ML20151W936

Summary of 980902 Meeting W/Duke Engineering & Services in Rockville,Md Re Plans to Transport Haddam Neck Reactor Vessel to Disposal Site in Barnwell,Sc.List of Meeting Attendees & Agenda,Encl

ML20151W936
Person / Time
Site:	Haddam Neck, 07109286  File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/10/1998
From:	Osgood N NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	Chappell C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
NUDOCS 9809160194
Download: ML20151W936 (22)

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p *a UNITED STATES g j NUCLEAR REGULATORY COMMISSION o WASHINGTON, D.c. ensamanni

• • • • • September 10, 1998 MEMORANDUM TO: Cass R. Chappell, Chief Package Certification Section Spent Fuel Project Office, NMSS FROM: ancy L. Osgood, Senior Project Manager Package Certification Section Spent Fuel Project Office, NMSS

SUBJECT:

MEETING

SUMMARY

REGARDING TRANSPORT OF THE HADDAM NECK REACTOR VESSEL Attendees NRC/SFPO Conn. Yankee Atomic Power Co. AE Andrew Barto John Haseltine Melissa Robinson Chris Brown Jim Kay Ross Chappell Gerry van Noordennen Bechtel Andrew Gaunt Keith Sickles Ray Ng Charley Haughney William F. Kane Duke Enoineerino NBC Lawrence Kokajko Cedric Child Elisabeth Mitchell Nancy Osgood Bruce W. Holmgren Susan Shankman D. R. LeFrancois Union of concerned scientists David Tiktinsky Adam Mancini David Lochbaum Bernie White Li Yang Chem-Nuclear Svstems Eloise Ziegler Hisham Shamkhani NRC/ACNW Northeast Utilities Howard Larson Raj S. Harnal NRC/NRR Maine Yankee Atomic Power Co.

Thomas Fredrichs John D. McCann l NRC/OPA Sue Gagner If[7)3 introduction A meeting was held on September 2,1998, in Rockville, Maryland at the request of Duke Engineering and Services. The meeting was held to discuss plans to transport the Haddam Neck reactor vessel to the disposal site at Barnwell, SC. The Haddam Neck power plant is loceted in Haddam, Connecticut, and is undergoing decommissioning. The plant was permanently shut down in December 1996. Duke Engineering requestea the meeting to 9809160194 980910 PDR ADOCK 07109286 C PDR 98-T3 Mhbbbbbbb

C. Chappell inform the NRC of its preliminary plans regarding the preparation of the reactor vessel package for transport, and the evaluation of the package in accordance with 10 CFR Part 71. Duke Engineering will be the applicant for package approval.

Discussion Duke Engineering provided a meeting handout, which is attached. The discussion followed the meeting handout. Duke Engineering emphasized that the information presented in the meeting was prel:minary, and that the plans for the vessel transport may change.

However, Duke Engineering was making the presentation to keep the NRC informed of their current plans regarding the disposition of the reactor vessel.

1. Package Design. The reactor vessel package will be approximately 20 feet in diameter and 33 feet leng, and will weigh approximately 860 tons. The package will be constructed of a steel shell three inches thick, with four-inch thick top and bottom steel closure plates. The reactor vessel head will be removed. The reflective insulation on the vessel will remain in place. The vessel will be filled with low density concrete, then placed in the outer packaging. The void region between the vessel and the outer packaging will be filled with a combination of normal and low density concrete. The top closure plate is connected to the outer shell by a full penetration weld, and will be attached to the vessel by the reactor head studs.

Additional steel plates will be added to the package at the vessel core region for radiation shielding.

2. Contents. The reactor vessel head, and some of the internals will be removed prior to transport. The transport package will contain only Class A, B and C wastes.

Duke Engineering presented preliminary estimates of radioactivity present in the remaining internals and in the reactor vessel. The vessel contains 12,000 curies, the lower internals contain 10,000 curies, and the upper internals contain 12,000.

The total radioactivity in the package is expected to be less than 50,000 curies.

3. Package Evaluation. The package will be evaluated in accordance with the performance standards in 10 CFR Part 71, including normal conditions of transport and hypothetical eccident conditions. Duke Engineering reviend the current plans for evaluating the package, as well as the acceptance criteria tr.at would be used.

The package will be evaluated for a normal conditions drop from the horizontal positiot. For accident conditions, four drop orientations will be considered. There  !

are no lifting or tie-down devices that are a structural part of the package. A  ;

system of cradles and wire ropes will be used to secure the package to the vehicle.

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4. Transport Plans. Duke Engineering reviewed the transport plans for the shipment.

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The package will ba loaded onto a barge on the Haddam Neck site. The barge will i transport the package to the DOE Savannah River Site. The package then will be transferred to a heavy hauler, and trar. sported by road to the waste disposal facility l near Barnwell, SC.

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l .1 C. Chappell 5. Package Fabrication. Duke Engineering reviewed the current plans for package j fabrication. 'The outer packaging will be febricated then transferred into the reactor

building. The reactor vessel will be filled with concrete, then placed within the outer l packaging. Fill holes in the outer packaging will be used to make multiple concrete l pouri. nd to ensure that each concrete placement will be free of voids.

6. Quality Assurance (QA) Program. NRC stated that all activities related to the l design, fabrication and certification of the package must be done under a QA

. program approved by the NRC under Part 71. Duke Engineering is currently l preparing a submittal for approval of their QA plan.

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[ 7. Schedule. Duke Engineering provided a general schedule for the vessel transport L project. The schedule shows transport of the vesselin the spring, year 2000. NRC stated that the period allowed for NRC review (six months) was optimistic.

L Docket No.: 71-9286 l

Attachment:

As stated l

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j. p'stribution w/sttacir  !

Docket File 71-9286 PUBLIC MWHodges DScrenci,R-l JHickey,NMSS

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Distribution w/out attach:

NMSS r/f SFPO r/f Meeting Attendees Meeting Notebook l -' .

j l *C" - Copy without attachinent/ enclosure 'E" = Copy with attachment / enclosure 'N" = No copy l 1

( OFC NMSS/SFPO E NMSS/SFPO C. NMSS/SFPO E  !

NAME LOsgood hERZiegler CRker DATE 9/ldiG8 9//0/98 9//#/98 OFFICIAL RECORD COPY j i l i

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l l MEETING AGENDA CONNECTICUT YANKEE ATOMIC POWER CO.

l DUKE ENGINEERING & SERVICES CHEM-NUCLEAR SYSTEMS l NUCLEAR REGULATORY COMMISSION l l

1 September 2,1998  ;

1:30 PM I 1 White Flint l l Room 6-B-11 i

l Removal Of The Haddam Neck Plant Reactor Pressure Vessel l

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4 HADDAM NECK PLANT REACTOR VESSEL REMOVAL L

MEETING AGENDA 4

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Introduction J. Haseltine Reactor Vessel / Package Description C. Child Waste Characterization A. Mancini

Shielding Evaluation A. Mancini ,

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Package Shipment R. Harnal l

Package Evaluation D. LeFrancois Project Schedule J. Haseltine l

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. 1 HADDAM NECK PLANT REACTOR VESSEL REMOVAL INTRODUCTION Meeting objectives Inform NRC of current HNP RPV transport and disposal plans Outline technical evaluation methodologies Solicit NRC input Establish SAR submittal / review schedule Transport and Disposal Mission Remove reactor vessel following partial segmentation ofinternals to remove GTCC material and ship to Barnwell, SC disposal facility in accordance with 10CFR71 requirements for a Type B package.

Safety Analysis Report Will comply with 10 CFR 71," Packaging and Transportation of Radioactive Material" .

1 Uses guidance from Regulatory Guide 7.9," Standard Format and l Content of Part 71 Applications for Approval of Packaging for l Radioactive Material" l l

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL INTRODUCTION Use Proven Technology and Expertise HNP / DE&S /CNS Team Members Have Successfully Completed Similar Projects:

Yankee Rowe RPV Remove.1 Millstone 2 Steam Generator Replacement Yankee Rowe Steam Generator Removal Yankee Rowe Pressurizer Removal Point Beach Steam Generator Replacement Salem Steam Generator Replacement St Lucie Steam Generator Replacement Catawba Steam Generator Replacement McGuire Steam Generator Replacement (2 Units) 4 N.

4 HADDAM NECK PLANT REACTOR VESSEL REMOVAL  !

REACTOR VESSEL DESCRIPTION 4-Loop Westinghouse PWR l Reactor Vessel And Internals Dimensions & Weights (Without Closure Head) l Vessel OD At Belt line 14 Feet 7-1/2 Inches i

Vessel Wall At Belt line 11 Inches Total Length 30 Feet 9 Inches

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Vessel Weight 423 Tons

Internals Weight (Less GTCC) 62 Tons Total 485 Tons i

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1 HADDAM NECK PLANT  ;

! REACTOR VESSEL REMOVAL  !

PRELIMINARY REACTOR VESSEL PACKAGE DESCRIPTION l

l Overall length 33 feet  !

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Outside diameter 20 feet Design Code ASME Section 111,1995 used as guide I Shell Material: SA-516 Grade 70 Carbon Steel 3 inch thick shell I 4 inch top and bottom covers 2 inch additional shield plates opposite belt line and on top cover Package does not include closure head, GTCC internals or upper internals above vessel flange i

Stainless steel reflective insulation on OD of RPV will be part of l

the package.

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL PRELIMINARY REACTOR VESSEL PACKAGE DESCRIPTION Void spaces filled with low density or standard density concrete except for hollow ring at bottom of packs, ^ 1 1

i Top cover bolted to RPV using modified vessel head studs l

l Only field weld required is top cover to cask after vessel is lifted 1 into cask l

Transport package will be fully compliant with all NRC and DHEC transport and disposal requirements  !

Weight empty cask and covers 222 tons Total package weight 862 tous Total package consists of the following:

RPV (less head)

RPV Internals (less GTCC and some upper internals)

Cask, covers and extra shield plates Internal and annular concrete 7

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t HADDAM NECK PLANT REACTOR VESSEL REMOVAL REACTOR VESSEL PACKAGE DESCRIPTION Proposed CY RPV Sh!pping Package

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TOP SHIELD 7 F i g ., "* j-/ nerate 2 z 2" ACTUAL CLR 1

31/8" NOM. CL l i

11" 50 J NOZZLE l N , PLUGS REFLECTNE INJULATION

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2" lNNER

. SHIELD 50 pcf

'Concreto I

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- 1 150ref

' Concrete

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Hollow Rmg 2 l l-l  !

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230" l 501rr I (1F-11")

NOTE: RPV intamals not shown.

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL REACTOR OPERATING HISTORY Reactor Critical, July 1967 Commercial Power, January 1968 600 Mwe i i

l 1825 MWt 21 Effective Full Power Y .rs 19 Completed Fuel Cycles Shutdown July 21,1996

- Permanent Shutdown, December 1996 9

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i HADDAM NECK PLANT '

REACTOR VESSEL REMOVAL  !

ACTIVATION ANALYSIS Neutron Fluence Code: DORT Discrete Ordinate Transport (2D)

Activation Code (s): ORIGEN-S and ACTIV Decayed to September,1997 l

Comparison Measurements Thermal Shield Survey Post Removal 1992 Vessel External Survey Vessel Internals Survey 10

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l' HADDAM NECK PLANT REACTOR VESSEL REMOVAL l

WASTE CHARACTERIZATION

- <50,000 Curies without GTCC internals '

Major radionuclides are from activation:

Fe55, Co60 and Ni63 10CFR61: All remaining reactor internals and vessel are Class l A, B, or C 10CFR71: Type B quantity, l Requires Advance Notification per 10CFR71.97 L

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HADDAM NECK P.LANT REACTOR VESSEL REMOVAL SHIELDING EVALUATION Package will meet 10 CFR 71.47 dose rate limits '

200 mrem /hr on cask surface 10 mrem /hr at 2 meters from cask surface QAD-UE computer code used for shielding evaluation also used for Yankee Rowe RPV package Source based on:

Activation analysis External vessel surveys i

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL SHIPMENT Secure RPV package on barge Complete barge and tug (s) inspections and survey per C of C and transport plan. Inspection to be done under the supervision of the National Cargo Bureau Barge shipment from HNP site to DOE's Savannah River Site (SRS)

Road transit from SRS to Barnwell l

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL l 1

l NORMAL CONDITIONS OF TRANSPORT l PER 10 CFR 71.71(C)

CONDITION DESCRIPTION i Condition Description 1

(1) Heat 161 F Maximum Cask Temperature (2) Cold Environment -40 F Minimum Cask Temperature (3) Reduced External Pressure +15.3 psi Differential (4) Increased External Pressure -8.5 psi Differential (5) Vibration Normally Incident to Barge Transport Transport Accelerations (6) Water Spray No Effect on Package (7) Free Drop 1 Foot Drop Horiz., Comer Drops (8) Corner Drop Not Applicable (9) Compression Not Applicable (10) Penetration 13 lb.,1.25" dia.

Missile, dropped 40" 14

HADDAM NECK PLANT

. REACTOR. VESSEL REMOVAL NORMAL CON'DITIONS OF TRANSPORT PER 10 CFR 71.71(C)

EVALUATION CRITERIA Condition Criteria u)

(1) Heat Level A Service Limits (2) Cold Environment Level A Service Limits (3) Reduced External Pressure Level A Service Limits (4) Increased External Pressure Level A Service Limits (5) Vibration Normally Incident to Level A Service Limits Transport  ;

(7) Free Drop (1 ft) Level D Service Limits "'

(10) Penetration Penetration Less Than Min.

Cask Thickness l NOTES:

1. Ref 1995 ASME Section III, Div.1, Subsection NB

2. Exclusive use package. Strain limited to % minimum elongation requirement of ASME/ ASTM material specification for localized I tensile stresses; for localized compression, strain is limited to l mmimum specified elongation.

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. HADDAM NECK PLANT l REACTOR VESSEL REMOVAL HYPOTHETICAL ACCIDENT CONDITIONS OF TRANSPORT PER 10 CFR 71.73 (C) l CONDITION DESCRIPTION I

Condition Description (1) Free Drop 30 Foot Drop I (2) Crush Not Applicable l (3) Puncture Horizontal drop onto steel cylinder under package CG (4) Thermal 100 F Initial Ambient 1475 F For30 Minuites (5) Immersion Not Applicable (fissile material)

(6) Immersion Maximum Barge Route Depth l

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HADDAM NECK PLANT REACTOR VESSEL ~ REMOVAL HYPOTHETICAL ACCIDENT CONDITIONS OF TRANSPORT PER 10 CFR 71.73(C)

EVALUATION CRITERIA Condition Criteria (1) Free Drop Release < 10 CFR 71.51

< 1R/Hr at 1 m external (3) Puncture Release < 10 CFR 71.51

< 1R/Hr at 1 m extemal (4) Thermal 1997 ASME Section III, Division 3 Subsection WB l

(6) Immersion Level D Service Limit"*

Notes:

(1) Ref.1995 ASME Section Ill, Division 1, Subsection NB.

j (2) Exclusive use package. Strain limited to % minimum i elongation requirement of ASME/ ASTM material l specification for localized tensile stresses; for localized l compression, strain is limited to minimum specified elongation.

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HADDAM NECK PLANT i REACTOR VESSEL REMOVAL HYPOTHETICAL ACCIDENT CONDITIONS OF TRANSPORT  :

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30 FT DROP ANALYSIS APPROACH -

I e Investigating impact limiter design feasibility impact limiter design must have good crushing strength and energy absorption characteristics ,

Evaluating design requirements 1

e Investigating best analysis tools to predict package l performance DYNA 3D appears to be one possible candidate e Originally developed at LLNL

= Explicit finite element program for nonlinear dynamic response of three-dimensional inelastic structures e Excellent large deformation and contact problem capabilities suitable for cask drop analyses

. Extensively used for analysis of nuclear shipping containers at LLNL, ORNL, for the last 8-10 years; more recently at LANL

! e Used by LLNL for the analysis of the Shippingport RV l package

. Preliminary discussions with Dr. John Hallquist, developer of the code, to provide technical support 18 I

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HADDAM NECK PLANT REACTOR VESSEL REMOVAL SCHEDULE OVERVIEW Completion of SAR analyses except for 30 foot drop Early 4th quarter 1998 l 1

Complete 30 foot drop analyses Mid 4th quarter 1998 Submit SAR January 1999 Request completion of NRC SAR review Mid 1999 Complete cask fabrication Mid 1st quarter 2000 Remove & ship RPV Spring 2000 l

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