Text

Rev 3 to License Change Application 237 Re Spent Fuel Cask Loading in Fuel Bldg
	ML20198T185
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	01/07/1999
From:	; PORTLAND GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20198T150	List: ML20198T145; ML20198T185;
References
	NUDOCS 9901120169
	Download: ML20198T185 (72)
	v • d • e

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Portland General Electric '

l LCA 237 - Spent Fuel Cask Loading l

in the Fuel Building )

Revision 3 1

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Trojan Nuclear Plant 9901120169 990107 PDR ADOCK 05000344 U PDR _

u j l l LICENSE CHANGE APPLICATION (LCA) 237

' SPENT FUEL CASK LOADING IN THE FUEL BUILDING I u

TABLE OF CONTENTS 4 l

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.1. BACKGROUN D AND REASON FOR Cl{ANGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I u m o d

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2. DESC RIPTION QF CHANGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -

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' ; 3. DESCRIPTION OF THE ISFSI COMPONENTS AND EOUTPMENIr . . . . . . . . . . . . . . . . . 2 3.1 G EN ERAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 E 3.2 S Y STEM DESI G N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l a

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- i DESCRIPTION OF THE SPFN[ FUEL CASK LOADING PROCliSS . . . . .. . . . . . . . . .. . . . 5 1

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> 5 ? S AFETY EVA LUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 8 a 5.1 NORMAL , OPERATIONAL CONSIDERATIONS . . . . . . . . . . . . . . . . . .. . . . . . . . . . 9 5.1.1 C.tilici@v Prevention . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . 9 l 112_Eu:LCladding Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 -

5.1.3 Radiation Shielding and Radiolonical Controls . . . . . . . . . . . . . . . . . . . . . . I1 -

5.2 POSTULATED OFF-NORMAL EVENTS AND ACCIDENTS . . . . . . . . . . . . . . . .13 l 5 .2 . I ' D rops . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 3 5.2.1.1 Fuel Assembly Drop into a Basket Loaded with Spent Fuel . . . . 13 .

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- 5.2.1.2 Basket Shield Lid Drop onto a Basket Loaded with Fuel , . . . . . . 14 j 5.2.1.3 Lifling Yoke Drop onto a Basket Loaded with Fuel . . . . . . . . . . . 15 -

5.2.1.4 Structural Faihire of Transfer Cask Lifting Devices . . . . . . . , . . . 16 l

' } 5.2.1.4.1 Transfer Cask Drop into the Cask Loadiagfit . . . . . . . 18 )

i 5.2.1.4.2 TransfttfCask Drop into the Cask Wash Plt . . . . . . . . . 19 5.2.1.4.3 Iransfer Cask Drop into the Fuel Buildinn Holstwgy . . 19 5.2.1.4.[hmsfer Cask Drop onto the Foel Building Floor (93' a Elevation) . . . . ...... . ....................20 C

5.2.l'.5 Transfer Cask Lid Assembly Drop onto a Basket Loaded with l F uel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 q i

5.2.1.6 Basket Structural Lid Drop onto Basket Loaded with Fuel . . . . . 22 ;

5.2.1.7 Basket Lift Rigging Drop onto Basket Loaded with Fuel . . . . . 23

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5.2.1.8 Basket Drop into Concrete Cask . . . . . . . . . . . . . . . . . . . . . 24 ]

L 5.2.1.9 Concrete Cask Shield Ring or Lid Drop onto a Basket Loaded with ,

i F u el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 l l

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i LICENSE CIIANGE APPLICATION (LCA) 237 SPENT FUEL CASK LOADING IN THE FUEL BUILDING TABLE OF CONTENTS 5.2.2 Tinovers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 S.2.2.1 Transfer Cask Tipover with Basket Loaded with Fuel . . . . . . . . . 25 5.2.2.1.1 Transfer Cask Tinover at the Cask Loading Pit . . . . . . . 25 5.2.2.1.2 Transfer Cask Tinover at the Cask Wash Pit . . . . . . . . . 27 5.2.2.1.3 Transfer Cask Tipover at the Fuel Building Hoistway . . 27 5.2.2.2 Transfer Cask Tipover during Concrete Cask Transfer Operations 27 5.2.3 Mishandlin g Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 5.2.3.1 Cask Loading Pit or Spent Fuel Pool Liner Tear / Breach During Handling Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 5.2.3.2 Crane Mishandling Operation with Transfer Cask / Basket 4

Resulting in Horizontal Impacts . . . . . . . . . . . . . . . . . . . . . . . . . 30 5.2.3.3 Interference while Lowering Basket during Transfer Operations . . -31 5.2.' A Misalignment of Transfer Cask Lifting Yoke . . . . . . . . . . . . . . . . 31 5.2.4 Oneradonal Errms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 5.2.4.1 Opening the Transfer Cask Bottom Doors Prior to Attaching the Basket Lifting Rig . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 !

5.2.4.2 Closing the Transfer Cask Bottom Doors while Lowering the !

B ask et . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 l 5.2.4.3 Brittle Fracture of the Transfer Cask . . . . . . . . . . .. . . . . . .. . . . . . 34 ;

5.2.4.4 Boron Dilution of Cask Loading Pit, Basket, or Spent Fuel Pool . 34

'k 5.2.4.5 Spread of Transfer Cask and/or Basket Contamination . . . . . . . . 35 ,

1 5.2.4.6 Inadvertent Basket Over Pressurization during Draining / Drying j M Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 6 5.2.5 Supoort System Malfunctions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 5.2.5.1 Welding System Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 5.2.5.2 Basket Draindown/ Vacuum Drying System Failures . . . . . . . . . . 37 5.2.5.2.1 Loss of Electrical Power or Component Failures . . . . . 37 5.2.5.2.2 Rupia e of a Pressurized Line . . . . . . . . . . . . . . . . . . . . 38 i 5.2.5.3 Air Pad System Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 5.2.6 Natural Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 5.2.6.1 Natural Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 5.2.6.1.1 Tornadoes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 5.2.6.1.2 Floods and Tsunami . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 5.2.6.1.3 Earthauakes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40

. LICENSE CHANGE APPLICATION (LCA) 237 i SPENT FUEL CASK LOADING IN THE FUEL BUILDING l o.

'l ABLE OF CONTENTS .

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. 5.3 OPERATIONAL CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41

'. 5.3.1 Handline Height for the Transfer Cask . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41.

.1 l 5.3.2 Backet Draindown Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 l 5.3.3 Handling Temnerature of the Transfer Cack . . . . . . . . . . . . . . . . . . . . . . . . 43 5.3.4 Air Pad Installation Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 ,

5.3.5 Concurrent Decommissionino Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 -

5.3.6 Auxiliary / Fuel Buildino Ventilation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 '

5.3.7 Recovery from Off-Normal Events and Accidents . . . . . . . . . . . . . . . . . . . 44 i l

' 6. SIGNIFICANT HAZARDS CONSIDERATION DETERMINATIDH . . . . . . . . . . . . . . . . . 44 .

7. ENVIRONMENTAL ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 7.1 ENVIRONMENTAL IMPACT OF PROPOSED ACTION . . . . . . . . . . . . . . . . . . . 53 l

7.1.1 Effects on Human Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 j 7.1.2 Effects on Terrain. Vegetation and Wildlife . . . . . . . . . . . . . . . . . . . . . . . . 54 '

7.1.3 Effects 'on Adiacent Waters and Aauatic Life . . . . . . . . . . . . . . . . . . . . . . . .- 55 7.1.4 Nonha7nrdous/Ha7nrdous Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 1 1

7.1.5 OccunationayPublic Radiation Exoosure . . . . . . . . . . . . . . . . . . . . . . . . . . 56 7.2. ENVIRONMENTAL IMPACT OF OFF-NORMAL EVENTS AND ACCIDENTS . 58 7.2.1 Fuel Assembly Drop into a Basket Loaded with Spent Fuel . . . . . . . . . . . . 59

7.2.2 Transfer Cask Dron and Mishandling Events Prior to Shield and Structural

. Lids Being Welded into the Basket . . . . . . . . . . . . . . . . . . . . 59 7.2.3 Basket Draindown/ Vacuum Drvino System Failure . . . . . . . . . . . . . . . . . . 60 7.3 ALTERNATIVES TO PROPOSED ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 '

7.3.1 Not Removing Soent Fuel from the Soent Fuel Pool . . . . . . . . . . . . . . . . . . 61 i 7.3.2 Shioment of Troian Soent Fuel to Another Facility . . . . . . . . . . . . . . . . . . . 62 I i

8. SCHFDULE CONSIDERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 ;

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A i Enclosure 2 to VPN-005-99 LCA 237, Revision 3 .

.Tanuary 7.1999 i

, L BACKGROUND AND REASON FOR CHANGE e

! 1 Following shutdown of the Trojan Nuclear Plant '(TNP) reactor in November 1992, PGE decided to )

decommission TNP. PGE submitted a Decommissioning Plan dated January 26,1995, that detailed

- - prompt decontamination and dismantlement of contaminated structures, systems and components.

In addition, PGE's Decommissioning Plan stated that relocating the contents of the Spent Fuel Pool to an Independent Spent Fuel Storage Installation (ISFSI) would be the most economical method for storing the TNP spent fuel until a permanent storage facility offsite would be available for offsite l

~ shipment of the spent fuel. Relocating the spent fuel and other associated radioactive waste from l the Spent Fuel Pool to the ISFSI also allows decontamination and dismantlement of structures, !

. systems, and components at TNP to occur sooner than if the spent fuel were left in the Spent Fuel )

. Pool until shipment to the offsite facility is possible. !

l PGE submitted a 10 CFR 72 license application to the NRC for construction and operation of an ISFSI in March 1996 and requested issuance of the 10 CFR 72 license prior to December 1997. As described in the safety evaluation, an evaluation of cask drop events has been completed, as well as an evaluation of a spectrum of postulated off-normal events and accidents. The license condition that prohibited bringing casks into the Fuel Building was removed from the license by License Amendment 196, dated May 19,1997. This LCA is being submitted to permit spent fuel casks to be loaded in the Fuel Building for movement of the spent fuel to the ISFSI on the schedule stated in

- Section 8 'of this submittal.

2. DESCRIPTION OF CHANGE

' Facil.ity Operating License No. NPF-1 was superseded in its entirety by Possession Only License No. NPF-1, which was issued as Amendment 190 to License No. NPF-1 on May 5,1993.

Amendments 191 through 198 have been subsequently issued.

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l Enclosure 2'to VPN-005-99 LCA 237, Revision 3 ii Januarv 7.1999 ,

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PGE proposes that paragraph 2.C(10) be added to License No. NPF-1 to allow movement and l

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L handling of spent fuel casks in the Trojan Fuel Building as described in the 10 CFR 72 heense ,

application. Limiting Condition for Operation (LCO) 3.1.4," Spent Fuel Pool Load Restrictions," l remains in effect and precludes movement of casks over the Spent Fuel Pool.- I The following license condition is proposed to authorize the loading of' spent fuel and other materials into casks in Ge Fuel Building:

"(10) f Londino of Fuel i@ Cash in the Fuel Building L The licensee is authorized to load spent nuclear fuel and other materials into transfer and storage casks in the Fuel Building in accordance with License Change Application (LCA) 237, transmitted by PGE letter dated January 7,1999. Changes to LCA 237 are permitted without changing this l l license providing they are evaluated in accordance with 10 CFR 50.59 and reported in accordance l with 10 CFR 50.71(e)."

A mark-up of the proposed change to Possession Only License No. NPF-1 (as provided by Amendment 198 to License No. NPF-1)is located in Enclosure 3 of this LCA. l

3. DESCRIPTION OF THE ISFSI COMPONENTS.AND EOUIPMENT 3.1 GENERAL PGE has selected Sierra Nuclear Corporation's Transtor Storage System for the Trojan ISFSI.

The TranStor Storage System is a vertical dry storage system which utilizes a ventilated concrete cask and a' seal-welded steel basket to safely store spent nuclear fuel assemblies and fuel debris. l TheTrojan ISFSI will consist of a reinforced concrete pad, supporting a maximum of 36 Sierra

. Nuclear Corporation Transtor Storage Casks. It is anticipated that 34 storage baskets and l Concrete Casks will be required. The baskets will be pressurized water reactor (PWR) baskets, l l

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 ,

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designed to safely store intact spent fuel assemblies, suspect / damaged fuel assemblies, and/or fuel l debris.

. l The system is designed to pennit transfer of the basket to a shipping cask once a permanent offsite repository or other offsite facility is available. The TranStor Storage System is also designed to accommodate recovery from postulated off-nomial events without reliance on the Spent Fuel Pool.

3.2I SYSTEM DESIGN The TranStor Storage System consists of the baskets, concrete casks, a storage pad and associated transfer equipment necessary for safe placement of spent nuclear fuel assemblies and fuel debris l into' dry storage. The baskets are metal containers that are seal welded closed. The baskets serve as l the confinement boundary for the materials that are stored within the baskets. i

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The PWR basket is a fuel storage canister designed to provide safe storage ofintact spent fuel, failed fuel, and fuel debris. The PWR basket consists of an internal sleeve assembly, an outer shell j assembly, a shield lid, and a structural lid. The internal sleeve assembly is fabricated from high strength steel plates formed into an array of 24 square storage sleeves, each holding one PWR spent fuel assembly. The storage sh eyes are sized to accommodate storage of control components within

. the fuel assembly. The shield lid contains penetrations for drying and filling the basket with helium.

. The shield lid and structural lid are welded to the basket shell. The outer shell and lids form a i confinement boundary for the spent fuel. The PWR basket relies only on geometry for suberiticality

- during storage.

Assemblies containing damaged fuel are stored in failed fuel cans in the PWR basket. The four peripheral cells in each PWR basket are oversized to accommodate failed fuel cans and will accommodate spent fuel assemblies as well.

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? A vacuum drying system is used to remove the water from the PWR basket 3 (following loading), l vacuum dry the basket, and backfill the basket with helium. The vacuum drying system is designed 3 ;

i Enclosure 2 to VPN-005-99

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' LCA 237, Revision 3 l January 7 1999 -

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. to evacuate the basket in a stepwise fashion.' During evacuation, the decay heat from the fuel further i helps remove residual moisture from the basket.

The transfer cask is used to move the loaded basket from the Cask Load Pit in the Fuel Building to the concrete cask in the Fuel Building crane bay. The transfer cask is constructed with multiple -

layers of various shielding materials to reduce the radiation exposure to personnel during spent fuel l loading and handling. The transfer cask has retractable doors at the bottom to permit lowering of j the loaded basket from the transfer cask into the concrete cask in order to maintain radiation l shielding during the transfer. The transfer cask is lifted and moved using a special lifting yoke and the Fuel Building overhead crane. 1 The concrete cask provides structural support, shielding, and natumi circulation cooling for the basket. The basket is stored in the central steel lined cavity of the concrete cask. A steel lid is l

placed on the concrete cask to protect the basket from the environment. The concrete cask is i l

ventilated by internal air flow paths which allows decay heat to be removed by natural circulation ;

. around the metal basket wall. Air flow paths are formed by channels at the bottom (air inlets), the ;

1 gap between the basket exterior and the concrete cask interior, and the air outlets. The air inlets and i 1

m outlets are steel lined penetrations that take non-planar paths to minimize radiation streaming. Side j surface radiation dose rates are limited by the thick steel and concrete walls of the concrete cask. ]

- j An air pad system is used to move a loaded concrete cask from the Fuel Building to the reinforced ' l concrete storage pad. . The air pad system lifts the concrete cask a few inches and floats the concrete cask on a cushion of air for movement.

,l The reinforced concrete pad is located about 200' north of the Fuel Building at the general site grade l elevation of about 45'.

i Except for commercially available equipment such as the r2 pad system, the components and equipment described above will be fabricated / constructed ia accordance with specifications and 4

L drawings prior to spent fuel loading and handling. )

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3

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4. DESCRIPTION OF THE SPENT FUEL CASK LOADING PROCESS e ,

F Each basket and the transfer cask are brought into the Fuel Building through the crane bay. After o examination and any needed cleaning, the transfer cask is moved along a safe load path by the Fuel i Building overhead crane and transfer cask lifting yoke to the Cask Wash Pit. The empty basket is

. then moved by the same crane and placed into the transfer cask. After installation ofradiation shielding shime in the gap between the transfer cask and basket, a cask lid assembly is bolted onto the top of the transfer cask. The cask lid assembly, which is a steel ring, ensures that the basket -

i cannot be inadvertently lifted out of the transfer cask while loaded. The cask lid assembly also has shield lid retainers which are retracted to load the basket and are manually extended after the basket is loaded to prevent the shield lid from coming out of the basket in the unlikely event of a tipover.

!- ~ The basket'is then filled with borated water, This filling may be done in the Cask Wash Pit or at the

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j: Cask Loading Pit before submergence. !

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! - The transfer cask (with empty basket) is then moved along a safe load path by the Fuel Building

' overhead crane and suspended over the Cask Loading Pit. A hose will be connected to a flushing i !

system connection' located near the bottom of the transfer cask designed to continuously flush

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borated water through the basket / transfer cask gap to minimize unnecessary contamination of the

[ basket external surface while the transfer cask is' submerged in the Cask Loading Pit. The water is -

filtered and/or demineralized, if necessary, to reduce the potential for contamination on the exterior of the basket', After the transfer cask is lowered onto the impact limiter at the bottom of the Cask Loading Pit, the lifting yoke will be removed and the pit filled with borated water. 1 1

The Spent Fuel Pool gate, which sepastes the Spent Fuel Pool from the Cask Loading Pit will be opened and spent fuel assemblies and failed fuel (e.g., non-intact fuel assemblies, non-fuel bearing ;

components, fuel rod storage unit number 1 and spent fuel pool deoris in process can capsules) will )

be loaded into the basket using the Fuel Handling Bridge crane. Handling operations in the Fuel l .

L Building that involve spent fuel movement will be directly supervised by Certified Fuel Handlers in accordance with approved procedures. Spent fuel assembly serial numbers will be verified and ;

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[ Enclosure 2 to VPN-005-99

' LCA 237, Revision 3 January 2 1999 o

recorded during the loading process. Failed fuel assemblies will be placed in a failed fuel can.

When basket loading is completed, the basket shield lid is placed in the basket while in the Cask

, Loading Pit.'

The loaded transfer cask is lifted from the Cask Loading Pit, the transfer cask is washed on the exterior to reduce contamination and the shield lid retainers are extended into position. Some water.

C above the shield lid is drained to prevent spillage. By use of the Fuel Building overhead crane, the

, : transfer cask is moved along a safe load path. Impact li. miters are placed as needed. It continues l

. along the safe load path to the west side of the Decontamination and Assembly Station (DAS)just g ' north of the Cask Wash Pit.

Decontamination of the transfer cask, if required, and shield lid welding may begin as soon as the tr nsfer cask is located in the DAS. The water level in the basket is lowered to ensure the shield lid weld is not affected by water percolation while providing adequate shielding. Shims are installed between the shield lid and the basket shell and the shield lid is welded. The basket is refilled with borated water and pressure tested in accordance with ISFSI SAR Section 5.1.1.2 " Basket Loading and Sealing Operations." After pressure tests have been successfully completed, the structural lid and backing ring are installed. The structural lid is welded and the weld between the two lids

' (inside the valve access port) is perfonr.ed. The radiation shielding shims are removed, as required, from the top' of the basket area to allow for completion of contamination surveys of the basket external surface. The exterior of the basket will be checked for loose surface contamination (to the extent possible because oflimited access while in the transfer cask) to determine if decontamination 7of the baske't is required. Ifloose surface contamination levels exceed 10d pCi/cm2 p-y or 10-5 pCi/cm 2a, then an evaluation will be performed to determine if decontamination of the basket is required.

Draindown and evacuation of the basket is initiated by pumping the borated water from the basket.

Residual moisture in the basket is removed by blowing dry nitrogen or other inert gas through the basket via the basket drain line (maximum pressure will be controlled to 15 psig) and out the vent line for a minimum of 15 minutes and until no water is visible coming from the vent line. The 6

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. Enclosure 2 to VPN-005-99.

LCA 237, R6 vision 3

' January 7.1999 - i

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outlet for the vent line will be connected to a suitable filtration system to minimize the possibility of l

particulate airborne contamination. ;

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j ' The vacuum drying system is used to perform multiple evacuations to achieve a stable internal basket vacuum tor a minimum of 30 minutes. The basket is then flushed with helium and the

' evacuation process is repeated.

The scaled basket is filled with helium. The vacuum drying systera is used to regulate internal pressure.-

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The vacuum drying system attachment couplings are disconnected and the two penetration cover k

plates are welded in place in the structural lid. By use of the Fuel Building overhead crane, the transfer cask containing the sealed basket is moved from the DAS along a safe load path to the Fuel ,

.1 Building hoistway; An impact limiter is positioned at the 45 ft. elevation of the hoistway prior to movement of the transfer cask over the crane bay area. The loaded transfer cask is then moved to -

the hoistway over two transfer cask guide beams, then rotated to fit through the guide beams and lowered into the crane bay. A prepared concrete cask is moved under the loaded transfer cask. The

,, - hydraulic cylinders for the transfer cask doors are installed. Ceramic tiles in the bottom of the -

concrete cask prevent the stainless steel basket from resting directly upon the carbon steel liner of the concrete cask., . !

The top of the' concrete cask walls are protected to minimize contamination from spreading from the bottom of the transfer cask to the concrete cask. The transfer cask is placed on top of the concrete

cask and correctly positioned by the use of alignment holes located on each side of the transfer cask.

Aner removing the transfer cask lining yoke and installing the basket lifting slings, the bas.ket is

lined slightly by the Fuel Building overhead crane to remove the weight from the transfer cask

- bottom doors. The bottom doors are opened, and the basket is lowered into the concrete cask.

When the basket is firmly resting on the ceramic tiles at the bottom of the concrete cask, the basket ilift rigging is removed with the aid of an extension device. The transfer cask bottom doors are then closed. The hydraulic ~ system is removed from the transfer cask, the lifting yoke is re-attached to 7

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f Enclosure 2 to VPN-005-99

' LCA 237, Revision 3 .

January 7.1999 ,

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the transfer cask and the transfer cask is lifted from the concrete cask to an elevation slightly above the top of the concrete cask. An air pad system is used to move the concrete cask to the west just out of the hoistway area.. The empty transfer cask is then moved from the hoistway back to the- '

Cask Wash Pit where the interior of the transfer cask is checked for loose surface contamination.

This is to provide an indirect indication of the surface contamination levels on the exterior of the -

basket that was just removed from the transfer cask. If contamination levels exceed the limits ,

2 2 previously stated (104 pCi/cm p-y or 10 5 pCi/cm a,), then the need to perform additional ,

^ decontamination of the basket previously placed in a concrete cask will be ' evaluated . l

- The shield ring is installed on top of the concrete cask and the concrete cask cover plate is bolted :

into position. A tamper indicating wire is threaded through cover bolts. The air pad system is removed.' The concrete cask exterior is surveyed for contamination and radiation levels are .

measured. In addition, a di!Terential temperature inspection will be performed at the inlet and oatlet '

air vents before moving the loaded concrete cask to the reinforced concrete storage pad.

The air pad system is inserted under the concrete cask in the air inlet openings and inflated by a standard service air compressor. The cask is moved by forklift, tractor, or other vehicle to the :

reinforced concrete storage pad. Once in position on the concrete storage pad smface, the air pad j system is removed from air inlets and the air inlet screens are installed. ;

1 5.' SAFETY EVALUATION' ;

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The safety evaluation first discusses nonnal operational considerations: criticality prevention, spent j

. fuel cladding integrity, and radiation . shielding and radiological controls. The second part of the .l safety evaluation addresses postulated off-normal events and accidents which are grouped into five categories:' drops, tipovers, mishandling events, operational errors, and support system malfunctions. The last part of the safety evaluation summarizes operational controls that are implemented to ensure that the assumptions of the safety evaluation and supporting analyses are l satisfied during the spent fuel loading and handling process.

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1 Encicsure 2 to VPN-005-99 LCA 237, Revision 3

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This sinfety evaluation is intended to address the complete spent fuel and process can capsule loading and handling process from the Spent Fuel Pool to the ISFSI storage pad. However, this y o- l safety evaluation does not address collection, processing, and loading ofWe spent fuel debris into ,

the process can capsules. A separate safety evaluation was prepared for spent fuel debris

)

collection, processing, and loading and was provided in LCA 239, which was issued as License Ametidment 198.

L-1 4 C5.1 NORMAL OPERATIONAL' CONSIDERATIONS 2 '

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. 5.1.1 Criticality PreventiQn A criticality analysis was performed for both flooded and dry conditions that show that the spent _

g. : fuel in the basket remains suberitical (km less than 0.95) during transportation and storage. The I r enticality analysis for transportation bounds the spent fuel loading and handling process. :The -

. criticality analysis assumed zero fuel burnup,4.2% U-235 fuel enrichment,24 fuel assemblies N' 4

[ loaded in the most reactive configuration,75% credit for boron plates, and the basket filled with

pure water. These assumptions ate conservative because each Trojan fuel assembly has some i

, L burnup, the maximum initial enrichment of Trojan fuel was 3.56%, and the basket will be filled !

[ ,

with borated water (2000 ppm boron or greater). In a dry condition, no moderator exists and K, :

WW ! drops even lower. The criticality analysis was performed using the KENO-Va module of the .

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gSCALE-PC Modult.r Code System for Performing Criticality Safety Analysis for Licensing

!?. ' Edluation, Version 4.1.' .

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5.121ml Cladding Integrity i 1

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The fuel cladding serves as a fission product barrier for intact fuel assemblies, therefore, this barrier ;

3 is maintained during the fuel loading process to minimize the potential for radiological releases. l Wlien the shield lid has been welded to the basket in the Decontamination and Assembly Stanon, j

the basket becomes an additional fission product barrier. The fuel cladding integrity may be

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 breached by physical mishandling or elevated temperatures that cause excessive stress. Physical mishandling (i.e., dropping a fuel assembly) is not likely due to stringent fuel handling prccedures, but is discussed under postulated off-normal events and accidents. Effects due to elevated temperatures are discussed in the following paragraphs.

A thermal analysis was performed to ensure that the spent fuel would not exceed the short term temperature limit of 1058 F during the loading process described in NUREG-1536, " Standard Review Plan for Dry Cask Storage System," Section 4.0, " Thermal Evaluation," Part V.1, " Spent Fuel Cladding." Exceeding 1058 F could adversely affect cladding integrity based on PNL-4835,

" Technical Basis for Storage of Zircaloy Clad Spent Fuel in Inert Gases" (Pacific Northwest Laboratory). The thermal analysis considered the basket in the transfer cask with a helium atmosphere and the basket in the transfer cask with a vacuum. The thennal analysis used a finite element model generated using ANSYS, Revision 5.0A. The results of the 26 kW longer term equilibrium thermal analysis were that the spent fuel cladding temperatures were considerably less than the 1058 F limit. The calculated values are equilibrium values and, therefore, not time dependent. Therefore, elevated temperatures -experienced by the fuel cladding during the loading piocess will not alTect cladding integrity. This is conservative because Trojan's loading pattern will not produce more than 18 kW of heat and currently proposed Technit ' Specifications limit the cask loading to 24 ~kW.

In order to determine the applicability of the 1058 F short-term limit for spent fuel clad temperature in NUREG-1536, the clad hoop stresses of the most limiting Trojan spent fuel were calculated and compared to the hoop stresses of the 28,000 MWD /MTU burnup fuel. As part of the explanation of Table 5, of PNL-4835, the maximum clad hoop stress for PWR fuel with 28,000 MWD /MTU burnup war 140 MPa at 325 C (617*F). At this same temperature,325 C, the maximum clad hoop stress of the limiting Trojan spent fuel rod was calculated to be 105 MPa. From this, it can be ;

concluded that the fuel tested, as described in Table 5, of PNL-4835, bounds the Trojan spent fuel and, therefore, the 1058'F temperature limit is applicable. Trojan fuel temperatures will remain below the limit for facil!.ty operations (e.g., initial loading, shipping cask loading and/or overpack loading) and worst-case credible accidents. It is further concluded that no administrative time limit

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,. 3 Enclosure 2 to VPN-005-99 '

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L l is required for the duration of fuel loading and vacuum drying operations for fuel clad considerations. I

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i Fuel cladding integrity is also mamtained by preventing oxidation that would be caused by

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prolonged exposure of the spent fuel to the atmosphere. Contact with the atmosphere may only be l for a few hours during the loading process, such as, during basket welding. The basket is then p; - drained and evacuated to a vacuum condition. After evacuation, the basket is filled with helium 1 which prevents oxygen from coming in direct contact with the fuel cladding during long term P

istorage. In addition, the basket lid is checked for helium leakage after being welded in place to ensure that the helium will remain in the basket for the storage period. Therefore, the likelihood of i

fuel cladding degradation by oxidation is minimized during the loading and storage process. 1 l

l The spent fuel to be stored in the ISFSI also inchues spent fuel pool debris and fuel assemblies -

1 identified as having suspect or failed fuel rods.. The debris will be loaded into and sealed inside a '

t . separate cordainer (process can capsule) that will serve as a containment boundary. Although j 1

cladding integrity has not been maintained for the suspe.:t/ failed fuel, release of fission products from the suspect / failed fhel is not anticipated during the loading and handling process. The fa4ed

$2 fuel cans are designed to contain fuel pellets under the maximum vertical and horizontal loads. Most i

of the suspect / failed fuel rods were identified on or before April 1982, thus, will have had 15 years l

. of storage in the Spant Fuel Pool prior to loading into a spent fuel cask. Short-lived fission products

.will have decayed and the longer lived fission products will have had several years to be released 4

from the suspect / failed fuel rods into the Spent Fuel Pool water. Therefore, even though the ;

i

. suspect / failed fuel does nat have cladding for a fission product barrier, the lack ofintact cladding

[ # will not represent a substamive radiological hazard during spent fuel loading and handling. The

. bdsket serves as the fission product barrier after being welded.

f, l 1].3 Radiation Shielding and Radiological Controls

~ The design of the basket, transfer cask, and concrete cask provides radiation shielding for the

. workers who will be loading and handling the spent fuel basket, transfer cask, and concrete cask.

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3

. January 7; 1999 The loaded basket will be inside the transfer cask or concrete cask throughout the loading and handling process including the time when the basket is being lowered from the transfer cask to the ,

concrete cask. The working dose rates (at I meter) are estimated as 0.141 rem /hr for the side surface transfer cask,0.180 rem /hr on top of the basket (where the basket lid welding will be

- " . performed),0.010 rem /hr for the side surface of the concrete cask, and 0.134 rem /hr for the top of the concrete cask.

Spent fuel loading of the basket will be performed underwater in the Cask Loading Pit with water over the basket to provide radiation shielding. Individual fuel assemblies that are being moved from l the Spent Fuel Pool to the Cask Loading Pit will be maintained about 10' underwater by the design of the Fuel Handling Bridge crane to provide radiation shielding for the fuel handlers. A shield lid will be placed on the basket before the loaded transfer cask is lilled from the Cask Loading Pit to provide radiation shielding while ihe basket is being moved to the Decontamination and Assembly Station (where the shield lid is welded to the basket).

The exterior of the transfer cask will be washed down upon removal from the Cask Loading Pit to minimize the spread of contamination in the Fuel Building. Plastic sheeting or other material will be used to minimize the spread of potential contamination from the bottom of the transfer cask to the top of the concrete cask. The basket, transfer cask, and concrete cask will be surveyed for contamination prior to moving the concrete cask out of the Fuel Building to ensure that contamination is not spread to the ISFSI storage pad.

The draining and vacuum drying process for the basket will ute filtration equipment to minimize airborne contamination and will be performed in the Fuel Building which has a monitored l 1 ventilation exhaust to ensure that any radioactive release is quantified and verified to be within -

established limits.

The loading and handling operations for the casks will be subject to the Trojan Radiation Protection Program. The doses received by occupational workers will be in accordance with 10 CFR 20 and as 12 i

Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 low as reasonably achievable (ALARA) in accordance with the Trojan Radiation Protection Program.-

5.2 POSTULATED OFF-NORMAL EVENTS AND ACCIDENTS The postulated off-normal events and accidents discussed below are grouped into five categories:

drops, tipovers, mishandling, operational errors, and support system malfunctions. Within these categories, the discussion is provided ic approximately the chronological order that these off-normal events or accidents could occur during the loading process. The postulated off-normal events and accidents that result in radiological consequences at the site boundary are listed for reference in Table 1. Recovery procedures, as described in Section 5.3.7, will be developed. The following

. event descriptions, therefore, only describe the immediate consequences of the event from a criticality and/or radiological release perspective.

' S.2.1 DEDS Drops ofloads onto fuel and drops of the transfer cask were analyzed to ensure offsite releases would be less than EPA Protective Action Guide (PAG) limits. Acceptance criteria for the drops included limiting damage to one fuel assembly for drops onto fuel; limiting decelerations of the fuel assemblies for drops of the transfer cask prior to closure; and limiting decelerations of the closed basket. Drops of the transfer cask onto structural elements were analyzed to preclude failure of the structural elements that could lead to subsequent drops and/or unacceptable loads on the cask and fuel.

5.2.1.1 Fuel Assembly Drop into a Basket Loaded with Spent Fuel The fuel tusembly drop is postulated to occur while spent fuel is being loaded into the basket when the basket is inside the transfer cask and submerged a the Cask Loading Pit. The design safety factors, load testing requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) for the fuel handling equipment minimize the possibility of a fuel 13

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. Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 l

assembly drop actually occurring. ' As described below, the fuel assembly drop will not cause spent fuel damage that results in a radiological release that exceeds regulatory limits or a k,y of more than 0.95. (Cask 'Loading Pit and Spent Fuel Pool liner damage from the dropped fuel assembly is discussed in'section 5.2.3.1 below.) -

The radiological consequences of a fuel assembly drop into the basket will be bounded by the existing analysis in the Defueled Safety Analysis Report (DSAR). The fuel will have had a I minimum of 6 years for the fission products to decay. The analysis in the DSAR only considers 6 l l months of decay time. The radiological consequences at the site boundary for the DSAR analysis are 0.0005 rem whole body,0.0006 rem thyroid, and 0.0455 rem skin. These doses are well below the EPA Protective Action Guides of I rem whole body,5 rem thyroid, and 50 rem skin for the i early phase of an event.

~ If the dropped fuel assembly is stacked vertically on top of another fuel assembly, or inclined at any -

l angle on top of the basket, the fuel in the dropped assembly would be 12 inches or greater from the active fuel region of the fuel assemblies in the basket, which would neutronically decouple the two

_ groups of fuel. Therefore, a dropped fuel assembly will not result in a k,,r that is 0.95 or greater.

5.2.1.2 Basket Shield Lid Drop onto a Basket Loaded with Fuel A basket shield lid drop is not postulated to occur after spent fuel is loaded into the basket (while submerged in the Cask Loading Pit) and the shield lid is being lowered into the basket to provide top axial radiation shielding prior to lifting the transfer cask from the Cask Loading Pit. In lieu of

~ determining the consequences of this event, the guidance of NUREG-0612 is being used to minimize the possibility of this event to the degree that the event need not be considered credible.

The design safety factors for the shield lid lifting equipment will be consistent with NUREG-0612 criteria, paragraphs 5.1.6(b)(ii) and 5.1.l(5) which specify using twice the safety factors specified in the guidelines of ANSI B30.9-1971," Slings." Load testing requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) will also be implemented.

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Enclosure 2 to VPN-005-99

LCA 237, Revision 3 January 7.1999

. Implementing these design features and controls makes the possibility of a shield lid drop extremfy small in accordance with NUREG-0612. Therefore, the consequences of a shield lid drop need not

. .be determined.

I i 5.2.1.3 Lifting Yoke Drop onto a Basket Loaded with Fuel The lifting y~oke drop is postulated to occur after spent fuel is loaded into the basket and the yoke is being lowered to lift the transfer cask out of the Cask Loading Pit. When the yoke is being lowered into the Cask Loading Pit, the shield lid is already inside the basket on the shield lid support ring.

~ Therefore, the analysis of the lifting yoke drop considers the ability of the shield lid and support

. ring to withstand the yoke impact. The design safety factors, load testing requirements, and l . administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling equipment minimize the possibility of a lifting yoke drop actually occurring. As described below, the lifting yoke drop will not cause spent fuel damage that results in a radiological release or a k,y of -

Emore than 0.95. (Potential damage to the Cask Loading Pit liner from the lifting yoke is discussed

. in section 5.2.3.1 below.)

The analysis considered a drop of the lifting yoke from 28' above the transfer cask when the yoke is being lowered into the Cask Loading Pit. The lifting yoke drop is conservatively assumed to be in air for the entire 28'. .

M The analysis shows that the shield lid, which is composed 'of a 3" and a 5" plate on top of each other, would plastically deform less than 2" and the shield lid would not grossly fail. In addition, the shield lid support ring inside the basket would not fail and would prevent the shield lid from

continuing further into the basket. If the lifting yoke impact is evaluated for perforation only, i.e.,

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no deformation of the shield lid other than the penetration, the lifting yoke would penetrate less than 2" through the 8" shield lid. Therefore, the fuel assemblies would not be damaged and there would be no radiological release and no increase in k,g. j 15

l Enclosure 2 to VPN-005-99 LCA 237, Revision 3

' January 7.1999 The lining yoke entering the water may spill some water from the Cask Loading Pit onto the Fuel Budding f!oor at the 93' elevation. The water would cause minimal exposure to workers because the water would be only slightly contaminated. Most of the water would be captured by floor drains. Clean up of the remaining spilled water would be relatively straightforward and not hazardous.

' 5.2.1.4 Structural Failure of Transfer Cask Lifting Devices Structural failure of the lifting yoke, a transfer cask trunnion, or other transfer cask lifting devices, e.g., sling, ring, or hook, is postulated to occur after spent fuel is loaded into the basket. The result of one of these failures would be a transfer cask drop when the transfer cask contains a basket loaded with spent fuel.-

The transfer cask drop analysis considers the spectrum of drops that could occur from the time the transfer cask is lifted from the Cask Loading Pit, after being loaded with fuel, until the time the basket loaded with fuel is lowered from the transfer cask into the concrete cask in the Fuel Building crane bay.- The design safety factors and load testing requirements for the lifting yoke, trunnions, other transfer cask lifting devices, and load handling equipment minimize the possibility of structural failure actually occurring, In addition, the maximum handling height of the transfer cask

-is procedurally limited. . Impact limiters and load distribution assemblies are used, where required, to ensure that the results of the analysis are bounding for any potential transfer cask drop. As described below, a transfer cask drop will not result in a radiological release that exceeds regulatory limits and will not result in a criticality concern.- (Potential damage to Cask Loading Pit liner from i the dropped transfer cask is discussed in section 5.2.3.1 below.)

The transfer cask drop analysis did not postulate a transfer cask' drop into the Spent Fuel Pool. The safe load path is sufficiently far from the Spent Fuel Pool, including lowering and lifting the transfer cask into and out of the Cask Wash Pit and Cask Loading Pit. Mechanical stops and electrical interlocks on the crane used to lift the transfer cask will ensure that sufficient distance from the Spent Fuel Pool is maintained. The floor over which the transfer cask is carried, with impact 16 L

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3

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. limiters and steel load distribution assemblies at required locations, has suflicient capacity to absorb the impact of the transfer cask in the unlikely event of a drop. The transfer cask will not be able to tipover and roll into the Spent Fuel Pool because the transfer cask lift height above the Fuel

- Building 93' elevation floor is well below the height at which the transfer cask could tipover.

If a transfer cask that contains only intact fuel is dropped, the analysis shows that a radiological release from intact fuel will not occur and karwill not increase to 0.95 or greater. The accelerations f . from these drops are less than the maximum allowable accelerations of 82g vertical and 44g horizontal. After the shield and structural lids are welded to the basket, the basket serves as the

- confinement boundary. The analysis shows that the drops that occur after the shield and structural

,- lids are welded to the basket do not exceed the maximum allowable accelerations of 124g vertical and 44g horizontal for the basket and shield / structural lid assembly. Therefore, a radiological

[ release will not occur as a result of a transfer cask drop after the shield and structural lids are welded to the basket.

1-If a transfer cask that contains failed fuel is dropped, the accelerations from a transfer cask drop may cause fuel pellets to leak from the fuel pins. Accumulation of fuel pellets at the bottom of a failed '

fuel can was evaluated for criticality. For each fuel assembly type, the optimum pitch for a rod

=

array inside the failed fuel can was determined. This is considered the most reactive possible contents for a failed fuel can, and is assumed in the TranStorm basket criticality analyses. No ;

possible configuration will be more reactive than the optimum pitch array, even if total rod failure is assumed and pellets are free to move about the failed fuel can. Thus, any number of failed fuel rods may be loaded into the failed fuel cans, even ifit cannot be determined that the rods will (structurally) survive a cask drop event. Therefore, failed fuel will not present a criticality concern as a result of a transfer cask drop.

If a transfer cask that contains failed fuel is dropped prior to the shield and structural lids being welded to the basket, a radiological release of fission product gases from the failed fuel could result

- because the failed fuel may not be able to withstand the same accelerations as the intact fuel in the basket. As discussed earlier, most of the suspect / failed fuel rods were identified on or before April 17 4

Enclosure 2 to VPN-005-99

- LCA 237, Revision 3 January 7.1999 1982, thus, will have had 15 years of storage in the Spent Fuel Pool prior to loading into a spent fuel cask. Short-lived fission products will have decayed and the longer lived fission products will have

had several years to be released from the suspect / failed fuel rods into the Spent Fuel Pool water.

. Therefore, the radiological release may be negligible. However, for conservatism, the radiological l consequences for dropping the transfer cask, prior to the shield and structural lids being welded to the basket, were determined by assuming 100% failure of the fuel pins in 4 suspect / failed fuel assemblies that could be loaded into the 4 oversized storage cells in a basket and 100% of the

! Krypton 85 was released. With these conservative assumptions, the dose at the site boundary would be about 0.003 rem. This dose is used in the drop event discussions that follow.

The postulated drops into the Cask Loading Pit, Cask Wash Pit, and Fuel Building hoistway and onto the floor in the Fuel Building are summarized in the paragraphs that follow.

5.2.1.4.1 Transfer Cask Drop into the Cask Loading Pit l

A transfer cask drop into the Cask Loading Pit was assumed to occur from elevation 93'8", which is about 6" above the curb elevation at the top of the Cask Loading Pit, to the top of an impact limiter at the bottom of the Cask Loading Pit (bottom is elevation 49' 4"). Only vertical and edge drops l were considered because the transfer cask will not fit sideways in the Cask Loading Pit. (Note that liner tears are addressed in section 5.2.3.1 below and a sideways tipover where the transfer cask -

impacts the opposite side of Cask Loading Pit is addressed in section 5.2.2.1.1 below.) -

The analysis shows that the transfer cask would drop through air, enter the water, and fall through the water to the impact limiter at the bottom of the Cask Loading Pit.

The deceleration force on the fuel as a result of the impact would be less than the maximum allowed 82g for intact fuel. Therefore, no damage would occur to intact fuel. As discussed in 5.2.1.4 above, this deceleration may cause damage to suspcet/ failed fuel pins. Therefore, a radiological release is postulated with a dose at the site boundary of 0.003 rem, which is well below the EPA Protective Action Guide of I rem whole body for the early phase of an event.

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l Enclosure 2 to VPN-005-99 LCA 237, Revision 3 Januarv 7.1999 l 1

There would be no change to k, from the intact fuel and the failed fuel would not represent a criticality concern, as discussed in 5.2.1.4 above, if fuel pellets were to leak from the failed fuel pins and accumulate at the bottom of the failed fuel cans.

If the transfer cask is dropped into the Cask Loading Pit, any resultant spill of water caused by the transfer cask entering the Cask Loading Pit would cause minimal exposure to workers because the water would be only slightly contaminated and would not be unduly difficult to clean up.

'5.2.1.4.2 Transfer Cask Drop into the Cank Wash Pit

~ A transfer cask drop into the Cask Wash Pit is not a fuel drop failure concern. The only time a l transfer cask will enter the Cask Wash Pit is when it is empty of fuel.

l The Cask Wash Pit may be used for surveying the inside of the transfer cask after the loaded basket i has been set into a Concrete Storage Cask, for loading and preparing an empty basket or for l temporary storage. Movement of the transfer cask in and out of the Cask Wash Pit will follow a j safe load path. ;

i Drop analysis in the Cask Wash Pit from a transfer cask with an unloaded basket indicates that there will be no damage of plant equipment important to safe storage of spent fuel.

l 5.2.1.4.3 Transfer Cask Drop into the Fuel Building Holstway l l

A transfer cask drop into the Fuel Building crane bay was assumed to occur from 93' 8", which is about 8" above the top of the Fuel Building hoistway, to the top of an impact limiter at the bottom of the Fuel Building crane bay (bay floor elevation is 45'). End, flat, edge, and corner drops were considered because the Fuel Building hoistway is large enough for the transfer cask to drop through sideways.

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3~ I L ' January 7.1999 '

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The structural integrity of the concrete slab in the Fuel Building bay was not considered because the slab is at ground level.

L These decelerations are less than the maximum allowables for the intact fuel of 82g vertical and 44g l- horizontal, hence, no damage would occur to intact fuel and there would be no increase in k, from intact fuel. 'As discussed in 5.2.1.4 above, these decelerations may cause damage to suspect / failed fuel pins. The failed fuel would not represent a criticality concern, as discussed in 5.2.1.4 above, if fuel pellets were to leak from the failed fuel pins and accumulate at the bottom or sides of the failed fuel cans.

The decelerations are less than the maimum allowables for the basket pressure boundary of 124g vertical and 44g horizontal. Since the shield and structural lids are welded to the basket, the basket pressure boundary serves as the confinement boundary and there would be no radiological release.

This is true even for suspect / failed fuel inside the basket, because the basket pressure boundary would remain intact.

5.2.1.4.4 Transfer Cask Dron onto the Fuel Building Floor (93' Elevation)

A transfer cask drop onto the Fuel Building floor slab (elevation 93') was assumed to occur while moving the transfer cask along the safe load paths between the Cask Leading Pit, the-1 Decontamination and Assembly Station, the Fuel Building hoistway, and the Cask Wash Pit. Only vertical drops were considered because the transfer cask will not be lifted high enough off the floor to tip over into a horizontal position. The maximum lift height will be 15" or less above the floor depending on the location along the safe load path. The maximum lift height will be procedurally controlled and bounded by the analysis (Section 5.3.1).

Along the safe load path, the transfer cask will be moved over floor slabs supported by a combination of rigid shear walls and steel beams. Multiple cases, which bound the floors structural

. configurations along the safe load path, were analyzed. The floors along the safe load path were L shown to have sutTicient strength to withstand a transfer cask drop when the following components L

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I Enclosure 2 to VPN-005 99 LCA 237, Revision 3 - '

- January 7.1999 are utilized during a fuel loade'd transfer cask movement: 1) A removable impact limiter either attached to the base of the transfer cask or set on the floor slab in areas along the load path between the Cask Load Pit and the Decontamination and Assembly Station,2) steel load distribution assemblies set on the 93 ft. elevation slabjust south of the Fuel Building hoistway, and 3) transfer cask guide beams installed in the hoistway.

The range of forces experienced by the transfer cask as a result of these drops are considerably less than the maximum allowed vertical deceleration of 82g for intact fuel. Therefore, no damage would occur to intact fuel and karwould remain <0.95. As discussed in section 5.2.1.4 above, this .

deceleration may cause damage to suspect / failed fuel pins. Therefore, a radiological release is

~

postulated for events that occur when the shield lid is not welded. The dose at the site boundary is ,

projected to be 0.003 rem, which is well below the EPA Protective Action Guide of I rem whole txxlv for the early phase of an event. The failed fuel would not present a criticality concern, as 4

discussed in 5.2.1.4 above, if fuel pellets were to leak from the failed fuel pins and accumulate at the bottom of the failed fuel cans.

- The calculated decelerations are less than the maximum allowable vertical deceleration for the

, basket pressure boundary of 124g. ' For drops that are postulated to occur after the shield and

- structural lids are welded to the basket, the basket pressure boundary serves as the confinement -

boundary and there would be no radiological release even from the suspect / failed fuel inside the basket because the basket pressure boundary would remain intact.

The safe load paths on the 93' elevation of the Fuel Building have been selected to minimize the

. possibility of an adverse effect of a dropped load on safety related equipment. The safety related equipment that could potentially be adversely affected by a load' drop is the Spent Fuel Pool,

, including the~ spent fuel racks. The safe load paths are sufliciently far from the Spent Fuel Pool to DS prevent a heavy load from dropping into the Spent Fuel Pool. _ Mechanical stops and electrical k - interlocks will prevent the Fuel Building overhead crane from moving over the Spent Fuel Pool. As

stmed above, loads are not lifted sufficiently far above the floor to allow a cask to tipover and roll

- into the Spent Fuel Pool. (Tipover events are discussed in section 5.2.2 below, but their occurrence Y

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999

- is only postulated in locations where the cask could fall back into a pit from, or into, which they are being lifted, orlowered).

5.2.1.5 Transfer Cask L' i d Assembly Drop onto a Basket Loaded with Fuel A transfer cask lid assembly drop is postulated to occur when the transfer cask is located at the Decontamination and Assembly Station and the cask lid assembly is being removed to weld the

' shield and structural lids to the basket ' The design safety factors load testing requirements, and l administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling equipment minimize the possibility of a cask lid assembly drop actually occurring. As described below, : cask lid assembly drop will not cause basket damage that results in a radiological release or spent fuel damage that results in a k,y of more than 0.95.

> The consequences of dropping the cask lid assembly on the basket loaded with fbel are bounded by the lifting yoke drop onto the basket (Section 5.2.1.3). A major difference would be that for this b event, the cask lid assembly weighs considerably less than the lifting yoke. .Also, the drop height p would be less than 28'. The consequences of the cask lid assembly were not specifically determined because the lifling yoke drop consequences are bounding.

5.2.1.6 Basket Structural Lid Drop onto Basket Loaded with Fuel .

A basket structural lid drop is postulated to occur when the transfer cask is located at the Decontamination and Assembly Station after the shield lid has been welded to the basket which is loaded with fuel. The design safety factors, load testing requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling equipment minimize the possibility,of a structural lid drop actually occurring. As described below, a structural lid drop will not cause basket damage that results in a radiological release or spent fuel damage that results in a

' k ,of more than 0.95.

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, Enclosure 2 to VPN-005-99 t L . LCA 237, Revision 3 Januarv 7.1999 !

The consequences of dropping the structural lid on the basket loaded with fuel are bounded by the i L . lifting yoke drop onto the basket (section 5.2.1.3). A major difference would be that for this event, L

the structural lid weighs considerably less than the lifting yoke. Also, the drop height would be less ;

than 28'. In addition, for a structural lid drop, the shield lid would be welded to the basket whc's the

. drop would occur. The shield lid would add considerable protection for the fuel. The consequences

'of the structural lid drop were not specifically determined because the lifting yoke drop

.. consequences are bounding.

l

, 5.2.1.7. Basket Lift Rigging Drop onto Basket Loaded with Fuel 1

' The basket lift rigging drop is postulated to occur when the basket loaded with fuel is being '

prepared for lowering from the transfer cask into the concrete cask in the Fuel Building bay. A ,

basket lift rigging drop would occur after the shield and structural lids have been welded to the l basket. .The design safety factors, load testing requirements, and administrative controls (i.e., ;

i '

procedures, training, maintenance, inspections) for the load handling equipment minimize the possibility of a basket lift rigging drop actually occurring. As described below, a basket lift rigging -

drop will not cause basket damage that results in a radiological release or spent fuel damage that !

. results in a kaof more than 0.95.

The consequences of a basket lift rigging drop are bounded by the lifting yoke drop onto the basket .:

shield lid (section 5.2.1.3). The basket lift rigging weighs considerably less than the lifting yoke and the drop height would be less than 28'. In addition, the shield lid would be welded in place and L the structural lid would be welded in place on top of the shield lid at the time that the basket lift

' rigging drop is postulated to occur. The structural lid provides an additional 3" of steel for

. protection of the fuel assemblies in addition to boti lids being welded in place, neither of which

_were considered in the lifting yoke drop. The consecuences of the basket lift rigging drop were not ,

L $ specifically determined because the lifting yoke drop consequences are bounding. l l

k 23 f ' .'

i

t l Enclosure 2 to VPN-005-99

' LCA 237, Revision 3 January 7; 1999 l.

5.2.1.8 Basket Drop into Concrete Cask l

The basket drop into the concrete cask is postulated to occur as the basket loaded with fuel is being l lowered from the transfer cask into the concrete cask in the Fuel Building bay. The basket drop

! would occur aner the shield and structural lids have been ;velded to the basket. The design safety g factors, load testing requirements, and administrative controls (i.e., procedures, training, i

maintenance, inspections) for the load handling equipment minimize the possibility of a basket drop actually occurring. As described below, a basket drop will not cause basket damage that results in a radiological release or spent fuel damage that results in a k,y of more than 0.95.

i The basket drop analysis considered a drop of the basket from the top of the transfer cask bottom

~ doors to the bottom of the concrete cask. An impact limiter will be installed in the Fuel Building,

'with its top surface at grade level (El. 45'), at the location of the basket transfer into the Concrete Cark. The impact limiter is designed to maintain basket hypothetical drop decelerations to less than the allowable values. Therefore, the basket drop would not result in a radiological release or an

-increase in k,.y of more than 0.95.

5.2.1.9 Concrete Cask Shield Ring or Lid Drop onto a Basket Loaded with Fuel The concrete cask shield ring or lid drop are postulated to occur aner the basket loaded with fuel has ben lowered from the transfer cask into the concrete cask in the Fuel Building bay and the transfer cask has been removed from the concrete cask. The drop would occur after the shield and structural lids have been welded to the basket. The design safety factors, load testing requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling E equipment minimize the possibility of this drop actually occurring. As described below, this drop will not cause basket damage that results in a radiological release or spent fuel damage that results

! . in a k,y of more than 0.95.

I.

j . The consequences of a concrete cask shield ring or lid drop are bounded by the lifting yoke drop onto the basket shield lid (section 5.2.1.3). The concrete cask shield ring and lid each weigh i 1 24

% i e- .my . - - .

Enclosure 2 to VPN-005 r

LCA 237, Revision 3 .

January 7.1999 ,-

l 1 l

considerably less than the lifting yoke and the drop height would be less than 28', In addition, the I

l shield lid would be welded in the basket and the structural lid would be welded in place on top of.

the shield lid at the time that the concrete cask shield ring or lid drop is postulated to occur. The !

- structural lid provides an additional 3" of steel for protection of the fuel assemblies in addition to

- both lids being welded in place, none of which were considered in the lifting yoke drop. The consequences of the concrete cask shield ring or lid drop were not specifically determined because .

- the lifting yoke drop consequences are bounding..

- 5.2.2 Tipovers 5.2.2.1 Transfer Cask Tipover with Basket Loaded with Fuel l This transfer cask tipover is postulated to occur at the top of the Cask Loading Pit while being withdrawn (after being loaded with spent fuel assemblies), and at the top of the Cask Wash Pit. The j

- tipover would be the result of a transfer cask drop while the transfer cask is partially over the Pit and I partially over the Fuel Building floor. The design safety factors, load testing requirements, and

- administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling equipment minimize the possibility of a transfer cask tipover actually occurring. As described l

below, a transfer cask tipover will not cause spent fuel or basket damage that results in a radiological release that exceeds regulatory limits or spent fuel damage that results in a k,y of more than 0.95. l 1

5.2.2.1'.1 Transfer Cask Tinover at the Cask Loading Pit A transfer cask tipover at the Cask Loading Pit was assumed to occur at the east wall of the Cask j Loading Pit where the transfer cask would be located after being withdrawn from the Cask Loading l' iPit.

l The transfer cask is assumed to tip sideways from the east wall and impact the top of an impact i limiter located on the west wall of the Cask Loading Pit. (The transfer cask is about 16' high and 25 ;

l o i i

i:

i

(

w - -r ,o 1 e r ~

l Enclosure 2 to VPN-005 L

. LCA 237, Revision 3 Japimrv 7.1999 L .

the Cask Loading Pit is about 12' across in the east-west direction.) For a worst case side impact, the transfer cask is assumed to tip such that its base remains at the top of the east wall and does not slide into the Cask Loading Pit which would create a glancing blow instead of a solid impact.

l (However, if the transfer cask slipped into the Cask Loading Pit, the results would be bounded by the drop into the Cask Loading pit discussed in section 5.2.1.4.1 above., j l

The analysis shows that with the use of a floor impact limiter positioned just west of the Cask Load )

l- Pit, the deceleration force on the fuel would be considerably less than the maximum allowed horizontal load of 44g for intact fuel. Therefore, no damage would occur to intact fuel. As !

discussed in section 5.2.1.4 above, this deceleration may cause damage to suspect / failed fuel pins.

Therefore, a radiological release is postulated with a dose at the site boundary of 0.003 rem, which j l is well below the EPA Protective Action Guide of I rem whole body for the early phase of an event.

l The value of kmwill remain less than 0.95 from the intact fuel and, as discussed in section 5.2.1.4 i above, the failed fuel would not represent a criticality concern if failed fuel pellets were to leak from the failed fuel pins and accumulate at the bottom of the failed fuel cans.

The analysis also shows that the strength of the concrete west wall of the Cask Loading Pit is j sufficient to withstand the transfer cask impact. j When the transfer cask is lifted from the Cask Loading Pit, there may be about 40 to 50 gallons of Spent Fuel Pool water on top of the shield lid (3" between the shield lid and the top of the basket) that could spill onto the Fuel Building floor if the transfer cask tipped over. (The water in the transfer cask / basket annulus would drain into the Cask Loading Pit through the transfer cask bottom doors.) This amount of water would cause minimal exposure to workers because the water would be only slightly contaminated. Most of the spilled water would be captured in floor drains. Clean l up of the remaining spilled water would be relatively straightforward.

l 26 i

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l

l l

Enclosure 2 to VPN-005-99 !

~

LCA 237, Revision 3 <

January 7.1999

]

l l

g 5.2.2.1.2 Transfer Cack Tinover at the Cack Wash Pit A transfer cask tipover at the Cask Wash Pit was assumed to occur at the west wall of the Cask Wash Pit where the transfer cask would be located prior to being lowered. This is no longer a fuel

, drop / failure concern since the transfer cask will be positioned in the Cask Wash Pit in support of

> loading an empty basket assembly or for temporary storage. A hypothetical transfer cask impact at -

this location would create structural damage but would not damage any plant equipment important to the safe storage of fuel.

5.2.2.1.3 Transfer Cack Tinover at the Fuel Building Hoistway )

A pair of guide beams will be installed across the top of the hoistway to prevent a transfer cask .i

' tipover. Therefore, a transfer cask tipover at the Fuel Building hoistway is not considered to be

)

credible. l 1

l L: 5.2.2.2 Transfer Cask Tipover during Concrete Cask Transfer Operations = I This transfer cask tipover is postulated to occur when the transfer cask is being placed on the -

concrete cask in the Fuel Building bay. After the transfer cask has been lowered until it is just a few N nes higher than the top of the concrete cask, the concrete cask will be moved undemeath the .

j i

transfer cask, and the transfer cask will be lowered the last few inches onto the top of the concrete

- cask. If a vettical drop occurred while a portion of the transfer cask was over the concrete cask, but l

! the transfer cask center of gravity was not over the concrete cask, the transfer cask could slide off l

the~ concrete cask and could tipover. The derign safety factors, load testing requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) for the load handling i equipment minimize the possibility of a transfer cask tipover actrally occurring. As described ;

, below, a transfer cask tipover will not cause spent fuel or basket damage that results in a 4

radiological release or spent fuel damage that results in a kmof more than 0.95.

l 1 l- j I

l r

F 27 l l

! j

, i L . . , . . .- - - . . - . .. -. .-

.. - ~ - . .- . - -- . - - . - -

)

Enclosure 2 to VPN-005-99 '

l LCA 237, Revision 3

( January 7.1999 l

The consequences of a transfer cask tipover during concrete cask transfer operations are bounded by I the transfer cask side drop in the Fuel Building hoistway (5.2.1.4.3). The transfer cask will be over I the impact limiter placed at the bottom of the Fuel Building hoistway. If the transfer cask is mispositioned on top of the concrete cask such that tipover were possible, its center of gravity l

would be over the impact limiter and if the transfer cask fell off the concrete cask, it would fall onto

! the impact limiter. The distance that the transfer cask could fall from the top of the concrete cask (about 17.5' high) to the top of the impact limiter is considerably less than the 43' drop used in the transfer cask drop for which the analysis showed that there would be no intact spent fuel or basket -

damage, no radiological release, and no increase in ke rrabove 0.95.

5.2.3 Mishandling Events 5.2.3.1 Cask Loading Pit or Spent Fuel Pool Liner Tear / Breach During Handlitig Operations ;

The Cask Loading Pit or Spent Fuel Pool liner tear / breach is postulated to occur by the drop of a l

. fuel assembly or heavy load such as the transfer cask. The design safety factors, load testmg i requirements, and administrative controls (i.e., procedures, training, maintenance, inspections) for I the fuel and load handling equipment minimize' the possibility of a liner tear / breach actually occurring. As described below, a liner tear / breach will not cause increased radiation levels in the Fuel Building as a result of water loss from the Cask Loading Pit or water loss from the Spent Fuel Pool.

i The water level in the Cask Loading Pit will be at the same level as the Spent Fuel Pool because the Spent Fuel Pool gate, which separates the Cask Loading Pit and Spent Fuel Pool, will be open for moving spent fuel from the Spent Fuel Pool to the Cask Loading Pit. The Spent Fuel Pool water ;

level is maintained 23' or more above the spent fuel stored in the Spent Fuel Pool per the Trojan Technical Specifications. The Trojan Technical Specification Bases state that 10' of water over the spent fuel provides an adequate heat sink for the fuel and shielding for personnel working in the area.

i I,

1 l 28 f

f-

, , - - - - . . - . - . . . - - - . . ~ .. = - - - - .- . - - - - .

t i l -

l ucosure 2 to VPN-005-99 ,

! LCA 237, Revision 3 January 7.1999 L ,

If the Cask L'oading Pit liner is tom or breached, the leakage from the tear or breach will be i

collected by tell-tale drains. The size of the drain lines from the Cask Loading Pit will restrict the leakage to 44 gpm. If the Spent Fuel Pool gate is closed and is not damaged by the event, the Cask

~

~ Loading Pit would drain to the minimum assumed height of water over the Transfer Cask, y hich is 10', in approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the initial level is near the Spent Fuel Pool level of greater tLm 23 feet above the fuel. This would be ample time to stop the leak by shutting the valves on the Cask Loading Pit tell-tale drain lines, patch / repair the liner, or place the shield lid on the basket (to prevent unacceptable radiation levels due to the loss of watcr shielding). In addition, water may be supplied from the various sources to makeup the water loss if more time is needed to stop the leakage. If the Spent Fuel Pool gate was also damaged such that water from the Spent Fuel Pool could flow into the Cask Loading Pit, the leak rate would be the same 44 epm, but the level decrease would be much slower due to the increased volume to be drained, i.e., the combined volume of the Cask Loading Pit"and Spent Fuel Pool. Once again, there would be ample time to secure the leak by shutting the valves on the Cask Loading Pit tell-tale drain lines. Water may be supplied from the various sources to makeup the water loss if more time is needed to stop the leakage. Ifleakage were to occur when the cask load pit level is reduced for placement or removal of the cask, either the basket will contain no fuel, or the shield lid will be in place to provide adequate shielding. Ample time for compensatory actions are available (greater than 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months ) to ensure fuel cooling.

If the Spent Fuel Pool liner is torn or breached, the leakage from the tear or breach will be collected by tell-tale drains. The size of the drain lines from the Spent Fuel Pool will restrict the leakage from a tear or breach of the Spent Fuel Pool liner to 42 gpm. This leakage is less than leakage from a Cask Loading Pit liner tear / breach and the Spent Fuel Pool is considerably larger than the Cask Loading Pit. Therefore, there will be more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to stop the leak by shutting the valves on the Spent Fuel Pool drain lines (with or without the gate closed between the Cask Loading Pit and

. Spent Fuel Pool)., As previously stated, water may be supplied from the various sources to makeup

- the water loss if more time is needed to stop the leakage.

L 29

\,

iK

. Enclosure 2 to VPN-005-99 LCA 237, Revision 3 L January 7.1999 i

The Trojan DSAR, Section 6.3.3, evaluates the effects of the loss of forced spent fuel cooling with concurrent spent fuel pool inventory loss. The evaluation credits only 10 feet of water being present r over the fuel (due to seismic failure of the gates and spent fuel cooling system piping). This evaluation demonstrates that there is ample time for operator action to provide make up water, from a variety of sources, to the spent fuel pool before level' decreases below 5 feet above the top of the fuel stored in the pool. :

5.2.3.2 Crane Mishandling Operation with Transfer Cask / Basket Resulting in Horizontal Impacts f :

A crane mishandling operation is postulated to occur while moving the transfer cask along the safe

' load path or when lowering the basket from the transfer cask into the concrete cask. .A crane mishandling operation is unlikely because of the administrative controls (procedures, inspections.

l maintenance, and testing) that are implemented. As described below, a crane mishandling operation !

may cause a radiological release but will not result in k,y of more than 0.95.

The analysis of a crane mishandling ' operation assumed a horizontal impact of 2' per second.

(Vertical impacts are bounded by the spectrum of drops addressed in sections 5.2.1.1 through l 5.2.1.9.) The impact corresponds to a deceleration of 17.5g. This deceleration is considerably less than the maximum allowable horizontal load of 44g for the intact fuel. Therefore, no intact fuel -

damage would occur and k,y would not increase. .

If the crane mishandling operation occurred prior to the shield lid being welded in place, then the ;

shield lid retainers will keep the shield lid in place on the basket and prevem any significant 1 quantity of potentially contaminated water from spilling out of the basket.

The Spent Fuel Pool is the only r.afety related equipment that could be potentially impacted by

' inadvertent horizontal movement of the transfer cask. However, this type ofimpact is not credible because the Fuel Building overhead crane's mechanical stops and electrical interlocks would l

[ prevent movement of the load over/into the Spent Fuel Pool. Therefore, safety-related plant equipment would not be adversely affected by a crane mishandling event.

e I

30 i l

l

i

't Enclosure 2 to VPN-005 5 - LCA 237, Revision 3 January 7.1999 j 5.2.3.3 Interference while Lowering Basket during Transfer Operations j' Interference while lowering the basket is postulated to occur while transferring the basket from the 4

transfer cask to the concrete cask in the Fuel Building bay. The interference would result f:om the j~ basket catching on the top edge of the concrete cask while being lowered which could be caused by misalignment of the transfer cask on top of the concrete cask or some sort of misalignment of the

[ basket lifting rig. The potential for interference is minimized because the concrete cask internal :

cavity is about 8" larger in diameter than the basket which allows additional horizontal clearance s

-the basket is being lowered into the concrete cask. In' addition, alignment holes are used to

? accurately position the transfer cask on top of the concrete cask prior to the basket lowering

operation. As described below, interference while lowering the basket will not cause basket damage

that would result in a radiological release or fuel damage.

j . The only force acting on the basket during lowering is gravity (Ig). Therefore, the worst case condition would be a load of Ig on the basket bottom or side ifit were to be completely supported by the interference. The stresses applied to the basket by the interference are bounded by those analyzed for the basket drop in the concrete cask described in section 5.2.1.8 above, which a

concluded that the basket accelerations were considerably less than the maximum allowed and -

[ would not result in a radiological release. Therefore, interference would not result in basket damage 4 that could cause a radiological release. . .

5.2.3.4' Misalignment of Transfer Cask Lifting Yoke i[

i Misalignment of the transfer cask lifting yoke ( i.e., less than full engagement of both transfer cask

-trunnions) is postulated to occur when the yoke is being placed on the transfer cask after fuel has y, _ been loaded into the transfer cask. The yoke is placed on the transfer cask, while the transfer cask is

! in the Cask Loading Pit in preparation for lifting the transfer cask out of the Cask Loading Pit. The yoke is also placed on the transfer cask at the Decontamination and Assembly Station after the

' shield and structural lids have been welded into the basket and the transfer cask is ready to be 31

1 1

L Enclosure 2 to VPN-005-99 L

LCA 237, Revision 3

'knuary 7.1999 i

I L {

o moved to the concrete cask. As described below, misaligning the transfer cask lifting yoke will not

- cause a drop of the transfer cask.

~ Misalignment of the transfer cask lifting yoke is highly unlikely because the design of the yoke l' ' ensures simultaneous full engagement of both transfer cask trunnions. The designed difference in ;

j diameter between the trunnion and cutout in the yoke arm into which the trunnion is placed will be about 0.2", which ensures that the yoke fits tightly on the trunnions with little chance for misalignment. The transfer cask' diameter will be about 0.5" less than the distance between the yoke arms, therefore, there is very little play between the yoke arms and the transfer cask which ensures i

that the yoke is not placed on the transfer cask in a misaligned position. Once the yoke is placed on the transfer cask trunnions, the yoke will not be able to spread apart and slip off the ends of the trunnions to a ' misaligned position because the trunnions are fitted with endcaps that are larger in diameter than the cutout in the yoke into whi :h the trunnions are placed. In addition, the yoke,

' trunnions, and transfer cask are checked for fit-up aner fabrication.

L

j. 5.2.4 - Onerational Errors 5.2.4.1 Opening the Transfer Cask Bottom Doors Prior to Attaching the Basket Lifting Rig l

! . Opening the transfer cask bottom doors is postulated to occur when the transfer cask has been placed on top of the concrete cask in the Fuel Building bay in preparation for transfer of the basket L (loaded with fuel) from the transfer cask to the concrete cask. Opening the transfer cask bottom e doors without first attaching the basket lifting rig to the basket will result in a drop of the basket into )

the concrete cask. Opening the transfer cask bottom doors prior to attaching the basket lihing rig is unlikely because procedural controls will minimize the possibility of occurrence of both of the two l.

l . independent actions that are required for opening the transfer cask bottom doors. As described below, this event is not considered credible.

L ,

Opening the transfer cask bottom doors will not occur because two separate and independent procedural errors are required to cause the bottom doors to be opened. The bottom doors are moved 32 !

I L

Enclosurc 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 by hydraulic actuators that are attached to the sides of the transfer cask. Operators must take deliberate action to actuate the hydraulica to open the bottom doors. This action would not be procedurally allowed until the basket lifting rig is attached to the basket. In addition, the transfer cask bottom doors are prevented from being inadvertently withdrawn from their closed position by locking bolts (2 per door,1 on either side). Operators must take deliberate action to remove the locking bolts to open the bottom doors. This action also would not be procedurally allowed until the basket lifting rig is attached to the basket. Therefore, opening the transfer cask bottom doors before the basket lifting rig is at' ched is not considered a credible event. Even if a basket drop were to occur, as described in section 5.2.1.8 above, the accelerations are less than the maximum allowable and there would be no radiological release or increase in km.

5.2.4.2 Closing the Transfer Cask Bottom Doors while Lowering the Basket Closing the transfer cask bottom doors on the basket is postulated to occur when the transfer cask has been placed on top of the concrete cask in the Fuel Building bay and the basket (loaded with fuel) is being lowered from the transfer cask into the concrete cask. Closing the transfer cask bottom doors could damage the basket. As described below, closing the transfer cask bottom doors on the basket is not considered a credible event.

Closing the transfer cask bottom doors on the basket is highly unlikely because two separate and independent operator actions or spurious actuations are required to cause the bottom doors to be closed while the basket is being lowered into the concrete cask. After the bottom doors are opened by the hydraulic actuators, the hydraulic pump (source of hydraulic motive force) will be stopped, and the hydraulic pump will be isolated fmm the hydraulic actuators by repositioning three way valves. Operators would need to reposition or misposition a three way valve and inadvertently start the hydraulic pump to close the bottom doors. Neither of these actions will be procedurally allowed until the basket is lowered completely into the concrete cask. Alternately, a spurious actuation of the hydraulic pump concurrent with the three way valve being physically repositioned or mispositioned is also unlikely. Therefore, closing the transfer cask bottoms doors while the basket is being lowered into the concrete cask is not considered a credible event.

33

j[ . '

,' s Enclosure 2 to VPN-005-99 LCA 237, Revision 3

. January 7.1999 3

5.2.4.3 i Brittle Fracture of the Transfer Cask

' Brittle fracture of the transfer cask will not occur because a service temperature of minus 3 *F.has l been established for the load bearing components of the transfer cask and handling operations will L be performed inside the temperature controlled Fuel Building.

Charpy v-notch energies have been determined for the transfer cask inner and outer shells, shield i doors and rails, cask trunnions, lifting yoke, and lift yoke pins using a service temperature of minus l 3 F, which is the lowest recorded temperature for Portland, Oregon. The fabrication specifications ' l require that the results of Charpy v-notch tests for these components are equal to or greater than the specified energies.-

" 5.2.4.4 Boron Dilution of Cask Leading Pit, Basket, or Spent Fuel Pool A boron dilution event is not likely because borated water will generally be used for operations involving the Cask Loading Pit and/or Spent Fuel Pool. The water may be filtered to reduce radioactivity / contamination levels or to improve water clarity, but the filtration equipment will be

operated to not reduce the boron concentration of the water being used to flush the basket / transfer

. cask annuhis. Limited quantities ofdemineralized water will be used to clean components as they

. are removed from the borated water in the Spent Fuel Pool or the Cask Load Pit, Spent Fuel Pool boron concentration is monitored, as required, by Technical Specification 3.1.2.

' As described above in section 5.2.3.1, a tear / breach of the Cask Loading Pit or Spent Fuel Pool liner although unlikely, could occur as the result of a load drop. One of the actions in response to a large b . leak would be addition of water to the Cask Loading Pit or Spent Fuel Pool from various L non-borated water sources. Addition of non-borated water would result in dilution of the Cask

~

Loading Pit or Spent Fuel Pool. The criticality analysis for a basket loaded with spent fuel shows

' that km is less than 0.95 even in non-borated water. Similarly, the Trojat. Technical Specification L .' Bases (B 3.1.2) state that the ka in the Spent Fuel Pool racks will be less than 0.95 even in

! < Lnon-borated water. Therefore, if a boron dilution did occur as a result of adding non-borated 34 i

b l

V w .w . . -, ,=- . - -- . --. --

1 a

a

1 Enclosure 2 to VPN-005-99

% LCA 237, Revision 3

{ ' January 7.1999 - l

. makeup water to the Cask Loading Pit or Spent Fuel Pool, the spent fuel in the basket and Spent I

Fuel Pool racks would remain suberitical. The PWR basket wet-case criticality analysis takes credit ,

for 75% of the installed BORAL plates.

i i

5.2.4.5; Spread of Transfer Cask and/or Basket Contamination 1

Spread of contamination is not likely because the surface contamination levels of the basket and concrete cask will be controlled and measured. The need for decontamination will be evaluated if loose surface contamination levels are above 104 Ci/cm2 -y or 104 pCi/cm2 a. This control

- ensures that even if the small amount ofcontamination that could be affixed to the basket became

' loose and was released,'the resulting dose would be negligible (0.0024 rem at 100 meters). The loose surface contamination level is determined by taking swipes'on the exterior of the basket, the

' internal surface of the transfer cask (which is representative of the contamination level of the inaccessible basket external surface), and the concrete cask external surface.

}

In addition,'the NRC's Dry Storage Action Plan identifies weeping as a potential source of contamination from casks' with bare metals, especially stainless steels, and states that weeping may -

be alleviated by coating the metal surfaces of the cask.' Normally, contamination caused by weeping

j. ' would occur only if the transfer cask and basket were submerged in the Cask Loading Pit over a

, , period of days. Since the loading operation will normally only require the transfer cask and basket

, to be submerged in the Cask Loading Pit for a short period of time, and the water in the Cask

. Loading Pit will not contain significant quantities ofloose contamination, absorption of p . contamination by the transfer cask and basket is not expected. However, as a precaution, the outer

, and inner surfaces of the transfer cask and the extemal surface of the basket will be painted with a L hard, smooth coating to prevent absorption of contamination by the transfer cask and basket which should further preclude the potential for weeping.

1 35 x

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F Enclosure 2 to VPN-005-99 o LCA 237, Revision 31 l January 7.1999 i

g '5.2.4.6 Inadvertent Basket Over Pressurization during Draining / Drying Operations l y

' Inadvertent basket over pressurization during draining and drying operations is not likely because i

multiple equipment failures and a procedural error are required. Until the shield lid is welded to the

basket, pressure from the basket could relieve through the shield lid vent penetration, shield lid drain penetration, or the gap between the shield lid and basket. After the shield lid is welded the

[ shield lid ' vent penetration and drain penetration could relieve pressure except during pressure l testing, basket drying (with pressurized inert gas), and helium filling. Pressure test procedures will j l

closely control the pressure in the basket and the test rig will have relief valves to protect the basket

[ from overpressurization. The relief valve set points are slightly higher than the minimum test c pressure (by approximately 2-5 psig) and well below the basket accident storage condition I maximum pressure described in ISFSI SAR Section 8.2.6.2. The inert gas supply used to dry the

!' . basket and the helium supply used to fill the basket after vacuum drying will have regulators that are .

! set at or below the test pressure of approximately 15 psig. In addition, pressure relief valves will be f

installed downstream of the regulator and upstream of the basket inlet penetration. The regulator and the relief valves all must fail in order to overpressurize the basket. In addition, procedures will h require that an operator monitor a pressure gauge and manually isolate the pressure source or relieve

' pressure if the pressure reaches approximately 15 psig (less the gauge tolerance). Therefore, over pressurization of the basket is not likely because multiple failures / errors must occur.

[ 5.2.5 Supoort System Malfunctions 5.2.5.1. Welding System Failures Welding system failures are postulated to occur while the shield and structural lids are being welded

to the basket at the Decontamination and Assembly Station. Welding equipment failures would include loss of electrical power or component failures that prevent completion of the welding process. As described below, delays in the basket draining / drying process caused by welding equipment failures will not result in fuel cladding temperatures that exceed short term temperature limits.

36

Enclosure 2 to VPN-005-99 LCA .237, Revision 3 January 7,1999 A calculation shows that at least 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months of heat-up are required for the water in the basket to reach boiling. This provides ample time to restore electrical power or repair / replace failed components.

In the unlikely event that electrical power cannot be restored or failed components cannot be repaired / replaced prior to steam being generated, procedures will specify appropriate methods and the equipment needed to be available to cool the basket or vent the steam generated by the basket until the power is restored or the tailed components are repaired / replaced. If the water inside the basket boils, the fuel cladding will not exceed any short term temperature limits nor experience significant strain that could cause long term cladding degradation because the temperature of boiling water is well below the normal fuel cledding operating temperature.

5.2.5.2 Basket Draindown/ Vacuum Drying System Failures Basket draindown/ vacuum drying system failures are postulated to occur while water is being pumped or blown from the basket er while the basket gases are being evacuated. Basket draindown/ vacuum drying system failures would include loss of electrical power, component failures that prevent completion of the draindown/ vacuum drying process, or rupture of a pressurized line. The bas'ket draindown/ vacuum drying systems are commercial grade components which means that failures may occur although the potential for failure would be unlikely. Passive failure of a pressurized line is unlikely because the operating pressures for the basket are very low, (i.e, atmospheric to about 15 psig). As described below, delays in the basket draining /drymg process caused by basket draindov,n/ vacuum drying system failures will not result in fuel cladding temperatures that exceed short term temperature limits and passive failure of a pressurized line will not result in a radiological release that exceeds regulatory limits.

5.2.5.2.1 Loss of Electrical Power or Comr>onent Failures l

l As stated in section 5.2.5.1 above, the loss of electrical power or component failures that delay j completion of the draindown/ drying process do not result in fuel cladding temperatures that exceed I short term limits. Procedures will specify appropriate methods and the equipment needed to be available to vent the steam generated by the basket (e.g., back into the Spent Fuel Pool through the l l

37 l 1

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Enclosure 2 to VPN-005-99 LCA 237, Revision 3 J;taugal 1999 line that will be used to return the basket water to the Spent Fuel Pool) until the power is restored or the failed components are repaired / replaced.

Once water is removed from the basket and the vacuum drying process begins, fuel cladding temperatures are higher because heat transfer from the fuel to the basket is less in a vacuum.

IIowever, the fuel cladding temperature in a vacuum will not exceed short term fuel cladding temperature limits. In the event of high temperatures, the basket could be backfilled with helium to cool the fuel. The helium atmosphere would increase heat transfer from the fuel to the basket and lower the fuel cladding temperatures as well as minimize fuel cladding contact with an air (oxidizing) atmosphere.

5.2.5.2.2 Rupture ofa Prenurized Ling Two cases were considered for a rupture of a pressurized line: the discharge line on the draindown pump ruptures while the basket is being pumped dry which results in a spill of contaminated water in the Fuel Building, and the inert gas supply line to the basket ruptures while filling the basket which would release airbome radioactivity to the Fuel Building atmosphere.

For the rupture of the discharge line on the draindown purnp, the analysis assumes that the basket is filled with water and there is no operator action which resuhs in the entire water volume of the basket, about 1572 gallons, beint, pumped onto the Fuel Building floor at the 93' elevation. With an assumed Spent Fuel Pool activity of 4.59 E-5 pCi/mi gross p,8.78 E-5 pCi/ml gross y, and 1.56 E-

.2 pCi/m! tritium, the radioactivity released in the Fuel Building would be approximately 2.73 E 4 Ci gross p,5.22 E-4 Ci gross y, and 9.3 E-2 Ci tritium which would result in a negligible dose at the site boundary due to the lack of airborne radioactivity. The water would cause minimal exposure to workers. Mest of the water would be captured by floor draias. Clean up of the y remaining spilled water would be a relatively straightforward and not hazardous.

For the rupture of the inert gas supply line, the analysis conservatively assumes that the basket is at

25 psig and that the effective surface contamination levels are 12 pCi/ square em with a normalized 38 l

i V

l Enclosure 2 to VPN-005-99

- LCA 237, Revision 3 Januarv 7.1999 _

i decay fractional activity for each radionuclide as of January 1,1998 Fineen percent of this coniamination is assuiaed to bc released as airborne particulate during the depressurization transient. The release to the Fuel Building atmosphere would result in dose at the site boundary of

' less than 1/1000th of the whole body PAG limit, if Auxiliary Building / Fuel Building ventilation is running, and less than 1/3rd of the whole body PAG limit if the Auxiliary Building / Fuel Building

. ventilation is not running. These doses are well below EPA Protective Action Guide of I rem whole body for the early phase of an event.

5.2.5.3 Air Pad System Failures Air pad system failures are postulated to occur after the basket has been loaded in the concrete cask,

- the concrete cask cover lid has been installed, and the concrete cask has been lifted by the air pad system for movement to the concrete storage pad. Air pad system failures would be malfunctions of the compressor, the air hoses that feed the air pads, or the air pads themselves that result in the loss of air pressure to the air pads. As described below, an air pad system failure will not cause basket damage that results in a radiological release or spent fuel damage that results in a kg of more than 0.95.

Failure of the air compressor, the air hoses that feed the air pads, or the air pads themselves will

- cause loss of air pressure that will result in the concrete cask being lowered to the concrete pad or surface over which the concrete cask is being moved. If a catastrophic failure of one of the four air

. pads undemeath the concrete cask occurs, then the concrete cask would be lowered rapidly on the

. corner being supported by the failed air pad The air manifold would automatically feed more air to

' the failed pad which would partially cushion the impact. If the failure is in the compressor or the air hoses that feed the air pads, then air pressure loss in the air pads would be slower and the concrete cask would be lowered to the concrete pad s!ower than in the case of the catastrophic failure of one of the air pads. In any event, the concrete cask will be lowered to the concrete pad or surface over which the concrete cask is being raoved from the lift height of the air pads, which is about 3" An analysis of a tipover of the concrete cask onto the concrete storage pad shows that the concrete cask can sustain a side drop (most limiting) from an equivalent drop height of approximately 59" without 39 e

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Enclosure 2 to VPN-005-99i !

.E 4 j /LCA 237, Revisiori 3);

"Jammev' 731999 1 n.. ,

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f unacceptable damage to the basket 'or intact fuel. Therefore, an air pad system failure would not ;

result in a radiological release or intact spent fuel damage. Nor would a k,y of more than 0.95 as i

discribeil in Section 5.2.1.4 ' result.

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3.2.6 Naiural Phenomena :

1 f 5.2.'6.1r Natural Phenoraena: i p

b - The spectrum of natural phenomena include tornados, floode, tsunami, and earthquakes.

L5.2.6.1.1 Tornadog .

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/ , J Procedures will be in place to preclude spent fuel loading and handling during high' winds and '

' unstable atmospheric conditions that could lead to a tornado threat. , ;

i

$5.2hl.2' Floods and Tsunami 7

n n : Spent fuel loading an'd handling will take place inside the Fuel Building at elevations above credible

[ external flood (internal flooding is'not postulated due to the limited sources of water) and tsunami i

{ < levels. ,

e.

w . 5.2.6.1.3 Earthounes' W '

. Earthquakes will have a negligible effect on the transfer cask, basket, and concrete cask individually

because of their rigid design. Therefore, an earthquake was only postulated to occur and cause 74

",, toppling of the transfer cask when the transfer cask is stacked on top of the concrete cask in the Fuel x e

.' Building bay while preparing to lower the basket loaded with fuel from the transfer cask to the e , concreie cask.t For the transfer cesk to topple, the earthquake would need to occur during the short ;

o o %.

J time that the lifting yoke is detached from the transfer cask and the basket lifting rig has not been c .4 fattached to the basket.7 oppling T would not occur while the basket or transfer cesk were attached to l, .. /4 '.

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L Janur.rv' 7.1949

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- the lifting equipment because the design safety factors would support the load. Toppihg while the

basket and transfer cask are both not attached to lining equipment is highly unlikely because the probability is extremely small that an earthquake will occur during the short period of time that is require'd to shin the liftmg equipment from the transfer cask to the basket. As described below, an earthquake will not cause basket damage that results in a radiological release or spent fuel damage l~ that results in a kmof more than 0.95.

An analysis was perfonned assuming that a Seismic Margin Earthquake occurred while the transfer

- cask (basket loaded with fuel inside) is stacked on top of a concrete cask in the Fuel Building bay,

[ the lifting yoke is detached from the transfer cask, and the basket lining rig has not been attached to l the basket. The calcuhtion concluded that some small amount of sliding of the transfer cask l across the top of the concrete cask may occur, but this would not be sufficient for the transfer cask

. center of gravity to move to the edge of the~ concrete cask. In addition, the rocking motion imparted by the earthquake would be much less thanwould be needed to overturn the transfer cask or transfer

. cask / concrete cask combination. Therefore, the transfer cask would remain on top of the concrete L cask during a Seismic Margin Earthquake, and no basket damage or spent fuel damage would occur. >

Criticality for damaged fuel within the basket is not a concern as described in Section 5.2.1.4.

5.3 OPERATIONAL CONTROLS The operational controls that are implemented to ensure that the assumptions of the safety evaluation and supporting analyses are satisfied are summarized below.

5.3.1 Handlino Height for the Transfer Cask The handling height of the' transfer cask (the height at which the transfer cask will be lifted above the floor or impact limiter on the 93 A. elevation) during movement along the designated safe load path will be limited to 4" to 15", depending on the location along the safe load path. A cask

. handling height on the order of 15 inches is required only where it is necessary to liR the cask onto or off of such items as the cask impact limiter, or load distribution assemblies. To ensure that the 41

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i Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 l

floor over which the transfer cask is moved has sufficient strength to withstand a transfer cask drop in the unlikely event that a drop cccurs, various components will be positioned along the safe load I path. Impact limiters will be placed in the Cask Loading Pit and Fuel Building hoistway. An

. impact limiter will also be placed at the base of the transfer cask or positioned at certain designated locations on the 93 ft. Fuel Building floor slab when movements of a fuel loaded transfer cask are made between the Cask Load Pit and the Decontamination and Assembly Station. A floor slab load distribution system will be positioned on the 93 ft. elevation of the Fuel Building just south of the hoistway. An impact limiter will also be placed on the 93 ft. Fuel . Building slabjust west of the Cask Loading Pit and transfer cask guide beams will be positioned across the Fuel Building ,

hoistway where tipover of the transfer cask was analyzed. These items will ensure the structural I integrity of the floors and the transfer cask in the unlikely event of a drop or tipover.

5.3.2 Basket Draindown Time l

The basket draindown time is administratively limited as described in ISFSI SAR Section 5.1.1.2 to ensure that the water inside the basket does not boil prior to removal of the water from the basket.

The draindown time starts when the top of the transfer cask is lifted from the water in the Cask Loading Pit and ends when the purge gas has blown the water out of the basket. This ensures that steam does not form which could potentially pressurize the basket above the pressure test pressure or could result in a radiological release (via steam escaping from the basket) into the Fuel Building.

Keeping water from boiling in the basket also ensures that the fuel cladding temperature remains well below the short term limits and does not cause fuel cladding strain which could potentially cause long term fuel cladding degradation. (Note that even without water, i.e., in a vacuum, an analysis shows that fuel cladding temperatures will not reach the short term temperature limits.)

Procedures will specify appropriate methods and the equipment needed to be available in the event that the basket draindown is not completed within the calculated time to boil.

42 i

. . . - _ . - . . _ - - ~ _ _ _ _ _ . . _ _ _ ._ _ _ _ . - . __

Enclosure 2 to VPN-005-99' LCA 237, Revision 3.

January 7.1999 '

~5.3.3 Handling Temnerature of the Transfer Cask The transfer cask will not be handled / lifted when ambient air temperatures in the transfer cask handling area are less than -3 *F. This requirement is satisfied by hand. ling operations being performed in the temperature controlled Fuel Building. This temperature limit minimizes the

possibility that the transfer cask components will experience a brittle fracture during handling / lifting operations.

5.3.4 Air Pad Installation Time The air pad system that is used to lift and move the concrete cask may inhibit natural circulation air flow through the concrete cask, which would prevent decay heat removal from the basket and cause the spent fuel cladding temperature to rise. Calculations show that blockage of the air inlets will not result in fuel cladding temperature limits that exceed the short term temperature limit.~ However, the -

amount of time that the air pad system can be inserted into the air inlet openings will be limited to' 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , which corresponds to the normal surveillance interval for measuring the concrete cask air outlet temperature.

-5.3.5 Concurrent Decommissioning Activities The postulated off-normal events and accidents in the safety evaluation do not consider interactions with concurrent decommissioning activities. Similarly, the postulated events in the Decommissioning Plan do not consider interactions with spent fuel loading and handling activities.

~ Therefore, a procedure will be written that specifies the spent fuel loading and handling activities and decommissioning activities that will not be performed concurrently because the activities could potentially interact with each other. This administrative control will ensure that concurrent

_ activities will not increase the probability or consequences of a postulated event or accident.

43

l. ._

~

' Enclosure 2 to VPN-005-99 LCA 237, Revision 3 -

Januarv 7.1999 1,5.3.6 Auxiliary / Fuel Buildino Ventilation

Auxiliary / Fuel Building ventilation will be in operation during spent fuel loading and handling inside the Fuel Building to minimize the amounts of airbome particulate radioactivity released from the Fuel Building to the environment. Procedures will specify actions to be taken in the event that

- the Auxiliary / Fuel Building Ventilation is not in operation during spent fuel loading and handling operations.

'5.3.7 Recovery from Off-Normal Events and Accidents Procedures will be written that specify appropriate methods and equipment that will be needed to recover from credible off-normal events and accidents. As a minimum, procedures will address recovery from a 1) fuel assembly drop,2) transfer cask / basket drops when the basket is loaded, 3) transfer cask tipovers when the basket is loaded,4) failures of pressurized lines connected to the basket loaded with fuel, and 5) spent fuel pool / pit liner tears / cracks.

6. SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION In accordance with the requirements of 10 CFR 50.92(c), implementation of the requested license amendment, spent fuel loading and handling in the Fuel Building, is analyzed using the following

-standards and found: 1) not to involu a significant increase in the probability of an accident previously evaluated; 2) not to invalys a significant increase in the consequences of an accident

- previously evaluated; 3) to create the possibility of a new or different kind of accident from any accident previously evaluated; and '4) not to involve a significant reduction in a margin of safety.

l. The requested license amendment does not involve a significant increase in the probability of an accident previously evaluated.

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j Enclosure 2 to VPN-005-99 LCA 237 Revision 3 <

j January 7.1999 With the issuance of a Possession Only License, the number of potential accidents was reduced

! to those types of accidents associated with the storage ofirradiated fuel and radioactive waste 1

i storage and handling. Additional events were postulated for decommissioning activities due to f the difference in the types of activities that were to be performed. The postulated accidents described in the Defueled Safety Analysis Report (DSAR) are generally classified as: 1)

, , radioactive release from a subsystem or component,2) fuel handling accident, and 3) loss of j spent fuel demy heat removal capability. The postulated events described in the

' Decommissioning Plan are grouped as: 1) decontamination, dismantlement, and materials handling events, 2) loss of support systems (offsite power, cooling water, and compressed air),

l

3) fire and explosions, and 4) external events (earthquake, external flooding, tornadoes, extreme

[ wmds, volcanoes, lightning, toxic chemical release). These types of accidents are discussed

below.

Radioactive release from a subsystem or component involves failure of a radioactive waste gas

[ decay tank (WGDT) or failure of a chemical and volume control systen; holdup tank (HUT).

3 For the failure of a WGDT, the radioactive contents were assumed to be principally noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and

[ trace quantities of halogens. For the failure of a HUT, the assumptions were full power

j. operation with 1-percent failed fuel,40 weeks since power operation, and 60,000 gallons of

[ 120 F liquid released over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period. However, the WGDTs and HUTS are no longer

active and have been drained or removed. Therefore, spent fuel loading and handling activities f- 'cannot increase the probability of occurrence of a failure of a WGDT or HUT.

3 The fuel handling accident involves a stuck or dropped fuel assembly which results in breaking g the cladding of the fuel rods in one assembly and releasing the gaseous fission products. Spent l fuel handling and loading will move the spent fuel assemblies, one by one, from the Spent Fuel Pool to the baskets that will be located in the Cask Loading Pit. The fuel handling equipment will be the same as before with the exception of special tools that will be used to manipulate the

. failed fuel. These special tools will be .similar in size and weight to other tools used for underwater manipulation, and therefore, would not present a new hazard. In addition, the same 45

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Enclosure 2 to VPN-005-99 I h : LCA 237, Revision 3 ,

. January 7.1999 L

' administrative controls and physical limitations imposed on any fuel handling operation will be l

L lused for spent fuel loading and handling. Thus, there is no increase in the probability of l l ' occurrence of a fuel handling accident over what would be expected for any routine fuel

- handling operation.

The loss of spent fuel decay heat removal capability involves the loss of forced spent fuel

cooling with and without concurrent Spent Fuel Pool inventory loss. The only requirement to L assure adequate decay heat removal capability for the spent fuel is to maintain the water level in

the Spent Fuel Pool so that the spent fuel assemblies remain covered (i.e., the capability to

. makeup water to the Spent Fuel Pool must be available when required). The potential events which could result in a loss of spent fuel decay heat removal capability include extemal events L (explosions, toxic chemicals, fires, ship collision with intake structure, oil or corrosive liquid

. spills in the river, cooling tower collapse, seismic events, severe meteorological events), and !

internal evente inchiding Spent Fuel Pool makeup water system malfunctions (multiple water L systems, electrical power, instrument air). Spent fuel loading and handling will not require the

. use of explosive materials (the gases used for electric arc welding are inert), toxic chemicals, or l

4 flammable materials (routine use of plastic sheeting or absorbent materials for contamination control is' not considered significantly hazardous). The probability of other extemal events (e.g.,

cooling tower collapse) would be unaffected by the spent fuel loading and handling activities I inside the Fuel Building. Spent fuel loading and handling activities will not directly interface r ,

! with the Spent Fuel Pool makeup water systems, therefore, could not afTect their probability of failure. (The Cask Loading Pit will be filled with borated water that will be cooled by the cooling systems, but use of this water in the Cask Loading Pit would not increase the failure probability of the Spent Fuel Pool or makeup water systems.) As described in the safety evaluation above, the safe load path and handling height limitations will ensure that a load drop ,

h does not adversely affect the Spent Fuel Pool or the makeup water systems. Therefore, there is F, no significant increase in the probability of a loss of spent fuel decay heat removal capability.

4

The events postulated in the Decommissioning Plan are similar to the DSAR with the exception of the decontamination, dismantlement, and materials handling events. Decontamination events ll

\

r 46 4:'

Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 involve gross liquid leakage from in-situ decontamination equipment (e.g., tanks), or accidental spraying ofliquids containing concentrated contamination. Dismantlement events involve segmentation of components and structures, or removal of concrete by rock splitting, explosives, or electric and/or pneumatic hammers. Dismantlement events potentially result in airborne contamination. Material handling events involve the dropping of contaminated components, concrete rubble, or filters or packages of particulate materials. As described in the operational controls above, administrative controls will be implemented to ensure that spent fuel loading /

and handling activities and decommissioning activities will not be performed concurrently if they interact with each other and could increase the probability or consequences of a postulated event or accident. Therefore, the probability of decontamination, dismantlement, and materials handling events would not be significantly increased.

Based on the above, the spent fuel loading and handling activities would not significantly increase the probability of any accident previously evaluated.

2. The requested license amendment does not involve a significant increase in the consequences of an accident previously evaluated.

The accidents described in the DSAR are generally classified as: 1) radioactive release from a I subsystem or component,2) fuel handling accident, and 3) loss of spent fuel decay heat removal capability. The events described in the Decommissioning Plan are grouped as:

1) decontamination, dismantlement, and materials handling events,2) loss of support systems (offsite power, cooling water, and compressed air),3) fire and explosions, and 4) external events

. (earthquake, external flooding, tornadoes, extreme winds, volcanoes, lightning, toxic chemical release).

As described above, the failure of a WGDT and HUT are no longer credible since these tanks have been deactivated or removed. Therefore, the consequences of a failure of a WGDT or HUT cannot significantly increase as a result of spent fuel loading and handling.

L 47 s

1 l

Enclosure 2 to VPN-005-99 ' '

LCA 237, Revision 3 -

January 7.1999 '

As discussed in the safety evaluation, if a fuel assembly was dropped while loading a basket in the Cask Loading Pit, then only I fuel assembly could be damaged. The previous analysis de' scribed in the DSA.R postulated the same results. Therefore, the consequences of a fuel assembly drop while loading a basket in the Cask Loading Pit would be the same as the l

-consequences of the analysis described in the DSAR. Therefore, the consequences of a fuel assembly drop while loading a basket in the Cask Loading Pit are not significantly increased as a result of spent fuel loading and handling.

There are no credible adverse consequences of the loss of spent fuel decay heat removal because the DSAR demonstrates that adequate time is available to establish a source of makeup water to

. the Spent Fuel Pool such that uncovering the fuel and an actual loss of spent fuel cooling is not credible. As described by the safety evaluation above, the postulated events that could affect the Spent Fuel Pool (liner tear / breach and heavy load drop) do not have a significant adverse effect.

In addition, establishment of the makeup water path and recovery of spent fuel cooling would l i

not be affected because postulated off-normal events and accidents would not affect the j capability to provide makeup water to the Spent Fuel Pool by various water sources. Basket j draindown events (loss of power, equipment failure) do not involve cladding temperatures above i short-term limits or result in radiological releases that exceed regulatory limits. Therefore, spent fuel loading and handling cannot significantly increase the consequences of the loss of spent fuel decay heat removal.

The events postulated in the Decommissioning Plan that are different from the DSAR are decontamination, dismantlement, and materials handling events. As described in the operational controls above, administrative controls will be implemented to ensure that spent fuel loading and handling activities and decommissioning activities will not be perfonned concurrently if

.they interact with each other and could increase the probability or consequences of a postulated l

> event or accident. Therefore, the consequences of decontamination, dismantlement, and I

materials handling events will not be significantly increased. )

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C y Enclosure 2 to VPN-005-99!

O LCA 237, Revision 3

? January 7.1999' Based on the above, the spent fuel lo'ading and handling activities do not involve a significant LJ . increase in the consequences of an accident previously evaluated.

h e

. 3. The requested license amendment does create the possibility of a new or different kind of

' accident from any accident previously evaluated.

1 The accidents described in the DSAR are generally classified as: 1) radioactive release from a

' subsystem or component,2) fuel handling accident, and 3) loss of spent fuel decay heat removal

>f : capability. The events described in the Decommissioning Plan are grouped as:

1) decontamination, dismantlement, and materials handling events,2) loss of support systems

. (offsite power, cooling water, and compressed air), 3) fire and explosions, and 4) cxternal events (earthquake, external flooding, tornadoes, extreme winds, volcanoes, lightning, toxic chemical release).-

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l DAs described in the safety evaluation of the proposed spent fuel loading and handling activities, I

only three types of off-normal events / accidents were determined to have radiological y consequences; 'a fuel assembly drop into a basket loaded with spent fuel, a transfer cask drop or mishandling event prior to the shield and structural lids being welded in the basket, and a basket i

' draindown/ vacuum drying system failure (passive failure of a pressurized line).

The postulated fuel assembly drop into a basket loaded with spent fuel is considered the same ;

- type or kind of event as the previously analyzed fuel handling accident, mainly because the :

(initiator for this postulated event is the same, i.e.,'a (non-specified) failure of the fuel handling

, iequipment or the Fuel 1%ndling Bridge Crane. The previous evaluation considered a dropped Lfuel assembly in Spent Fud Pool..: During spent fuel loading and handling, a fuel assembly may i T be dropped in the Spent Fuel Pool or the Cask Loading Pit. As the Cask Loading Pit is similar

- in construction to the Spent Fuel Pool and the Cask Loading Pit will be flooded with borated

' water of the same concentration as the Spent Fuel Pool, the differences between the two events are negligible and the two . events may be considered the same type or kind of event. Therefore, j {the fuel assembl'y drop is not a new or different type or kind of accident.

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- Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 721999 The postulated transfer cask drop or mishandling event prior to the shield and structural lids

.being welded in the basket is similar to a fuel handling accident. However, the fuel handling -

accident only considers dropping one fuel assembly because routine fuel handling operations only involve moving one fuel assembly at a time. In addition, normal fuel handling involves movement of the single fuel assembly underwater in the Spent Fuel Pool, transfer canal, etc. In the case of the transfer cask / basket, up to 24 fuel assemblies will be moved at one time, and for

. the portion of their movement where a radiological release is postulated, will not be within the

!- confines of the Spent Fuel Pool or transfer caual. The postulated transfer cask drop or mishandling events are considered to be a different kind of accident than any previously evaluated. As stated in the safety evahiation, the consequences of transfer cask drop or mishandling event prior to the shield and structural lids being welded in the basket are about 0.003 rem whole body at the site boundary. This dose is well below EPA Protective Action

. Guides of I rem whole body for the early phase of an event. Therefore, the consequences of drops or mishandling events prior to the shield and structural lids being welded in the basket would not represent a significant hazard to public health and safety.

The postulated passive failure of a pressurized line on the basket during the vacuum drying

! process is similar to events described in the Decommissioning Plan. The Decommissioning Plan considers airbome release of radioactive materials as a result of cutting, dropping, demolishing, etc., plant subsystems and components during decommissioning activities. The postulated passive failure of a pressurized line on the basket constitutes a different kind of L accident because it involves the airbome release of radioactive material by a different initiation process than previously evaluated. As described in the safety evaluation, the passive failure of a b pressurized line connected to a basket loaded with spent fuel results in a whole body dose at the site area boundary ofless than the 1 rem EPA Protective Action Guide for the early phase of an event. In a'ddition, the Auxiliary / Fuel Building ventilation will be operating during spent fuel l handling evolutions as a precaution (see operational controls above). Therefore, the consequences of a passive failure of a pressurized line would not represent a significant hazard to public health and safety.-

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l Enclosure 2 to VPN-005-99 LCA 237, Revision 3 -

January 7.1999

- Based on the above, spent fuel loading and handling activities create new types of accidents, but the consequences of the new types of accidents are well below EPA Protective Action Guide of I rem whole body for the early phase of an event. Therefore, although different kinds of accidents are created by sper: fuel handling and loading, the consequences of the different kinds

~of accidents do not represent a significant hazard to public health and safety.

4. The re quested license amendment does not involve a significant reduction in the margin of safety.

The Trojan Permanently Defueled Technical Specifications contain four limiting conditions of operation that cddress: 1) Spent Fuel Pool Water level,2) Spent Fuel Pool Boron Concentration,

3) Spent Fuel Pool Temperature, and 4) Spent Fuel Pool load restrictions. These Technical Specifications will remain in effect as long as spent fuel is stored in the Spent Fuel Pool, which is in accordance with their applicability statements. The sp:nt fuel loading and handling activities will not affect these Technical Specifications nor their bases.

The Cask Loading Pit,5 sre spent fuel will be loaded into the basket, is immediately adjacent to the Spent Fuel Pool. 71he x .te between the Cask Loading Pit and Spent Fe.el Pool will be opened to allow spent fuel awemblies to be moved from the spent fuel storage racks in the Spent Fuel Pool to the basket in the Cask Loading Pit. Opening the gate will allow free exchange of the water between the Cask Loading Pit and the Spent Fuel Pool. The water in the Cask Loading Pit must be at est entially the same level, boron concentration, and temperature as the Spent Fuel Pool prior to the first opening of the gate to ensure that the limiting conditions of operation are continuously satisfied for the Spent Fuel Pool. Therefore, the Cask Loading Pit will be filled, to about the same level as the Spent Fuel Pool, with water that is about the same boron ~ concentration and temperature as the Spent Fuel Pool. With these precautions, the limiting conditions of operation pertaining to Spent Fuel Pool level, boron concentration, and ,

temperature will be continuously maintained for the Spent Fuel Pool and the margin of safety will be unaffected. The level in the Cask Load Pit will not be reduced until the shield lid has been placed on the loaded basket. This configuration is consistent with the objective of 51 1

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l l Enclosure 2 to VPN-005-99 LCA 237, Revision 3 l-January 7.1999 I

protecting operations personnel from excessive dose. The loading process does not involve a significant reduction in the margin of safety.

i Spent fuel loading and handling activities will involve lifting and moving heavy loads (e.g.,

! transfer cask). Loads that will be carried over fuel in the Spent Fuel Pool racks and the heights

at which they may be carried will be limited in such a way as to preclude impact energies over [

240,000 in-lbs if the loads are dropped in accordance with LCO 3.1.4, " Spent Fuel Pool Load ;

l Restrictions." With this precaution, the limiting condition of operation pertaining to load restrictions over the Spent Fuel Pool will be satisfied for fuel stored in the Spent Fuel Pool racks ,

and the margin of safety will be unaffected. The safe load path for heavy loads being lified and ;

moved outside the Spent Fuel Pool will be located sufficiently far from the Spent Fuel Pool as to I not have an adverse effect on the Spent Fuel Pool in the unlikely event of a load drop. In addition, the mechanical stops and electrical interlocks on the Fuel Building overhead crane will provide additional assurance that heavy loads are not carried over the fuel in the Spent Fuel Pool !

racks.

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Based on the above, the spent fuel loading and iandling activities will not reduce the margin of :

l safety.

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l 7; ENVIRONMENTAL ASSESSMENT '

l Spent fuel loading and handling will be conducted in a manner having minimal impact on the environment.~ The spent fuel loading and handling process will comply with applicable Nuclear ;

i Regulatory Commission (NRC) and State of Oregon requirements to ensure that the activity does l not adversely affect the environment or the health and safety of workers or the public. As discusced j in the safety evaluation above, potential off-normal events and accidents have been postulated and l analyzed for the spent fuel loading and handling process. The results of those analyses show that 52 i i

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l l Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 l 1

spent fuel loading and handling process are well below the EPA Protective Action Guides for the

early phase of an event.

i This environmercal assessment addresses spent fuel loading and handling in preparation for storage of the spent fuet on the reinforced concrete storage pad. The environmental impact of construction of the ISFSI components and operation of the ISFSI is addressed in the Trojan ISFSI Environmental H

Report (PGE-1070), which was included as part of the 10 CFR 72 license application that PGE submitted to the NRC in March 1996. By letter dated November 25,1996, the NRC staff transmitted the Environmental Assessment (EA) and Finding of No Significant Impact (FONSI) for the construction and operation of the ISFSI. J 7.1 ENVIRONMENTALIMPACT OF PROPOSED AC' LION The overall environmental impact of spent fuel loading and handling is beneficial because moving j the spent fuel to the ISFSI will allow further decommissioning activities in the Fuel Building and i ;

Containment Building which will reduce the site inventory of radionuclides and return much of the j Trojan site to a condition suitable for unrestricted use.

7.1.1 Effects on Human Activities j- The activities related to spent fuel loading and handling will take place primarily inside the Fuel Building. This building is on the portion of the Trojan site t!at has already been developed and j used as an industrial facility. Portions of the existing Trojan .,ite are set aside for recreational use by the public. These areas include a 28-acre ecreational lake, picnic area, hiking and bicycle paths, and parking facilities. These public use r.reas are located separate from the industrial portion of the site and will be unaffected by the spent fuel loading and handling.

The work force required to complete the spent fuel loading and handling will be comparable to that associated with minor projects previously conducted at the Trojan site and will be much smaller

! ' than the work force required for refueling outages and the initial construction of the facility. The i

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Enclosure 2 to VPN-005 i LCA 237, Revision 3 l Januarv 7.' 1999

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surrounding communities have experienced the construction of several large industrial facilities m the past and facilities exist to accommodate the temporary workers that may be required for spent j _ fuel loading and handling. Therefore, there will be'no significant adverse impcets on temporary l

. housing or schools as a result of the additional work force that may be required for spent fuel i i

. losding and handling.' i Spent fuel loading and handling will have negligible effects related to ambient noise in the crea of n.

the Trojan site. Any noise generated will be localized and of temporary duration. The Trojan site is bor'dered on the east by the Columbia River and on the west by a stata highway. The nearest l residential area is located approximately one mile north of the facility. Therefore, there will be no

' significant impact due to noiw generated during spent fuel loading and handling.

l Dust generation will be minimized during spent fuel loading and handling by using existing building ventilation systems and supplemental high efficiency particulate air filtration units, if necessary Other dust control measures will be implemented if necessitated by forklift / truck movement of the concrete cask to the ISFSI pad.

l Spent fuel loading and handling will be conducted in accordance with the existing Portland General Electric (PGE) safety program to minimize the risk of an industrial accident. The PGE safety I program has contributed to a solid safety record during Trojan's seventeen years of operation and includes detailed procedures on topics such as emergency preparedness; hazard communication; accident investigation and reporting; electrical hazards; welding and cutting practices; placement

- and use ofladders and scaffolding; use of cranes, hoists, and associated rigging; and fire protection -

and prevention. Vendor organizations participating in the project will be required to submit industrial health and safety programs for PGE approval or work under the PGE program.

' 7.1.2 Effects on Terrain. Vegetation and Wildlife TAs stated above, spent fuel loading and handling activities will take place in the Fuel Building, l which are located on a portion of the Trojan site that has been previously developed for industrial 1

54 I

l Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 usage. Portions of the site which have been left in their natural state will not be disturbed by these activities. Following loading of the basket into a concrete cask, the concrete cask will be moved by a forklift or'similar vehicle to the ISFSI reinforced concrete storage pad. The storage pad and the route to the storage pad are also within the portion of the Trojan site previously developed for industrial use, hence, are also separated from the areas of the site that have been lefl in their natural I

. state. Therefore, there will be no impacts on terrain, vegetation, or wildlife on or near the Trojan ;

site. I 7.13 Effects on Adiacent Waters and Aanatic Life I

The Columbia River is adjacent to the Trojan site. The Service Water System uses water from the Columbia River to cool the Component Cooling Water System, which in turn cools the Spent Fuel Pool water. In this way, decay heat from the spent fuel is rejected to the Columbia River. As the spent fuel is removed from the Spent Fuel Pool and loaded into casks, fewer spent fuel assemblies :

l will be left in the Spent Fuel Pool, which will reduce the amount ofdecay heat that is rejected to the Columbia River. Therefore, the loading of spent fuel casks will reduce the impact of the Trojan Nuclear Plant on the Columbia River and aquatic life. If alternative cooling systems are used

- (modular air cooling system), there will be a concomitant reduction in impact on the Columbia River and aquatic life.

Water usage on-site during the spent fuel loading and handling will be a small fraction of previous water usage at Trojan Nuclear Plant during power operations.

i 7.1.4 Nonha7ardous/ Hazardous Waste !

< 1 Radioactive wastes generated during spent fuel loading and handling will be handled and disposed ofin accordance with the existing Radiation Protection Program. Such wastes include airbome particulate activity, filters used for processing liquid / gaseous radioactive wastes, and liquids !

associated with decontamination activities. Liquid and gaseous radioactive wastes will be processed, sampled, monitored, and discharged in accordance with the plant effluent requirements 55

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Enclosure 2 to VPN-005-99 1 l LCA 237, Revision 3 Jaluary 7.1999

. specified in the Offsite Dose Calculation Manual. The process control program and approved plant procedures will be used to radiologically monitor, handle, and dispose of solid radioactive waste.

No hazardous or mixed (radiologically contaminated and chemically toxic) wastes are expected as a result of spent fuel loading and handling. However, handling and disposal of any radiologically uncontaminated hazardous waste that is encountered will be controlled by the existing hazardous

. materials processes. These processes involve evaluation of the hazardous material and approval of 5 methods for its handling and disposal. ' Approved plant procedures are in place to reduce the potential for generation of mixed wastes and to control its handling and storage onsite.

Nonhazardous, nonradiologically contaminated waste will be handled in accordance with Trojan's normal waste disposal practices.

7.1.5 Occupational /Public Radiation Exposure >

Occupational radiation exposures resulting from spent fuel loading and handling are anticipated to be low. The activities, number of personnel required, estimated duration, estimated working dose rate, and estimated exposure for the loading and handling of spent fuel casks are provided in the l ISFSI SAR Table 7.4-3. The esti. mated exposure compares favorably to that typically experienced by workers at operating nuclear plants.

The spent fuel loading and handling operations will occur inside the Fuel Building. The general radiation levels in the proposed work areas near the Spent Fuel Pool, Cask Loading Pit, DAS, and Cask Wash Pit inside the Fuel Building are about 0.0002 roentgen per hour. Once the basket is l loaded with spent fuel and the transfa cask containing the loaded basket is lifted from the Cask l

~

l Loading Pit, occupational workers will be near the sides of the transfer cask or on top of the basket. I to weld the lids into the basket, connect vent and drain lines, etc. The radiation levels adjacent to the transfer cask and on top of the basket will be considerably higher than the general radiation j . levels near the Spent Fuel Pool, Cask Loading Pit, DAS, and Cask Wash Pit because the radiation

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L LCA 237, Revision 3 1 l January 7.1999 l

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source (spent fuel) will no longer be at a distance of 23' with 23' ofintervening water for radiation I shielding. ;

The design of the basket, shield lid, and transfer cask reduce the radiation levels on top of the basket 1

and sides of the transfer cask to accommodate the work that must be performed in close proximity to the basket. In addition to the radiation shielding provided by the design of the basket, shield lid, and transfer cask, the design of the basket assembly, such as a ball valve on the basket vent line, and the c hoice of equipment, such as the automatic welding machine for the basket lids, reduces the time that workers need to be in close proximity to the basket.

Procedures and practices will be employed to maintain occupational worker exposures ALARA beciuse the estimated working dose rates are in excess of those specified for a radiation area and a !

high radiation area as defined by 10 CFR 20. These measures will include: 1) fuel loading l pro:edares that follow accepted practice and build on existing experience; 2) loading spent nuclear j fue'l in the basket within the controlled environment of the Fuel Building to minimize the spread of i cor tamination; 3) generally loading the most radioactive fuel in interior basket positions; l

4) injecting filtered, bomted water into the annulus between the transfer cask and basket to minimize coritamination of the basket extemal surface; 5) placing the shielding lid on the basket while the ;

transfer cask and basket remain in the Cask Loading Pit; 6) decontaminating the exterior of the transfer cask and welding the basket lids while the basket is still filled with water; 7) draining the bas ket while still housed in the transfer cask; 8) using portable shielding as necessary; 9) using the shielded transfer cask with remotely operated bottom doors to transfer the basket to the concrete cask; 10) placing a shielding ring over the annular gap between the concrete cask and basket; 11) using pre-job briefings prior to fuel movement and cask loading sequence.

In addition to using procedures and pra s to maintain personnel exposure ALARA, controls and monitoring are implemented to ensure tk. wrkers do not receive an annual dose in excess of Tmjan's administrative dose guidelines. These controls and monitoring are governed by the Trojan Radiation Protection Program.

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[ Enclosure 2 to VPN-005-99 L LCA 237, Revision 3 i l ' January 7.1999 l In summary, the design of the basket, shield lid, and transfer cask includes radiation shielding to accommodate the spent fuel loading and handling work processes that must be performed in close proximity to the radiation source (spent fuel). Also, the basket, transfer cask, and support I Lequipment have been designed (or chosen if commercially available) to be operated quickly or remotely to minimize personnel radiation exposures. Additional procedures and practices will be

- implemented to ensure that personnel radiation exposures are ALARA and are controlled and monitored to ensure that regulatory limits are net exceeded. As a result of these design features, l ' procedures, controls, and monitoring, spent fuel loading and har.dling will not significantly increase

- individual and cumulative occupational exposures.

t Public exposure was not separately determined because no significant amounts of radioactive mawrials will be transported out of the industrial area and no significant amounts of gaseous or liquid radioactive effluents will be released to unrestricted areas as a result of spent fuel loading and handling activities. I'ublic exposure associated with ISFSI operation is addressed by separate licensing actions.

7.2 ENVIRONMENTAL IMPACT OF OFF-NORMAL EVENTS AND ACCIDENTS As discussed in the safety evaluation above, postulated off-normal events and accidents and natural ;

. phenomena that could potentially have an adverse impact on the environment or on the health and L safety of workers or the public have been analyzed. The results of those analyses demonstrate that 1

- the postulated events have minimal significance from a public health and safety and environmental impact standpoint. Of the postulated off-normal events and accidents, only three types were l ' determined to have radiological consequences: a fuel assembly drop into a basket loaded with spent L

fuel, a transfer cask drop or mishandling event prior to the shield and structural lids being welded in the basket, and a basket draindown/ vacuum drying system failure (passive failure of a pressurized l line).

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l l-L Enclosure 2 to VPN-005-99 LCA 237, Revision 3 January 7.1999 '

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7.2.1 Fuel Assembly Drop into a Basket Loaded with Soent Fuel i

j- The potential for a fuel assembly drop into a basket loaded with spent fuel is highly unlikely L because the fuel handling equipment is designed with safety factors that ensure that the structural and lifting capability of the equipment well exceeds the weight that will be lifted. In addition, the fuel handling equipment will be load tested prior to use to ensure that the structural and lifting

. capabilities are verified to be adequate. Finally, administrative controls in the form of approved load handling procedures, training and certification of fuel handlers, maintenance of the fuel handling equipment, and inspections of the fuel handling equipment provide additional assurance that a fuel assembly drop will not actually occur.

The consequences at the site boundary from a postulated fuel assembly drop into a basket loaded l with spent fuel would be about 0.0005 rem whole body, about 0.0006 rem thyroid, and about 0.0455 l rem skin. These doses are well below the EPA Protective Action Guides of I rem whole body,5 I rem thyroid, and 50 rem skin for the early phase of an event. These doses are conservatively estimated by assuming failure of 100% of the fuel rods in the dropped assembly and r. cooling time of 6 months (Trojan fuel will have been cooling for a minimum of 6 years and some fuel will have l

! cooled for 17 years when cask loading will occur). In addition, this event was previously evaluated for the defueled condition. Therefore, this event dc,es not introduce any consequences that have not been previously evaluated and the environmental impact is negligible.

7.2.2 Transfer Cask Drop and Mishandling Events Prior to Shield and Structural Lids Being Welded into the Basket t

The potential for a transfer cask drop or mishandling event is highly unlikely because the load handling equipment is designed with safety factors that ensure that the structural and lifting

. capability of the equipment well exceeds the weight that will be lifted. In addition, the load l.

handling equipment will be load tested and interlock operation verified prior to use to ensure that they are adequate. Finally, administrative controls in the form of approved load handling

. procedures, training of op.ators, maintenance of the load handling equipment, and inspections of 59 l

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. Enclosure 2 to NPN-005-99 '

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' LCA 237, Revision 3 -

January 7.1999 i

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~ the load handling equipment provide additional assurance that the potential for transfer cask drop or

!' mishandling event is minimized.

7.2.3 = Basket Draindown/ Vacuum Drving System Failure The potential for passive failure of a pressurized line connected to a basket loaded with spent fuel is i

~

unlikely because the pipe and flexible hose lines are rated to greater than 1,000 psig. The I probability of a passive failure of one of these lines is low, especially considering that the maximum )

operating pressure of these lines is about 15 psig. The consequences at the site boundary from a postulated passive failure of a pressurized line connected to a basket loaded with spent fuel would be less than 1/1000th the.whole body PAG if Auxiliary Building / Fuel Building ventilation is l running and about 1/3rd the whole body PAG limit if the Auxiliary Building / Fuel Building I ventilation is not running. These doses are well below the EPA Protective Action Guide of I rem

' whole body for the early phase of an event. These doses are conservatively estimated by assuming 15 percent of the contamination on the fuel becomes airborne.

~. Based on the limited likelihood of an off-normal event or accident occurring during spent fuel loading and handling, the conservatisms used in estimating the potential radiological consequences of postulated off-normal events and accidents, and the potential radiological consequences of postulated off-normal events and accidents being well below the EPA Protective Action Guides, it is concluded that the potential for and radiological consequences of off-normal events or accidents during the spent fuel loading and handling at the Trojan Nuclear Plant will not significantly impact the environment.

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Enclosure 2 to VPN-005-99

. LCA 237, Revision 3 Januatd. I999 -

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7.3' ALTERNATIVES TO PROPOSED ACTION The proposed action is to place spent fuel in spent fbel storage casks inside the Fuel Building in order to transfer spent fuel from the Spent Fuel Pool to the Trojan ISFSI. The alternative to this ;

proposed action would be to not remove the spent fuel from the Spent Fuel Pool (i.e., no action) or j to ship the fuel to another facility for storage. !

- 7.3.1 Not Removing Spent Fuel from the Sntut Fuel Pool Alternatives that do not remove the spent fuel from the Spent Fuel Pool are not viable .

decommissioning attematives. The objective of decommissioning is to restore a raciaactively l contaminated facility to a condition such that there is no unreasonable risk to the public health and

' safety following license termination. If the spent fuel is not removed from the Spent Fuel Pool, then

!. the Trojan Nuclear Plant could not be decommis sioned. ;

G The Trojan Nuclear Plant will be decommissioned using the DECON method as described in I

r NUREG-0586. DECON is a decommissioning alternative that provides for equipment, structures, and portions of a facility containing radioactive contaminants to be removed or sufficiently.

l -

decontaminated to allow release of the property for unrestricted use relatively soon after cessation of power operations. PGE anticipates terminating the 10 CFR 50 license in 2002. Therefore, spent ;

fuel must be remover A m the Spent Fuel Pool far enough in advance of 2002 to allow the

. completion of decommissioning activities and free release surveys. As the DECON

' decommissioning option has been chosen, removing the spent fuel from the Spent Fuel Pool will be necessary.

U The Trojan Nuclear Plant Deconunissioning Plan, which describes use of the DECON method for

decommissioning, has been reviewed and found to be acceptable by the Nuclear Regulatory 1

Commission Staff. The Staff's review and conclusions were issued on December 18,1995 in a ;

Safety Evaluation Report, Environmental Assessment, and Final Finding of No Significant Impact.

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E' nc19sure 2 to VPN-005-99 '

< LCA 237, Revision 3 l January 7.1999

'In summary, the DECON method has been chosen for decommissioning of the Trojan Nuclear Plant

, and the Nuclear Regulatory Staff has found the Trojan Nuclear Plant Decommissioning Plan 4

L acceptable. The Trojan Nuc' ear Plant Decommissioning necessitates removing the spent fuel from the Spent Fuel Pool and the DECON ' option necessitates removing the spent fuel from the Spent

[ Fuel Pool relatively soon after cessation of power operations. For these reasons, not removing the I spent fuel from the Spent Fuel Pool was not considered a viable altemative.

1.12 Shipment of Troian Soent Fuel to Another Facility -

t I POE anticipates terminating the 10 CFR 50 license in 2002. Therefore, spent fuel must be removed L from the Spent Fuel Pool far enough in advance of 2002 to allow the completion of

} decommissioning activities and free release surveys. As the DECON decommissioning option has

[ been chosen, removing the spent fuel from the Spent Fuel Pool will be necessary. There are

currently no'other facilities that can or will accept Trojan's spent fuel for wet or dry storage.

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8. SCHEDULE CONSIDERATION h ]

POE plans to begin loading spent fuel into casks and moving the casks to the ISFSI concrete storage :

pad on April 12,1999. Accordingly, PGE requests approval of this 10 CFR 50 license amendment . -l 1

by March 1999, to allow adequate time for finalization of spent fuct loading and handling l procedures as a result of mock-up and dry run testing.

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1 ENCLOSURE 3 4

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NOTE: This document is a reproduction of Facility Operating License NPF-1 that includes License Amendments through 198. This was produced for case of reference for anyone reviewing this document.

PORTLAND GENERAL ELECTRIC COMPANY JJiE CITY OF EUGENE. OREGON PACIFIC POWER AND LIGHT COMPANY

DOCKET NO. 50-344

- TROJAN NUCLEAR PLANT FACILITY OPERATING LICENSE Amendment No.198 License No. NPF-1

..1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for license filed by Portland General Electric Company, the City of Eugene, Oregon, acting by and through the Eugene Water and Electric Board, and Pacific Power and Light Company' (the licensees), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the

Commission rules and regulations set forth in 10 CFR Chapter I and all required notifications to othtr agencies or bodies have been duly made; B. Construction of the Trojan Nuclear Plant (facility) has been substantially completed in conformity with Construction Permit No. CPPR-79 and the application, as amended, the s

provisions of the Act and the rules and regulations of the Commission; ,

e C. The facility will be maintained in conformity with the application, as amended, the N provisions of the Act, and the rules and regulations of the Commission; -

m D. ' There is reasonable assurance: (i) that the activities authorized by this license can be

. conducted without endangering the health and safety of the public, and (ii) t6t such activities will be conducted in compiiance with the rules and regulations of the 7 . Commission; g' i

'l ' ' Pacific Power and ' Light Company, whi':h has a 2.5 percent ownership

interest in Trojan, has merged with Utah Power and Light Company to become a new corporation named PC/UP&L Merging Corporation, which will change its name to

, PacifiCorp, but will operate under the assumed business name of Pacific Power and Light Company.

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E. Portland General Electric Company is technically qualified and the licensees are financially qualified to engage in the activities authorized by this license in accordance with the rules and regulations of the Commissions; F. The licensees have satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection Requirements and indemnity Agreements," of the Commission regulations; G. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the pubi;c; H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available altematives, the issuance of Facility Operating License No. NPF-1 set forth herein is in accordance with Appendix D to 10 CFR Part 50, of the Commission regulations and all applicable requirements have been satisfied; and

1. The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this license will be in accordance with the Commission regulations in 10 CFR Part 30,4u, and 70, including 10 CFR Sections 30.33,40.32, and 70.23 and 70.31.

2. Facility Operating License No. NPF-1 is superseded in its entirety by Possession Only License (POL) No. NPF-1 hereby issued to the Portland General Electric Company, The City of Eugene, Oregon, and Pacific Power and Light Company to read as follows:

A. ,This license applies to the Trojan Nuclear Plant, a pressurized water nuclear reactor and associated equipment (the facility) owned by the licensees. The facility is located

-.on Portland General Electric Company site on the west shore of the Columbia River in Columbia County, Oregon, and is described in the Updated Final Safety Analysis Report, as supplemented and amended in accordance with 10 CFR 50.71(e), and the Environmental Report as supplemented and amended (Supplernents 1 through 3).

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

-(1) Pursuant to Section 103 of the Act and 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," to possess, use, but not operate the facility at the designated location in Columbia County, Oregon in accordance with the

. procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear

. material as reactor fuel, in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended;

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(3) Pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess and use

' at any time any byproduct, source, and special nuclear material as sealed neutron sources, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the' Act and 1' 0 CFR Parts 30,40, and 70, to receive, possess and use

.in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

'(5) Pursuant to the Act and 10 CFR Parte 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operations of the facility!

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulatiens in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is not authorized to operate the facility as a nuclear reactor.

(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.197, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. The licensee chall maintain the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Soent Fuel Pool Modificatiq0

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The licensee is authorized to modify the spent fuel pool as described in the application dated August 1,1983, and amended October 31,1983.

f (4) QWAl!!ylaturance Activities to_which a Quality Assurance Program is applicable shall, after the date of issue of this license, be conducted in accordance with the Nuclear Quality Assurance Prograrn.

(5) The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.18 of the NRC Fire Protection Safety

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4 Evaluation Report on the Trojan Nuclear Plant dated March 22,1978. These modifications shall be completed by the end of the second refueling outage of the Trojan facility and prior to retum to operation for Cycle 3. In addition, the licensee shall submit the additionalinformation identified in Table 3.2 of this Safety Evaluation Report in accordance with the schedule contained therein, in the event these dates for submittal cannot be met, the licensee shall submit a report explaining the circumstances, together with a revised schedule.

T (6) Primarv Shieldino Modification The licensee is authorized to modify the primary shielding design as described in t

PGE letter dated April 22,1977, as supplemented and amended by letters dated September 22 and 23,1977, December 22,1977, January 4 and 24,1978, March 20, and April 4,1978.

(7) Soent Fuel Assembiv Shiocino Cask Deleted (Amendment 196)

(8) Fire ProtectiQD The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in PGE-1012, " Trojan Nuclear Plant Fire Protection Plan" and as approved in the SER dated September 22,1993 subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely impact the safe storage of irradiated fuel or increase the likelihood of an offsite release of radio'active material due to a fire.

(9) Soent Fuel Pool Debris Processing The licensee is authorized to process spent fuel pool debris in accordance with PGE letter dated October 23,1996.

(10) Loadino of Fuelinto Casks in the Fuel Building l l

The licensee is authorized to load spent nuclear fuel and other materials into transfer and l storage casks in the Fuel Building in accordance with License Change Application (LCA) l 237, transmitted by PGE letter dated January 7,1999. Changes to LCA 237 are permitted l without changing this license providing they are evaluated in accordance with 10 CFR l 50.59 and reported in accordance with 10 CFR 50.71(e). l t

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D. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provision of the l Miscellaneous Amendments and Search Requirements to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards information protected under 10 CFR 73.21, are entitled:

" Trojan Nuclear Plant Security Plan", with revisions submitted through October 7,1988;

" Trojan Nuclear Plant Security Force Training and Qualification Plan", with revisions submitted through June 10,1988; and " Trojan Nuclear Plant Safeguards Contingency Plan', with revisions submitted through October 7,1988. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth ,

herein.

l j E. This license is effective as of the date of issuance.

l l FOR THE NUCLEAR REGULATORY COMMISSION 1

i Brian K. Grimes, Director l l !

l Division of Operating Reactor Support l l Office of Nuclear Reactor Regulation .

l l

Attachment:

Appendix A Technical Specifications Date of issuance: June 9,1997 l

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ML20198T185

Text

Attachment:

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