0 OE+00 0 OE+00 0OE+00 0 OE+00 0.0E+00 0 OE+00 0 OE+00 0 DE+00 EXT 2 0F+01 1OE+00 61E-01 4 7E-01 3 2E-01 2 3E-02 4 6E-04 2.5E-06 3 3E-09 0.0E+00 TOT 2.0E + 01 1 OE+00 61 E-01 4 7E-01 3 2E-01 2 3E 02 4 6E-04 2 5E-06 3 3E-09 0 OE+00 l SKIN 6 6F+01 3.4E+00 2 OE+00 1 SE+00 1 1E400 7 6E-02 1 SE-03 8 2E 06 1.1E-08 0 OE+00 uCi/cc 11Et91 ;,6 4F, 00 4 9E+00,.3 4E.00g 7 7 24E-01 4.8E-03 2 6E 05 3 SE-08 0 OE+00

OCT 7 '97 - 8:37 FROM BRP DECOM PLAN PAGE 016 g w z, O 11 l DECOM CALC OFFSITE DOSE FOR CASK DROP (SPREADSHEET CSKDRP3 WK4) - GROUND LEVEL RELEASE - PAGE 2 l NUCUD RF & OCF Ct/ Mw and Dose m Rem (All Activity released) @ Days After Shutdown 05 5.0 70 10 0 30 0 60 0 100 0 150 0 365 0 HFLFE xe130m 1.0E 01 21E+03 5 9E-02 9.7E-59 0OE+00 0 CE*00 0.0E+00 0 OE+00 0 OE< 00 0.0E+00 1.1 E-02 THY 0OE+00 0.0E+ 00 0.0E+00 0 OE+00 0 OE+00 0.0E + 00 0 CE+00 0 OE+00 0 DE+00 0.0E+00 EXT 2 SE+02 51E 01 14E 05 2 3E-62 0OE+00 0 OE+00 0 OE+00 0 OE+00 0 OE+00 0 CE+00 TOT 2 GE+02 5.1E-01 14E-05 2 3E-62 0CE+00 0 OE+00 0 CE+00 0.0E+00 0 OE+00 0 OE+00 l SKIN 4 OE+02 81 E-01 2 2E-05 3 6E-62 0CE+00 0 CE+00 0 CE+00 0 CE+00 0 OE+00 0.0E+00 l uCvoc 4 3E 01 12E 05 19E-62 0CE+00 0 OE+00 0 OE+00 0 OE+00 0.0E+00 0 DE+00 Xe-135 1 OE-01 12E+04 1 1 E+ 01 2.8E-01 1.3E-03 0.0E +00 0.0E+00 0.0E+00 0 OE*00 0 OE400 3 8E-01 Tl(Y 0OE+00 0 OE+00 0 OE+00 0 OE+00 0 CE+00 0 OE+00 0 OE+00 0 OG+00 0 CE+00 0 OE+00 EXT 14E+02 16E+00 1 SE 03 3 8E-05 18E-07 0OE+00 0 OE+00 0.0E+00 0 OE+00 0 OE+00 TOT 14E+02 16E+00 1 SE-03 3 8E-05 18E-07 0OE+00 0 DE+00 0 OE+00 0 CE+00 0 CE+00 SKIN 4 2E+02 4 8E+00 4 3E-03 1 1E-04 5 3E-07 0DE+00 0.0E+00 0 CE*00 0 CE+00 0 OE*00 uCvec 2 4E+00 2 2E 03 5 6E-05 2 7E-07 0OE+00 0 OE+00 0 OE+00 0.0E+00 0 OE+00 Kr 85 3 OE-01 3 OE+02 3 DE*02 3.0E+02 3 OE*02 3 OE+02 3 OE+02 3 OE+02 3.0E+02 2 8E+02 3 3E+03 THY 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 0 0E+00 EXT 13E+00 1 1 E-03 1 1E-03 1 1E-03 1.1 E-03 1.1E-03 1.1E-03 1.1E-03 1.1E-03 1 OE-03 TOT 13E+00 1 1E-03 1 1E-03 1 1E-03 1 1E-03 1.1E-03 1 1E-03 1.1E-03 1 1E-03 1 OE-03 SKIN 1BE+02 1 5E-01 1 SE-01 1 SE-01 1 SE-01 1.5E-01 1 SE-01 1.5E-01 1 SE-01 1 4 E-01 uCvcc 1 8E-01 1 BE 01 1 8E-01 1 8E-01 1.8E-01 18E-01 1.8E-01 1 BE-01 1 7E-01 SUMS THY 42E*02 2 4E+02 2 OE+02 1SE+02 2.7E +01 2.1E+00 6.7E-02 11E-03 2 OE 04 (CASK EXT 3 4E+00 6 6E 01 51E-01 3 5E-01 2 9E-02 2 OE-03 1.1E-03 1 1E-03 1 OE-03 DROP) TOT

2. 0E +01 1 OE+01 8 6E+00 6 6E+00 1 1E+00 8 6E-02 3 8E-03 1 1E-03 1 OE-03 SKIN 9 8E*00 2.3E+ 00 18E400 1.3E +00 2 4E-01 1.5E 01 1.5E-01 1 SE-01 1 4 E-0

1 BNDL THY 3 4E+01 19E+01 16E+01 12E+01 2.2E+00 17E 01 5 4E 03 8 GE-05 16E-05 015 TOT 16E+00 8 4E-01 7.0E-01 5 3E-01 9 2E-02 7 OE-03 31E-04 9 3E-05 8 SE-05 PKING SKIN 8 OE-01 19E-01 1 SE 01 1OE-01 1.9E-02 12E-02 1.2E-02 1.2E-02 1.1E-02 CORE THY 1.9E 403 11E+03 91E+02 7 OE+02 12E+02 9 4E+00 3 CE-01 5 0E-03 91E-04 TOT 9 OE+01 4 7E+01 3 9E*01 3 OE+01 51E+00 3 9E-01 1.7E 02 5 2E-03 4 7E-03 SKIN 4 SE+01 1 1E+01 8 2E+00 5 9E+00 1 1 E +"')

7.0E-01 6.8E-01 6 8E-01 6 4E-01 FULL THY 1.9E + 03 11E+03 91E+02 7.0E*02 1.2E+02 9 4E+00 3.1E 01 9 SE-03 5 4E 03 POOL TOT 9 OE+01 4.7E+01 3 9E+01 3 OE+01 5 2E+00 41E-01 41E-02 2 9E-02 2 BE 02 SKIN 4 BE+01 14E+01 1 1E+01 9 OE+00 4 2E+00 3 9E*00 3.8E+00 3 8E+00 3 8E+00 i m

_.OCT 7 '97 8:37 FROM BRP DECOM PLAN PAGE.017 EA-BRP.DP CHS-1

)CC i) 1-4)

CSKDRP2.WK4 (PAGE 3) - ELEVATED RELEASE DOSES AT SITE BOUNDARY Ci/ Mw and Dose in Rem (All Activity released) @ Days After Shutdown D5 5.0 7.0 10.0 30.0 60 0 100.0 150.0 365.0 SUMS THY 1.2E+02 7 OE+01 5 9E+01 4 SE+01 8.0E+00 6.1 E-01 2.0E 02 3.2E-04 5 9E. (CASK TOT 5.8E+00 3 OE+00 2 SE+00 1.9E+00 3.3E-01 2.5E-02 1.1E-03 3.4E-04 31E-04 DROP) SKIN 2 9E+00 6 8E-01 5 3E 01 3 8E-01 7.0E-02 4 SE-02 4.4E 02 4 4E-02 4.1E-02 1BNDL THY 1OE+01 5.7E+00 4.8E+00 3.7E+00 6 SE-01 4.9E-02 1.6E-03 2.6E 05 4.8E 06 01.5 TOT 4.7E-01 2.5E-01 2.1E-01 1.6E 01 2.7E-02 2.1E-03 9.2E-05.2.7E-05 2.5E-05 ' PKING SKIN 2 3E-01 5 5E 02 4.3E-02 3.1 E-02 5.7E-03 3.7E 03 3 6E-03 3.6E-03 3.3E 03 CORE THY 5.6E402 3 2E+02 2.7E+02 2.0E + 02 3.6E+01 2.8E+00 8.9E-02 1.5E 03 2.7E-04 TOT 2.6E+ 01 1.4E+01 1.2E+01 8 8E+00 1.5E+00 1.2E-01 51E-03 1.5E-03 1.4E-03 SKIN 1.3E + 01 31E+00 2.4E+00 1.7E+00 3.2E-01 2.0E-01 2 0E-01 2.0E-01 1.9E-01 FULL THY 56E+02 3.2E+02 2.7E+02 2.0E+02 3.6E + 01 2.8E+00 91E-02 2.8E 03 1.6E-03 POOL TOT 2 6E+01 1.4E+01 1.2E+01 8.8E+00 1.5E+00 1.2E-01 12E 02 8.4E 03 8 3E 03 SKIN 1.4 E + 01 4.0E+00 3 3E+00 2.7E+00 1 2E+00 1.1 E+00 1.1 E +00 1.1 E+00 1.1 E+ 00 I I ( i,, 12/09/94 06.22 PM

- TOTAL PAGE.Ol? **

i

616 547 8340 OC1 7 '97 9:23 FROM BRP DECOM PLAT 1 PAGE.001 FAX TRANSMITTAL CONSUMERS ENERGY COMPANY 8 BIG ROCK POINT PLANT 10269 US-31 NORTH CHARLEVOIX MI 49720 FAX NUMBER [616] 547-8340 Internal Access: 1371-340 DATE: IO ~1 97 TO: PAUL W HARRIS NAME: coxvisv: US NRC/NRR (301) 415-1169 omcE riroxE uuMnEn: (301) 415-3313 FAX NUMBER: FROM: NAME: Mike Bourassa (616) 547-8244 ) OFFICE PIlONE NUMBER: PAGES TO FOLLOW G Fou 2e%e h ce-< ssio~,se Acceek 6)/v e L Y 5 / S 4 g 1

l OCT 7 '9,7, 9:23 FROM BRP DECOM PLAN PAGE.002 I i TO: RAEnglish, Big Rock Point CONSUMERS i { POWER FROM: RGChristie, P24-106 r COMPANY DATE: October 10, 1994 Internal Correspondence

SUBJECT:

BIG ROCK POINT TISSION PRODUCT INVENTORY CALCULATION USING ORIGEN2 RGC 94-18 { Attached are the results of an ORIGEN2 calculation which was performed to estimate the fission product inventory in a hypothetical core exposed to 33,000 MDW/MTU. All 84 ?ssemblies are initially assumed fresh and are cesumed to remain in the e for 5 years while the reactor continues to operate at-240 MWt, (This is a conservative, but unrealistic ocsumption, as criticality could not be achieved in the fourth and fif th year without replacing some of the fuel.) At shutdown, the release fractions from NEDO-24782 are applied: 100% Noble gases 50% Halogens 1% All others. The results are in units of Ci/MWt. Tha basis for the input file development is also attached. i l n,,i i. ,i i O, i y

I 1 OCT -7 1'97 9:24 FROM BRP DECOM PLAT 4 PAGE 003 I 2 l l ACTIVITY AFTER SHUTDOWN FOR BIG ROCK POINT USING ORIGEN2 1 Arsumptions: 11. All fuel assemblies are identical in composit'su.a. Mu= 131,752 grams / assembly = 0.131752 Te/ assembly (from Kevin Shields) i Enrichment.= 3.4284% (from Kevin Shields) i N = 84 assemblies Power = 240 MWt Natural uranium = 0.000054 Un, l 3.007200 Un3 0.992746 Ung Assume fraction of Un, is unchanged and ' only U s is depleted by n enrichment process. Enriched uranium = 0.000054 Un, 0.034284 Un3 0.965662 Ung Mu, = 0.000054

0.131752 Te/ assembly
84 assemblies

= 0.0005976 Te i Mn3 = 0.034284

0.131752 Te/ assembly
84 assemblies

= 0.3794 Te 0000 .1 j "'. ';'. O,:i,1, -

F 1 OCT 7 '97. 9:24 FROM BRP DECOM PLAH PAGE.004 j l 3 Mn 0.965662

0.131752 Te/ assembly
54 assemblies

= 10.6871 Te = Mu 0.131752. Te/ assembly

84 assemblies l

11.0672 Te = l [ M (2 15.9994) / (238.029 + 2

15.9994)
Mu

= o ~ = 0.11850

84 assemblies
0.131752 Te/ assembly 1.3115 Te

= 1 Divide by core power so results will.be on a MWt basis. M ' n. 0.0005976 Te / 240 MWt = 2.4900 x 10 Te/MWt 4 = l l M'm = 0.3794 Te / 240 MWt = 1.5808 x 104 Te/MWt M ' n, 10.6871 Te / 240 MWt = 4.4530 x 104 = Te/MWt M'u 11.0672 Te / 240 MWt = 5.6.'13 x 104 = Te/MWt M'o 1.3115 Te / 240 MWt = 5.4645 x 104 = Te/MWt 2. All fuel assemblies enter the reactor at the same time, and remain in use until the specified burnup is reached. Burnup = 30 GWD/ ton-U = 30 GWD/ ton-U

1.10232 ton /Te

= 33.07 GWD/MTU Exposure Time = 33.07 GWD/MTU

5.6113 x 104 Te/MWt 1856 days (5.08 years) l Note: This exposure time is in excess of 4 years since the maXigym burnup l

was used, rather than the core average, for all assemblies. I 'I O n 1..t (; 1 OA :l

l - OCT 7 '9.7 9.:24 FROM BRP DECOM PLAN PAGE.005 I 4 l ORIGEN2 Input Deck Development The original NEDO-24782 source core fractions are available for release at shutdown: terms were developed assuming the fo 100% noble gas l 50% halogen 1% all others The implication of this assumption is that at the instant of shutdown the i stated fraction of the core inventory of fission products is separated from the remaining radioactive 1 material, and then decayed taking into account parent-daughter relationships. This assumption has certain deficiencies. For example, since only 1% of the tellurium is assumed present at shutdown, only 1% of the tellurium (parent). feeds 50% of the iodine (daughter). For most radionuclides, this is not a bad simplifying assumption. For I-132 it is a terrible assumption: the half life of Te-132 is 78 hours and that of I-132 only 2.29 hours, so af ter =10 hours the amount of I-132 will decay with the 78 hour half life of To-132. The NEDO-27482 shutdown fraction assumption makes availab.le only 1% of the Te-132 to feed I-132, so the amount of I-132 cfter 10 houte is a f actor of 100 too low. There may be other nuclide decay chains that are similarly affected by this problem. This ORIGEN2 analysis will use the same NEDO-24782 assumption, so that a one-to-one comparison can be performed. The easiest way to do this in ORIGEN2 is to use the default element group fractional recoveries. This is described in Sections 3.5 and 3.6 of the ORIGEN2 User's Manual (ORNL/TM-7175). Set 10 of the element-group fractional recoveries will be redefined to the fractions shown above (Set 10 default values are 0.0 for all element groups, i see Tablo 3.2). Table 3.3 defines the element men.bersh ip in each of the 20 groups that are-available. The default groups already define noble gases and halogens in separate groups, so the input is straight forward. All groups except 12 and 13 are assigned a fractional recovery value of 0.01. Group 12 is the halogens, so it is assigned a value of 0.50. Group 13 is the noble gases so it is assigned a value of 1.00. Since Groups 15-20 have no alements assigned, they are left at the default value of 0.0. These inputs are at the top of the input deck. The first -1 input indicates that individual element fractional recoveries are not to be modified. The n3xt fifteen lines (ending with -1) are the revisions to" the Set 10 ( Oloment-group fractional recovery f actors. The final -l' input indicates that the default element assignments to the groups will be used. I. The OPTL/OPTA/OPTF commands control what output tables are to be produced. Sstting OPTF(7)-7 results in only the fission product summary table, O 0 1:11.1 4 1.. :. O.111

l l OCT' 7 ;9 7,, 9:25 FROM BRP DECOM PLAN PAGE,006 l l l 5 i The uIB command specifies use of the BWR libraries (251/252/253). The PHO command links in the standard photon library, even though the photon tables are not required. l The INP command specifies tnat the input vector will be vector -1. The irradiation of the fuel is done using the IRP command. The IRP's are enclosed between two RUP commands so that the summary output will chow l Information based on the entire irradiation time. The time units of the IRP commands is set to days, and the specific power set to 1.00 MWt. The irradiation is carried out over 1856 days resulting in an 1856 MWD exposure. Since the input mass of uranium (all isotopes) was 5.6113 x 10 Te, the burnup 4 10 33,076 MWD /MTU. f The PRO command is the command that separates the fission product inventory at shutdown using the group recoveries. The shutdown inventory is in voctor -7, the separated material is to be placed in vector 1, the remaining material in vector -10, and group recovery set 10 (indicated by the negative i sign) is to be applied. Remember that set 10 of the group' recoveries was l modifled by the first group of input lines. l The remainder of the inputs are straightforward, except for the CUT commands. l The numerical value of the CUT operand was determined by first running the input deck with the default cut, and then calculating the appropriate value for the final input deck. ~The lines'following the STP command but before the END command is the input l vector. The ORIGEN2 nuclide numbering scheme is used and the input units are { i A 2 on the start of the line indicates that the input is for specific grams. l nuclides, a 4 indicates that the input is for an element, with the natural i obundance used to establish the amount of each nuclide. l l r, e,3 j l i I j

1 OCT '7 '97' 9:2,5 FROM BRP DECOM PLAN PAGE.007 l 6 BRP_FP_0.INP -1 1 10 0.01 2 10 0.01 3 10 0.01 4 10 0.01 5: 10 0.01 6 10 0.01 7 10 0.01 8 10 0.01 9 10 0.01 i 10 10 0.01 i 111 10 0.01 12 10 0.50 13-10 1.00 14 10 0.01 -1 -l' RDA OUTPUT TABLE OPTIONS - OPTL 888 8 8888 8 8888 8888 8 8888 88 OPTA 88 88888 8 888 8 86 38 88888888 OPTF 888888.7 88 8 8888 8888 88888 8 i RDA BAS BIG ROCK POINT - 1 MWt REACTOR i E .RDA VECTOR -1 =' FRESH FUEL AS DEFINED AT END OF INPUT DECK LIP O 0-0 LIB 0 1 2. 3 251 252 253 9 3 0 1 4 PHO 101 102 103 10 TIT INITIAL COMPOSITION OF BIG ROCK 1 MWt REACTOR RDA READ. FUEL COMPOSITION FOR 1 MWt REACTOR BASED ON BIG-ROCK-INP -1 1 11 1 TIT FIVE YEAR IRRADIATION WITH NO SHUTDOWNS RDA-BURNUP - ASSUMES 33,070 MWD /MTU BUP IRP 30.00 1.00 -1 -2 42 IRP 90.00 1.00 -2 -3 4 0 IRP. 180.0 1.00 -3 -4 40 IRP 270.0 1.00 -4 -5 4 0 i' IRP 360.0 1.00 -5 -6 40 IRP 540.0 1.00 -6 -7 4 0 t 'IRP 720.0 1.00 -7 -8 4 0 IRP 900.0 1.00 -8 -9 40 IRP 1080.0. 1.00 -9 -2 40 IRP 1260.0 1.00 -2 -3 4 0 IRP 1440.0 1.00 -3 -4 4 0 IRP 1620.0 1.00 -4 -5 4 0- 'IRP 1800.0 1.00 -5 ~6 4 0 IRP -1856.0 1.00 -6 -7 4 0 RDA BURNUP = (1856 DAYS

1 MWt) / 0. 056113 MTU = 33,076 MWD /MTU

~BUP-u a i.~s s. : d :i m. ::. O d 1 6. I ) l

OCT 7 '97 -9226 FROM BRP DECOM PLAN PAGE.008 7 RDA THE PRO COMMAND WILL SEPERATE OUT THE DESIRED CORE FRACTIONS. RDA VECTOR -7 HOLDS THE TOTAL CORE INVENTORY AT THE END OF THE IRRADIATION. RDA IT IS MUTIPLIED BY ELEMENT-GROUP FRACTION SET 10 (ELEMENT-GROUP i RDA FRACTIONS ARE USED BECAUSE THE LAST ARGUMENT IS NEGATIVE). THE RDA RESULT IS PUT IN VECTOR 1, THE REMAINDER IS PUT IN VECTOR -8. PRO -7 1 -8 -10 DEC 0.5 1 2 4 1 DEC 1.0 2 3 4 0 DEC 3.0 3 4 4 0 DEC 5.0 4 5 4 0 DEC 10.0 5 6 4 0 DEC 30.0 6 7 4 0-RDA CUT VALUE ALLOWS ONLY NUCLIDES > 1. 0E-6 CI (luCi) AT 0.5 DAYS TO PRINT CUT 7 8.09061E-12 -1 OUT 7 1 -1 -1 RDA NEW PAGE DEC 60.0 7 2 4 3 DEC 90.0 2 3 4 0 DEC 120.0 3 4 4 0 DEC 180.0 4 5 4 0 DEC 210.0 5 6 4 0 DEC 364.6 6 7 4 0 RDA CUT VALUE ALLOWS ONLY NUCLIDES > 1.0E-6 CI (luCi) AT 60 DAYS TO PRINT CUT 7 3.07503E-10 -1 OUT 7 1 -1 -1 STP 4 3 922340 2.490 922350 1580.800 922380 44530.000 0 0.0 4 080000 5464.500 0 0.0 i 0 END 1 l l l 4 hl l f fl l I, 9 g mm

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v t e Letter Sequence Approval
TAC:M99688, (Approved, Closed) TAC:M99689, (Approved, Closed)
Initiation Acceptance... Administration Questions 2 February 1998 23 February 1998 Answers Results Approval, Approval, Approval Other: ML20211P867
MONTHYEAR1998-02-23 [Table View]

ML20216C943

Text

SUBJECT:

SUBJECT:

Dear Mr. Mogg:

Enclosures:

2.0 REFERENCES

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Title:

=

Title:

2.0 DESCRIPTION

Reference:

2.0 DESCRIPTION

4.0 REFERENCES

2.0 DESCRIPTION

4.0 REFERENCES

SUBJECT:

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