ML20137M768

Rev 1 to Defueling Canisters, Technical Evaluation Rept

ML20137M768
Person / Time
Site:	Three Mile Island
Issue date:	09/03/1985
From:	Boldt G, Rider R, Smith E GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20137M741	List: ML20137M735 ML20137M768 ML20137M776
References
0334Y, 15737-2-G03-114, 15737-2-G03-114-R01, 15737-2-G3-114, 15737-2-G3-114-R1, 334Y, NUDOCS 8509130323
Download: ML20137M768 (33)

Text

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Nuclear TER 15737-2-G03-114 REV.1 ISSUE DATE 3/2%N5' O ITS El NSR 0 NITS TMI-2 DIVISION TECHNICAL EVALUATION REPORT FOR Defueling Canisters COG ENG M148/N88[ DATE 2/2MW RTR SYwrJ YE4/ / /W bd b DATE DATE 3/> 2 3 -u,/F4 ~

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N O.

Nuclear 2-G03-114 Title PAGE OF TER for Defueling Canisters 2 33 Rev.

SUMMARY

OF CHANGE O Issued for Initial Use 1 Update to incorporate design change from vibrapacked B4C powder to sintered B4 C pellets, discussion of maximum particle size expected in filter canister, increase in load limit on fuel canister lower support plate from 350 to 550 lbs, addition of keff criteria for plant accident condition (< 0.99),

discussion of effects on criticality analyses caused by a) change to B 4C pellets, b) lower storage pool water temperature, and c) fuel particle size, addition of section regarding hydrogen controls within the canister.

(PD 3215 1944

O LATEST LATEST LATEST LATEST LATEST LATEST 1 LATEST REV. SHEET REV. SHEET REV. SHEET REV. SHEET REV. SHEET REV.

SHEET REV. SHEET 1 1 .

2 1 3 1 4 1 5 0 6 1 7 1 8 1 9 1 10 1 11 1 12 1 13 1 14 1 15 1 16 1 17 1 18 1 19 1 20 1 21 1 22 1 23 1 24 1 25 1 26 1 27 1 28 1 29 1 30 1 31 1 12 1 13 1

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REVISION STATUS SHEET JOB 15737 REV.

g g SPEC. N O.

M gNT N Technical Evaluation Report for 2-G03-ll4 DOCUMENT TITLE: Defuelinn Canisters p PAGE 3 OF 33 e

15737-2-C03-114 Table of Contents

.P,, age i

5 j 1.0 Introduction 1.1 Purpose 5 '

1.2 Scope 5

• i -

2.0 Canister Description 6 l >

1 6

2.1 Codes and Standards 8

2.2 Fuel Canister 8 i 2.3 Knockout Canister 9

' 2.4 Filter Canister i

17 j 3.0 Technical Evaluation 17 l 3.1 Canister Structural Evaluation 3.2 Canister Criticality Evaluation 19 l

3.3 Canister Hydrogen Control Evaluation 23 4.0 Radiological Considerations 29

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30 l 5.0 10CFR 50.59 Evaluation 6.0 Conclusions 32 f

33 7.0 References Attachments

1. TMI-2 Transfer System Criticality Technical Report j

2. Assessment of a Drained Pool Scenario i t

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1.0 Introduction i

' canisters are required during the defueling at TMI-2 to retain core j debris ranging from very small fines to partial length fuel assemblies.

These canisters provide effective long term storage of the TMI-2 core i debris. Three types of canisters are required to support the defueling j . system to be used at TMI-2: filter, knockout, and fuel canisters. '

i 1.1 Purpose l The purpose of this report is to show that the canisters are i designed to remain safe under normal operation and handling -!

conditions as well as postulated drop accidents and storage. I i

Section 2.0 of this report describes the three types of canisters. l l

Section 3.0 addresses the safety of the canister design considering i design drop analyses and drop tests and criticality analyses. i Requirea/nts for spacing of the canisters in an array under normal conditions are also addressed. Section 4.0 outlines the j radiological concerns associated with the handling and storage of ,

! the canisters. Section 5.0 draws conclusions about the safe

} operation and handling of the canisters. ,

i ,

j 1.2 Scope l

l This report addresses only those safety issues associated with the loading, handling and storage of the canisters as related to canister design. Analyses of the design drop considers only the  ;

effect of that drop on a canister; damage to other components is not >

j considered. Actual handling of the canisters is not addressed in this report and neither are the shielding requirements for canister.

handling with the exception that the criticality concern associated with the use of lead shields around the canisters is addressed in t Attachment 1. Also, the criticality concern associated with a i

drained spent fuel pool is addressed in Attachment 2. Canister l

performance during defueling is addressed here only as it impacts 4 the safe use of the canister. Canister interfaces with the f l

defueling equipment, canister handling equipment and the fuel '

{

transfer system are not covered in this report. The issues related 1 to canister use (e.g. shielding requirements, load drops, etc.) are i

! evaluated in the Safety Evaluation Report for Early Defueling of the r

' THI-2 Reactor Vessel (reference'3). The transportation requirements l for the canisters will be separately addressed.

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15737-2-G03-114 2.0 Canister Description This section presents the designs of three canisters to be used in defueling TMI-2. Compatible with the RCS and spent fuel pool environment, these canisters provide long term storage of the TMI-2 core debris. In conjunction with the defueling system, the canisters will retain and encapsulate debris ranging from micron size particles to partial length fuel assemblies.

The canisters consist of a circular pressure vessel housing one of three l

types of internals, depending on the function of the canister. Except i

for the top closures, the outer shell is the same for all three types of l canister design. It serves as a pressure vessel protecting against

! leakage of the canister contents as well as providing structural support for the neutron absorbing materials. It is designed to withstand the pressures associated with normal operating conditions. A reversed dish end is used for the lower closure head for all of the canisters while the upper closure head design varies according to the canister's function.

The canisters are non-buoyant under all storage and operational conditions.

Each canister contains a recombiner catalyst package incorporated into the upper and lower heads. The catalyst recombines the hydrogen and oxygen gases formed by radiolytic decomposition of water in the canisters.

Each canister has two pressure relief valves which are connected to the canisters using Hansen quick disconnect couplings. The low pressure relief valve has a pressure setpoint of 25 psig. The high pressure ASME code relief valve has a 150 psig setpoint.

l 2.1 Codes and Standards l

l The defueling canisters have been classified as Nuclear Safety ,

j Related for criticality control purposes. l l

They are designed and designated for fabrication in accordance with the following codes and standards:

ANSI /ANS 8.1 (1983) American National Standards Institute /

American National Standard, Nuclear Criti-cality Safety in Operations with Fissionable Materials Outside Reactors l

ANSI /ANS 8.17 (1984) American National Standards Institute /

[ American National Standard, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors ANSI N45.2 (1977) American National Standards Institute, Quality Assurance Program Requirements for Nuclear Power Plants 1

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1 15737-2-G03-114 ANSI N45.2.2 (1972) American National Standards Institute, Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants ANSI N45.2.11 (1974) American National Standards Institute, Quality Assurance Requirements for the Design of Nuclear Power Plants

! ANSI N45.2.13 (1976) American National Standards Institute, Quality Assurance Requirements for Control of Procurement of Items and Services for i Nuclear Power Plants ANSI /ASME NQA-1 (1979) Quality Assurance Program Requirements for Appendix 17A-1 Nuclear Power Plants, Nonmandatory (including ANSI /ASME Guidance on Quality Assurance Records NQA-la-1981 Addenda)

ANSI /ASME NQA-1 (1979) Quality Assurance Program Requirements for Supplement 17S-1 Nuclear Power Plants, Supplementary (including ANSI /ASME Requirements for Quality Assurance Records NQA-la-1981 Addenda)

ASME Boiler and Pressure American Society of Mechanical Engineers, Vessel Code, Section Pressure Vessels VIII, Part UW (lethal)

(1983)

ASME Boiler and Pressure American Society of Mechanical Engineers, Vessel Code,Section IX Welding and Brazing Qualifications j (1980)

ASTM A 312 (1982) American Society for Testing and Materials, Seamless and Welded Austenitic Stainless Steel Pipe SNT-TC-1A (1980) American Society for Nondestructive Testing, Recommended Practice for Nondestrutive Testing, Personnel Qualification and Certification 10 CFR 21 Reporting of Defects and Noncompliance 10 CFR 50, Appendix A General Design Criteria for Nuclear Power Plants 10 CFR 50, Appendix B Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR 72 Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation NUREG-0612 Control of Heavy Loads at Nuclear Power

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15737-2-G03-114 i

,I 4 2.2 Fuel Canister i

i The fuel canister is a receptacle for large pieces of core debris to be picked up and placed in the canister. The fuel canister consists 1: of a cylindrical pressure vessel with a flat upper closure head.- It uses the same outer shell as ebe other canisters. Within the shell,

! a full length square shroud forms the internal cavity (see Figure i 2.2-1). This shroud is supported at the top by a bulkhead that mates with the upper closure head (see Figure 2.2-2). Both the ,

shroud and core debris rest on a support plate that is welded to the  ;

j shell. The support plate has impact plates attached to absorb canister drop loads and payload drop loads.

The shroud assembly consists of a pair of concentric square stainless steel plates seal welded to completely enclose four sheets .

l of Boral, a neutron absorbing material (see Figure 2.2-1). The

> shroud internal dimensions are larger'than the cross section of an undamaged fuel assembly. The shroud external dimensions are

! slightly smaller than the inner diameter of the canister, thus

{

providing support at the shroud corners for lateral loads. The void area outside of the. shroud is filled with a cement / glass bead  ;

mixture to the maximum extent practical to eliminate migration of i

the debris to an area outside of the shroud during a design basis '

accident. s g

j The upper closure head is attached to the canister by eight equally i spaced bolts. These bolts are designed for the design pressure  :

1 loads, handling loads, and postulated impact force due to shifting l of the canister contents during an in plant load drop or a shipping (

l accident. ,

1 i 2.3 Knockout Canister i

i Designed to separate debris ranging in size from 140 microns up to i' approximately the size of whole fuel pellets (whole fuel pellets i included), the kneckout canister, Figure 2.3-1, is part of the j Fines / Debris Vacuun System. The influent comes directly from the -

! defueling vacuum' system inside the reactor while the outlet flow goes to a filter canister for further treatment. Flow fittings are 2" can and groove type similar to the filter canister fittings and

are capped or plugged after use. Externally, the knockout canister i

! is similar to the other canisters, using the same outer shell ,

I design. It also incorporates the same handling tool interface.

i j The internals module for the knockout canister is supported from a 1 lower header welded to the outer shell. An array of four outer neutron absorber rods around a central neutron absorber rod is located in the canister for criticality control. The four outer rods are 1.315" 0.D. tubes filled with sintered B 4C pellets'. l I The central absorber rod is comprised of an outer strongback tube surrounding a 2.125" 0.D. tube filled with sintered 5 C 4 pellets. l i Lateral support for the neutron absorber rods and center assembly is provided by intermediate support plates.

ReV. 1 0334Y

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15737-2-G03-114 The influent flow is directed tangentially along the inner diameter of the shell, setting up a swirling action of the water within the canister. The large particulates settle out and the water moves upwards, exiting the canister through a machined outlet in the head. A full flow screen ensures that particles larger than 850 microns will not escape from the knockout canister. This screen has been designed to withstand the maximum pressure differential across the screen that can be developed by the vacuum system equipment.

2.4 Filter Canister As part of either the Defueling Water Cleanup System or the Fines / Debris Vacuun System, the filter canisters are designed to remove small debris particles from the water. Externally, it is similar to the other canister types. The f11ter assembly bundle l that fits inside the canister shell was designed to remove particulates down to 0.5 (nominal) microns. Flow into and out of the filter canister is through 2 1/2" can and groove quick disconnect fittings (Figure 2.4-1).

The internal filter assembly bundle consists of a circular cluster l of 17 filter elements, a drain line and a neutron absorber assembly (Figure 2.4-2). The influent enters the upper plenum region, flows d on past the support plate, through the filter media and down the fiher element drain tube to the lower sump. . The flow is from outside to inside with the particulate reasining around the outer perime*.er of the filter elements. The filtered. water exits the canister via the drain line.

A filter element consists of 11 modules. Each module consists of pleated filter media forming an annulus around a central, perforated drain tube (Figure 2.4-3). Fabricated from a porous stainless steel material, the media is pre-coated with a sintered metal powder to control pore size. Bands are placed around the outer perimeter of the pleated filter media to restrict the unfolding of the pleats.

The filter assembly bundle is held in place by an upper support l plate and lower header. The lower header is welded to the outer shell of the canister to provide a boundary between the primary and secondary side of the filter system. The upper header is equipped with a series of openings to allow for the passage of the influent into the filter section of the canister and to protect the filter media from direct impingement of particles carried in the influent flow. Six tie rods position the upper plate axially relative to the

-lower support plate.

The filter canister has a central neutron absorber rod that is comprised of an outer strong back tube surrounding a 2.125" 0.D.

tube' filled with sintered B4 C pellets.

The filter canisters are not expected to contain significant quantities of fuel particles larger than 850 microns. The filter canisters are used with the defueling water cleanup system (DWCS)

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15737-2-G03-114

and the defueling vacuum system. The DWCS is used to process both spent fuel pool / fuel transfer canal water and reactor coolant system (RCS) water. In the RCS, the DWCS suction is located in the upper region of the reactor vessel, where large fuel debris (i.e., >

850p) would not be expected to be suspended in solution. The spent fuel pool / fuel transfer canal is not expected to contain significant quantities of fuel particles larger than 850 microns.

Consequently, the DWCS filter canisters are not expected to contain significant quantities of fuel particles larger than 850 microns.

When the filter canisters are used in conjunction with the defueling vacuum system, they are located downstream of the knockout canisters. Proof of principle testing (Reference 11) has shown that for the planned vacuum system flowrates, minimal quantities, if any, of 850 micron or larger sized particles would be carried out of the knockout canister. Additionally, the discharge of the knockout canisters are equipped with a 841 micron screen to prevent larger fuel particles from exiting the knockout canister. Thus the vacuum system filter canisters are not expected to contain significant quantities of fuel particles larger than 850 microns.

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15737-2-c03-114 , ,

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15737-2-C03-114 Figure 2.4 2 -

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15737-2-G03-114 i

i 3.0 Technical Evaluation This section summarizes the safety issues which were evaluated during the 4 design of the canisters. These issues deal with the expected performance

of the canisters during normal operations and various design basis ,

i events. Safety issues which were evaluated. include structural forces on  ;

a canister as a result of a drop accident, criticality issues associated with both single canisters and canisters in the storage racks and the

! l t canister / storage rack interface, including any constraints on the storage rack design.

f 3.1' Canister Structural Evaluation i'

A structural evaluation has been performed (Reference 1) which addresses both the loads imposed on the canister during normal j operations (loading and handling) as well as postulated drops.

A combination of analytical methods and component testing is used to l Verify.the adequacy of the design. Acceptance criteria for normal

! operation is based on the ASME Pressure Vessel Code,Section VIII,

! Part UW (lethal).

i l Normal operation of the canister imposes very small loads on the

! canister internals. The largest load on the internals is the

combined weight of the debris and internals. The configuration of the canisters is such that only the lower plate assembly that l supports both the debris and internals experiences any significant j

1 loads. Results of the stress analysis shows a large margin of j safety for the lower plate assembly and its weld to the outer shell  !

i for all canister types. The canister shell is subject to ASME Code, I Section VIII standards. Verification of the canister shell structural design to the ASME requirements has been performed (Reference 1). The canisters'are designed for a combined (canister, debris, and water) static weight of 3500 pounds.

During normal handling operations (lifting), the static plus dynamic  ;

i loading considered in the design of the handling features of the f i canister is 1.15 times the static lifted weight. Results froe the  !

{ structural evaluation show an acceptable margin of safety

! considering the stress design factors specified in NUREG-0612 and, 3

ANSI N14.6. -

3 Normal loading of the fuel canister presents two cases for i{

j evaluation. First is the capability of the lower support plate to [

! absorb the impact of debris accidently dropped into the canister.  !

! Results of the dynamic impact evaluation show that the support plate i 1 can accommodate loads of up to 350_lbs (23% of a fuel assembly) dropped, in air, the full canister length without a failure of the i lower plate to shell weld. This weight limit increases to 550 lbs.

l (in air weight) if credit is taken for the drag forces of the m eer i in the canister. Second is the verification that placement of

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15737-2-G03-114 >

i jt f debris within the canister will not rupture the shroud's inner

vall. This would expose the Boral sheets to the RCS water which 4 could cause corrosion of the boral. However, examination of the i shrouds subjected to drop tests (reference 10) indicate that the inner wall is resistant to debris impacts and scrapes.

L l A dewatering system is used to remove water from all canisters prior f- to shipment. During this procedure, a pressure differential is

{

developed across the debris screen, lower support plate and drain j tube. The maximum pressure differential allowed, via a safety

- relief valve in the dewatering system, across canister internal components during dewatering is 55 psi. The canister internals are  !

4 designed for a maalaus differential pressure of 150 psi although l filter media differential pressure is limited by design to 60 poid. i

/

Hence, an adequate margin of safety exists for the dewatering [

process.  ;

4

! The canisters are capable of withstanding enveloping accidents.

Vertical drops' of 6'-l 1/2" in air followed by 19'-6" in water, or ,

11'-7" in air are considered along with a combination of vertical

and horizontal drops. These drops were analysed to bound a drop in t any orientation. For these cases, the structural integrity of the i poison components must be maintained and the canister must remain i subcritical. Deformation of the canister is acceptable. Although ,

! not expected based on the BW drop test results, leakage of core 1 material from the canister, up to its full contents, is allowed j i provided that the contents left in the canisters remain

! subcritical. An equivalent drop in air was calculated for the worst  !

j case cad this equivalent air drop was used as the basis for the i 1 structural analysis. Structural analysis methods were used to l determine the extent of the deformation of the shell and canister l l internals. Impact velocities were calculated for the specified

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canister drops. Based on these velocities, strain energy methods were used to compute the impact loads associated with the various ,

i- postulated drops. Vector combinations of the horisontal and vertical components were used to determine the effect of a drop at i any orientation.

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) In the vertical drop cases (reference 10), the same deformation will ,

p occur regardless of the canister type, since it is shell dependent.

j Test results from the actual canister drops have verified that for j the bottom impact, all deformation occurs below the lower support plate in the lower head region. An upper bound shell deformation  ;

I was computed using the ANSYS (Reference 5) computer code and the l

results are presented in Figure 3.1-1 along with the actual test

results. l ,

To determine the consequences of a vertical and horisontal drop on j the filter and knockout canisters, their internals were analysed  ;

with finite element' methods using the ANSYS computer program. This  ;

analysis incorporated the actual non-linear properties of the esterial. Geometric constraints imposed by the shell were accounted I for by liatting the displacement of the supports.

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1 15737-2-G03-114-i i1 In the filter canister, criticality control is provided by the l central B4C poison rod coupled with the mass of steel in the i filter element drain tubes and tie rods. ' Using the end caps of the s

- filter modules as deflection limiters, the entire tube array deflection is limited to 1.6" under postulated accidents. This l

l i analysis is conservative because it does not take into account the 5 l circumferential bands around the array or the viscosity of the <

filter cake bed, both of which would tend to maintain the standard spacing. Using the maximum calculated deformed geometry (before the l array bounced back closer to its original position), the criticality criterion given in section 3.2 was set.

t In the knockout canister, criticality control is provided by the i _ central B4 C poison rod coupled with four absorber rods. Results from the structural analysis show that the poison rods remain I

essentially elastic during all postulated accidents and the maximum . ,

instantaneous displacements are less than 0.75 inch. As in the case s 3

of the filter canister, the resultant deformed geometry successfully met the criticality criterion given in section 3.2.

t 4 The fuel canisters, with their square-within-a circle geometry, exhibit. different drop behavior than the_ other canisters. For both the verrical and side drops, the fuel canister internals will not experience significant deformations other than the shell ,

j deformations discussed above. Lightweight concrete filling the void between the square inner shroud and the circular outer shell

, provides continous lateral support to both the outer shell and the ,

shroud. This results in a distributed loading function for i

horizontal drops resulting in no calculated deformation to the shroud shape. Testing has demonstrated that the lower support plate j

,. remains in place for design drops while supporting a mass equal to s j the shroud, payload and the concrete. The lack of significant

deformation af ter a drop (reference 10) makes the criticality l analysis for the standard design applicable to the drop cases as well.

1

+ 3.2 Canister criticality Evaluation Criticality calculations were performed to' ensure that individual canisters as well as an array of canisters will remain below the established keff criterion under normal and faulted conditions.

The criticality safety criterion established is that no single canister or array of canisters shall have a k gf greater than 0.95 1 during normal handling and storage at the TMI-2 site. For plant accidents (e.g., drained spent fuel pool), the criticality safety 4 criterion established is a k,gg < 0.99. These criteria are

l. satisfied for all canister configurations. ,

The computer codes used in this work were NULIF, NITAWL, XSDRNPM and ,

j KEN 01V (References 6, 7, 8 and 9). The NULIF code was used primarily for fuel optimization studies in a 111 energy group -

i representation. NITAWL and XSDRNPM were used _ for processing cross sections from the 123 group AMPX master cross section library.

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y.  !

'NITAWL provides the resonance treatment and formats the cross

! section for use by either XSDRNPM or KEN 0lV. In most cases, XSDRNPM

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-cell weighted cross sections were used in the KEN 01V calculations but for some comparative fuel optimization runs the NITAWL output

,  ;

• library was used directly by KEN 0lV.

g.t

-1 The calculational models assume the following conditions for the canister contents: ,

.1. Batch 3 fresh fuel'only ,

2. Enrichment: batch 3 average + 20 (highest core enrichment)

3. No cladding or core structural material j 4. No soluble poison or' control material from the core

5. Credible fuel size land optimal volume fraction and moderator

, s density w r x l

I 5,s Canister fuel regions are completely filled without weight >

1 l restrictions

7. Uniform 500F temperature i

8. B-10 surface density was assumed to be 0.040 ga/cm2 in the

Boral used for the fuel canister. (Actual B-10 surface density 4 will be 0.040 ga/cm2 with a 95/95% confidence level in the testing to provide at least a 2o margin.)

j ,

9. B4 C density used is the poison tubes for the filter and 1 i

i knockout canister was assumed to be 1.35 ga/cm3 with the boron

! weightpercentassugedtobe70%. (Actual B4 C density will be l

{ at least 1.38 ga/cm with a boron weight percent meeting requirements for ASTM-C-750 Type 2 B 4C powder, minimum boron weight percent 73%.)

I i Optimization studies were performed to determine the value of these j=  ; parameters.. These optimization studies are presented in Reference 1 ,

l 'along with other parametric studies performed for special cases.  !

The XENO analysis employs a fuel model that bounds all debris loading'conffgurations. Three basic configurations were analyzed

'1 for each canisters a single canister surrounded by water, an array

. ,of canisters in the storage pool and a disrupted canister model'

{ resulting from an enveloping drop. The standard canister configuration assumed that some minimum degree of damage could have i i t.' ,

occurred in the canisters during normal loading operations. All the

[

s

' canisters analyzed in an array were assumed to have this minimus damage. A 17.'3" center-to-center spacing was analyzed for the array

{

!- , cases. .The 17.3" center-to-center spacing accounts for all storage rack tolerances and is the minimum center-to-center spacing possible s

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15737-2-G03-114 l 1

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a for any two canisters. The canisters are assumed to be loaded with debris consisting of whole fuel pellets enriched to 2.98 w/o, optimally moderated with 500F unborated water. This provides the g

most reactive fuel configuration possible for the canisters. Thus, the analysis will provide conservative results and bound any actual configuration including draining of the canisters during the dewatering operation. For accident conditions, it is assumed that optimized fuel is present in both normal fuel locations and in all void regions internal to the canister. Filling all void regions with fuel has the effect of adding fuel to the canister after a drop.

The canister shell, including the lower head, is identical for all three canisters. The cylindrical shell is modelled using the maximum shell OD of 14.093" and the nominal 0.25" wall thickness.

The model explicitly describes the concave inner surface but squares off the rounded corners. This increases the volume of the lower j head.

All three canisters contain catalytic material for hydrogen recombination in both the lower and upper head. This material and i its structural supports are not included in the models. The volume i occupied by these materials is replaced with fuel. In addition, the

protective skirt and nozzles on the upper canister head are not

modelled.

The storage rack cases assume the canisters are stored in unborated water with a 17.3" minimum center-to-center spacing. Sensitivity l studies were performed on the nominal 18" center to center spacing to determine the effect of a canister dropped outside of the rack.

These analysis show that keff < 0.95 for canisters dropped l outside the rack as long as the side of the dropped canister does not come within 2" of the side of the nearest canister in the rack.

Thio requirement is met by the storage rack design (Reference 2).

i Three cases are examined for a dropped canister: a vertical drop,'a horizontal drop and a combined vertical and horizontal drop. The i shell deformation is essentially the same for all cases. For these drops, the cylindrical shell is assumed not to deform. Any

] deviation from the cylindrical shape would increase the surface to j volume ratio and increase the neutron leakage from the system. In the lower head region of the shell, a tear drop shape expansion is assumed to occur. The bottom-head is modelled as a flat plate with the internal components resting on it. To bound all drop cases, the canister was assumed to rotate during a drop and land on its head.

, A similar tear drop shape will result. Both of these cases were merged into a single model that assumes the tear drop deformation at both the top and bottom with the internals displaced to the i,

flattened lower head surface. For the combined vertical-horizontal drop, the radial displacement of the internal components is combined i with the double tear drop model. This drop model bounds any

conceivable drop configuration by exceeding conservative stress estimates of deformation.

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t Results f

l The results of KENO, using basic three dimensional canister models are presented in Table 3-1. These results represent bounding values for any

configuration of the canisters at 21-2.

Basically, they show that for any configuration, the effective-

multiplication factor, with uncertainties included, will be less than i

0.95. Due to the conservatism built into the models, the k.fg of any actual configuration will be less than these bounding values.

i Three assumptions used in the analyses reported in Table 3-1 have been i

reevaluated. The affected assumptions aret

1. type of poison used in the filter and knockout canistert., j

] 2. storage pool water temperature, and i

3. fuel particle size. t i ,

The values reported in Table 3-1 for the filter and knockout canisters j .are based on the assumption that the poison tubes for the canisters are l

filled with vibrapacked B4 C powder. Actual fabricated filter and i knockout canisters contain compressed sintered B4 C pellets. This L change resulted in a small reduction to the diameter of the poison in the canisters which results in a small increase in the multiplication value ,

l (kegg) of the'two canister types. Based on analyses the increase in l i multiplication will not exceed 0.'4% ak.

l The values reported in Table 3-1 assume a minimum temperature of 500F -

f for all canister types. For canisters stored in the spent fuel pool the

temperature could be as low as 320F. Explicit criticality array t j calculations were not performed at this lower temperature. Rather, an

evaluation was performed to determine the maximum increase in

! multiplication due to cooling from 500F to 320F. The maximum change I in multiplication was determined to be an increase of 0.1%'ak.  !

! The results reported in Table 3-1 are also based on the assumption that j no single fuel mass greater than a whole fuel pellet exists in the THI-2 4

core. Examinations of the core have indicated that fuel melting any have 9

occurred. To assess the impact of this possibility an evaluation was ,

. performed to determine the k for the most reactive batch 3 fuel particle

! size. The k for the large particle was only 0.07% Ak higher than the k,, for the standard whole pellet.

l j In conclusion, the changes in k,gg resulting from the three modified  ;

assumptions.wlli not result in exceeding the k.gg criteria of 0.95 for t t the< cases reported in Table 3-1.

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3.3 Canister Hydrogen Control Evaluation l 1  !

! A generic feature of the canisters is the recombiner catalyst package incorporated into the upper and lower heads of all the ,

j canisters. The catalyst recombines the hydrogen and oxygen gases '

formed by radiolytic decomposition of the water trapped in the damp debris. This reduces the buildup of internal pressure in the i canister and keeps the gases below the flammability limit. The L

{ redundant locations ensure that an adequate amount of catalyst is

available for any canister orientation in which hydrogen might be generated (e.g., an accident which leaves a canister upside down).

! Test results (Reference 4) have shown that the catalyst will perform l

effectively when dripping wet, but not when submerged.

li A total of 200 grams of catalyst is initially installed in each canister. Then extra catalyst is installed in the beds to fill remaining volds. The 200 gran quantity was determined from the

! catalyst tests run by RHO (Reference 4) which used 100 grams and a l H2 /02 generator which simulated the maximum gas generation

< stated in the report of 0.076 liter /hr hydrogen. Additionally, the j beds were designed'to meet the shape and volume requirements l established by the tested catalyst beds. A total of at least 200 j grams of catalyst is installed in the canister in order to be  !

assured that at least 100 grams is above the maximum water level for i all canister orientations. At least 100 grams of catalyst is at

} either end of the canister and the bed arrangement at each end is j syneetrical.

i The maximum predicted gas generation rate in a canister has been determined by two separate models; (1) the maximum theoretical gas l generation rate t.nd (2) the maximum realistic gas generation rate. i The maximum theoretical gas generation rate was determined by Rockwell Hanford Operations (RHO) in their document RHO-WM-EV-7

, (GEND-051) for purpose of developing the catalytic recombiner bed j design. The maximum realistic gas generation rates were determined j by GPU for purposes of predicting canister internal pressures during i periods when the canisters are water solid.

1

! .Both models are based on the Turner paper, "Radiolytic Decomposition .f 1 of Water in Water-Moderated Reactors Under Accident Conditions", t i referenced in the RHO report. The basic relationship is: ,

i H2 = (W)(F)(G)(r) 8.4 x 10-3 liters / hour '

-l where:

F = fraction of Y and 6 energy absorbed in water e j G=H2 generation value in moles /100 eV l t r = ratio of peak to average decay heat energy in the fuel debris W = ionizing radiation per canister (watts) r

! '8.4 x 10-3 = unit conversions (L ev/W.hr)

I l

Rev. I

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15737-2-G03-114 For the maximum theoretical generation, the above factors are maximized as follows:

o W - the maximum quantity of fuel debris in any canister, not including residual water weight or weighing accuracy, is assumed. (W = 54.2) o F - The fraction of Y and 8 energy absorbed is conservatively high and large amounts of water are also assumed to be available for absorbtion which is in excess of what is possible in the canisters. (F = 0.2) o G - The hydrogen gas generation value is based on a) completely curbulent/ boiling conditions when the radiolytic gases are instantly removed from the generation site and b) no build up of hydrogen overpressure which tends to retard radiolysis. (G

= 0.44) o r - The ratio of peak-to-average decay heat energy in the fuel is based on the most active region of an undamaged core. This assumes the fuel is intact and not scattered to other regions. (r = 1.9)

For the maximum realistic generation of hydrogen and oxygen, the worst case realistic factors for the damaged TMI core are used as follows:

o W - The maximum quantity of fuel debris expected in any canister is used which includes allowances for residual water and weighing accuracy. (W = 50) o F - The fraction of Y and 8 energy absorbed is based on the maximum amount of water possible in an actual canister.

(F = 0.07) o G - The hydrogen gas generation value is based on the actual worst case core debris conditions expected in a canister which includes lower temperature, quiescent conditions.

(G = 0.12) o r - The ratio of peak to average decay heat energy in the fuel debris is based on the worst case conditions in the damaged TMI core. (r = 1.4)

The resulting hydrogen / oxygen generation rates for the two models are:

Max. Theoretical Max. Realistic I liter / hour liter / hour H2 7.6 x 10-2 5.0 x 10-3 02 3.8 x 10-2 2.5 x 10-3 Total 1.14 x 10-1 7.5 x 10-3 Rev. 1 0334Y aw .

o 15737-2-G03-114 The 6eneration of other gases was not considered. Since the amount of contaminants in the RCS is small, the generation of other gases from the radiolytic decomposition of these contaminants is not expected to be significant.

Using the maximum realistic gas generation rate of 0.0075 liters / hour and assuming no recombination or scavenging of oxygen, the 25 psig relief valve is estimated to first open in approximately 25 days for the worst case canister. Released gas will be vented through the pool water directly to the containment or fuel handling building and is such a small quantity that it will cause no combustion concerns in the atmosphere of these buildings.

To address the issue of canister pressurization resulting from failure of the 25 psig relief valve a second relief valve is installed on the canisters. This relief valve will ensure that canister pressure does not exceed the design limit of 150 psig. The additicnal relief valve will make the canister single failure proof with regards to pressurization. This second valve will also be installed in such a manner to eliminate common mode failure of the two pressure relief valves.

The recombiner catalyst is ineffective when it is under water. An evaluation has been performed to determine how long it takes an undewatered canister to reach 150 psig if the 25 psig relief valve fails closed. This time for the worst case canister is 139 days. A similar concern exists for the dewatered canister should a signficiant amount of oxygen scavenging occur and the 25 psig relief valve fails closed. Assuming no recombination, (i.e. complete oxygen scavenging) the canister will reach the design pressure in 4286 days for the worst case canister.

If the relief valve should fail open while the canisters are being stored chere is the possibility that fuel debris can be released into the pool water. If contaminants are released into the pool the defueling water cleanup system (DWCS) can be used as necessary to limit the contamination level of the water. Hence, a failed open relief valve does not pose a safety concern. Additionally, given that it is planned, although not required, to dewater the canisters shortly after they are loaded, pressurization of the canisters caused by hydrogen / oxygen generation will be minimal and the relief valve is not expected to open.

Although not considered a credible event, the consequences of a hydrogen ignition inside a canister has been evaluated. The maximum pressure that can be reached inside a canister under normal conditions, because of the 25 psig relief valve, is approximately 42 paia. This pressure includes the 25 psig set pressure and 5 feet of water submergence. Under the assumption that the recombiner catalyst does not function properly, a flammable mixture of hydrogen and oxygen can accumulate within a canister. If an ignition of this mixture is postulated, an overpressurization of the canister could occur. The ultimate stresses will be reached for various canister components at the estimated pressures:

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15737-2-G03-114 2

o canister shell - 2160 poi o fuel canister bolts - 2900 psi o threaded connections - 2500 psi Considering the large margin that exists between these pressures and the maximum, normal condition canister pressure (i.e., approximately a factor of 50), the overpressurization resulting from an ignition of hydrogen within the canister is not expected to affect the overall canister integrity.

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15737-2-G03-114 Table 3-1 Results of 3D KENO Criticality Calculation Description k ff+2a Histories Maximus k f g*

Filter Canister ** . l Single, Ruptured Filters 0.795 j; 0.024 9331 0.839 17.3" Array, Ruptured Filters 0.823 j; 0.021 52374 0.867 Vertical Drop, Ruptured, without filter screens 0.798 j; 0.025 8127 0.843 Horizontal Drop, Ruptured, without screens 0.843 j; 0.010 15050 0.873 Combined Horizontal / Vertical Drop, Ruptured, without screens 0.851 j; 0.021 44849 0.892 Fuel Canister Single, Standard Configuration 0.825 j; 0.012 15050 0.857 17.3" Array, Standard Configuration 0.829 + 0.025 6321 0.877 Knockout Canister ** l Single, Standard Configuration 0.835 j; 0.018 10535 0.873 17.3" Array, Standard Configuration U.877 + 0.015 11438 0.915 Vertical Drop, Single 0.843 j; 0.019 9933 0.882 Horizontal Drop, Single 0.853 + 0.008 26488 0.881 Combined Horizontal /Veritical Drop, Single 0.851 j; 0.016 12943 0.887

• k gg + 20 t calculational bias (see Reference 1)

• • results are based on vibrapackel 34C powder in the poison tubes l I

\

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15737-2-G03-114 I

Figure 3.1-1 SHELL DEFORMATIONS - VERTICAL DROP (ALL CANISTERS)_

.M-I PREDICTED

/ \ DEFORMED

/ \

SHAPE

/ SHAPE BETORE TEST b \

I i \ \

l h \

I h

.l*/

i I

'f f -

\l\, ,

/ ///

/ .

AFTER TEST

-- / (DEFORMED SHAPE)

.o > .2 ACTUAL 4y .6 PREDICTED g,y, g

15737-2-G03-114 4.0 Radiological Considerations The canisters are designed to be loaded with core debris from the TMI-2 RCS. These canisters do not contain internal shielding and must be shielded during all handling and storage operations.

The shielding requirements for the various canister operations (e.g.

loading, handling, and storage) are discussed in reference 3.

Personnel exposure from the loaded canisters will be addressed in Reference 3 as part of the canister handling sequence.

Rev.1 l 0334Y

15737-2-G03-114 5.0 10CFR 50.59 Evaluation Changes, Tests and Experiments, 10CFR 50, paragraph 50.59, permits the holder of an operating license to make changes to the facility or perform a test or experiment, provided the change, test or experiment is determined not to be an unreviewed safety question and does not involve a modification of the plant technical specifications. A proposed change involves an unreviewed safety question ift a) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the saiety analysis report may be increased; or b) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or l

c) the margin of safety, as defined in the basis for any technical specification, is reduced.

The defueling canisters replace the fuel cladding lost during the l

accident as the barrier for containing the fuel. As discussed in Section l

1.1 of this TER, the purpose of this evaluation ir to show that the canisters are designed to remain safe under normal operation and handling conditions as well as postulated drop accidents and storage. The scope of the evaluation relates only to design aspects and not in field canister use which is addrer ed in the Safety Evaluation Report for Early Defueling of the THI-2 Reactor Vessel (Reference 3). On this basis the l

scope of this 10 CFR 50.59 Evaluation is limited to design aspects of the l canister.

The issues of concern with canister design are criticality control and overpressurization protection. With respect to criticality control, this evaluation shows that the canister will remain subcritical under any configuration or following structural deformation due to a load drop.

With respect to overpressurization protection, two relief valves will be l l

installed on each canister to prevent the possibility of a single failure or common mode failure from overpressurizing the canister. Thus, it can be concluded that the design of the defueling canisters neither increases the probability of any accident previously evaluated nor creates the possibility of a different type of accident. Additionally, as the current THI-2 Technical Specifications do not specifically address containment of the fuel debris, the margin of safety as defined in the basis of the Technical Specifications is not reduced.

l As discussed above, these canisters are critically safe by design.

Additionally, activities associated with canister closure and handling, including installation of the relief devices, will be performed in accordance with procedures prepared, reviewed and approved in accordance with TH1-2 Technical Specifications Section 6.8, which requires NRC

( approval of certain types of procedures. Therefore, as no further l

engineering controls are needed to ensure criticality safety and i

activities associated with canister closure and handling will be l

controlled in accordance with procedures subject to Technical l Specification Section 6.8, it is GPU Nuclear's belief that no changes to

! the Technical Specifications are required.

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15737-2-G03-114 l i

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t In conclusion, within the bounds described in this report, the design and l use of the defueling canisters do not result in an unreviewed safety question, nor require chan&es to the TMI-2 Technical Specifications.

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- , , - - , , - , - - - - - , rn,-, --.. ,--r--,4 , - - - e._ ,,,_ .- ,-+-----y. . , , , - - - ~ , - - , - --_n-..

15737-2-G03-114 6.0 Conclusions Canisters are needed to provide effective long tera storage for the THI-2 core debris. Three types of canisters are required to support the defueling systems fuel, filter and knockout canisters. These canisters have been evaluated to determine if they could safely perfore their function under normal and accident conditions. The results of this evaluation show that the canisters will remain subcritical under normal operations, handling and accident conditions. A structural evaluation of the canisters has shown that they maintain their integrity and will function as designed under normal operating conditions. Drop analyses and drop tests were used to determine the effect of a design basis drop on the canister shell and internals. The results from these analyses were used in determining the reactivity of the canisters under accident conditions. Therefore, based on structural and criticality considerations, it can be concluded that these canisters can safely function under normal and accident conditions at TMI-2.

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15737-2-G03-114 l

. t l .

7.0 References l 1. TMI-2 Defueling Canisters Final Design Technical Report, Babcock and i Wilcox, Document No. 77-1153937-04, May 24, 1985.

l

2. Technical Evaluation Report for Fuel Canister Storage Racks,  !

15737-2-G03-113, Rev. O. i l

3. Safety Evaluation Report for Early Defueling of the TMI-2 Reactor Vessel, 15737-2-G07-107.

4. Evaluation of Special Safety Issues Associated with Handling the TMI-2 Core Debris, RHO-WM-EV-7, Rockwell Hanford Operations, l February 1985. l l

S. Computer Code "ANSYS" Revision 4.1, March 1, 1983, Swanson Analysis l System Inc., Houston, PA.  !

6. "NULIF-Neutron Spectrum Generator, Few Group Constant Calculator and l Fuel Depletion Code", RAW-426, Rev. 5.

7. "NITAWL, Nordheim Integral Treatment and Working Library

• Production," NPGD-TM-505.

l 8. "XSDRNPM AMPX Module with One Dimensional a S Capability for l

Spatial Weighting," AMPX-11, RSIC-RSP-63, OKNL.

9. " KEN 04, An Improved Monte Carlo Criticality Prograa," NPCD-TM-503, '

Rev. B.

10. TMI-2 Drop Testing of Defueling Canisters Final Report, Babcock and Wilcox, Document No. 77-1156372-00, February 1985. .

l

11. TM1-2 Early Defueling Fines / Debris Vacuus Systen Proof-of-Principle Test Report, TM1-AD-84-018 Westinghouse Electric Corporation, Advanced Energy Systems Division, October 1984. [

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15737-2-G03-114 Attachment 1 THI-2 Transfer System Criticality Technical Report Rev. 1 0334Y