NUREG/CR-4517

PN L-5321

Design Features for Enhancing International Safeguards of

Away-from- Reactor Dry Storage for

Spent LWR Fuel

Prepared by F. P. Roberts, N. L. Harms

Pacific Northwest Laboratory Operated by Battelle Memorial Institute

Prepared for

U.S. Nuclear Regulatory Cornmission
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NUREG/CR-4517

PN L-5321

Design Features for Enhancing International Safeguards of

Away-f romn-Reactor Dry Storage for

Spent LWR Fuel

Manuscript Completed: February 1986

Date Published: October 1986

Prepared by

F. P. Roberts, N. L. Harms

Pacific Northwest Laboratory

Richland, WA 99352

Prepared for

Division of Safeguards Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission

Washington., D.C. 205655

NRC FIN B2475

ABSTRACT

The Pacific Northwest Laboratory has performed a study for the Nuclear

Regulatory Commission to identify and analyze design features that can facilitate the implementation of IAEA safeguards at away-from-reactor facilities for

dry storage of light water reactor spent fuels. Thirteen specific design

features are identified that can enhance verification of nuclear material flow

and inventory. These are assessed from the viewpoint of safeguards effectiveness and possible impacts on the IAEA and the operator of the AFR facility.

ii i

CONTENTS

1.0 INTRODUCTION ................... **** ...... ... ......... .. . . ... . 1

2.0 SUMMARY AND CONCLUSIONS . .. . .... . .. .. . ... . .. .. * ........... o.. . ... s . .. 3

3.0 BASELINE ASSUMPTIONS ........ .. ...... ........ *... o .. ...... .. ... .. 5

3.1 SAFEGUARDS OBJECTIVES ...................... s** ........ ... o.....s 5

3.2 SPENT FUEL RECEIVING .... . ... ...... . ... .. .... *........ . w .... so.** 5

3.3 MATERIAL FLOW FOR BASELINE FACILITY ............................. .7

3.4 OPERATIONS IN FUEL HANDLING HOT CELL ............................ o 7

3.5 LOADOUT AND STORAGE OF FUEL ..................... o..... ooosol ..0..0 10

4.0 BASELINE SAFEGUARDS APPROACH .. oo..o .o..oe.. . .oo. .ooooo...o....o. ooo. 11

4.1 RECORDS AND REPORTS ....... o...... 11

4.2 VERIFYING FLOW AND INVENTORY .. ... oo.o.o...o..................... 11

5.0 BASELINE SAFEGUARDS SYSTEM EVALUATION .......... o ............ 15

5.2 INSPECTION MANPOWER FOR FLOW VERIFICATION ............ o... o........ 16

5.3 INSPECTION MANPOWER FOR PHYSICAL INVENTORY VERIFICATION . .... oo. 16

5.4 TOTAL INSPECTION EFFORT .... .... ..... ...... o.........sooosooo .... 16

6.0 DESIGN FEATURES FOR AFR STORAGE FACILITIES .. o.o.oooo.ooo.oo.oc....... 17

6.1 SPECIFIC DESIGN FEATURES . .. . . . .. . .. . o. .. . . .. .. . .. .. .. . .. .. 17

6.2 GENERAL FACILITY CHARACTERISTICS o.o..... oooosooo................ 21

V
FIGURES

1 Material Flow Diagram: Transport Cask Handling Operations ............ 8

2 Material Flow Diagram: Spent Fuel Handling ........... . . 9

3 Safeguards Strategic Points at the Baseline AFR Dry

Storage Facility ..... ooo...... o... o.......o............... . oo........... 13

TABLES

1 Detection Goals .. o..o......o.........o........o................... ooo 5

2 Physical Characteristics of Some Spent Fuel Shipping Casks ......... 6

3 Characteristics of Typical LWR U02 Fuels . ... o.o...... ..... ....... 0.0 6,

4 Material Flow at the Baseline Facility o..o.o.ooooooooooo..ooooooooo.. 12

vi
EXECUTIVE SUMMARY

Including safeguards considerations in the early design phases of nuclear

facilities can have substantial beneficial impacts on later application of IAEA

safeguards. The study described in this report is the fifth of a series of

studies carried out for the Nuclear Regulatory Commission's Office of Nuclear

Materials Safety and Safeguards. The studies all have the specific objective

of identifying and analyzing design features for nuclear facilities that can

facilitate the implementation of IAEA safeguards. The previously reported

studies involved light water reactors, spent fuel reprocessing plants, mixedoxide fuel fabrication plants, and low enriched uranium conversion and fuel

fabrication plants. This study involves away-from-reactor dry storage facilities for light water reactor spent fuels.

The baseline facility for the study is a 5,000 metric-ton dry storage

facility. Spent fuel is received from pressurized water or boiling water

reactors both as bare intact assemblies and as canistered, consolidated fuel

rods. The spent fuel is received into a fuel receiving and handling buildi~ng

where it is inspected, packaged in sealed metal canisters, and loaded into

concrete-shielded storage casks. The casks are welded closed and placed on

concrete pads in a storage area. Ultimately the spent fuel is retrieved and

shipped to a reprocessing plant or to a disposal site.

The assumed safeguards approach consists of item accountancy heavily

supported by containment and surveillance (C/S) measures.

An analysis of the baseline facility and the safeguards approach identified areas that can be improved by facility designs. The following specific

design features were identified and evaluated:

" Provide quarantine area in the fuel handling cell for storing spent

fuel receipts after unloading from the transport casks. The area

would be protected by C/S measures to prevent undetected passback of

material into the casks.

* Design the equipment location and processes used for spent fuel' rod

consolidation and the canisterization operations to permit them to be

viewed and recorded on video tape with IAEA CCTV systems.

" Install equipment and processes to photograph weld beads on canister

covers.

" Install a station for accurately weighing spent fuel canisters after

they are prepared for loading into the storage casks.

" Install NDA capability to quantitatively measure the plutonium content in spent fuel receipts and canisters prepared for loading in the

storage casks.

vii
" Install NDA equipment to qualitatively measure the gamma ray emissions from spent fuel temporarily stored in the fuel handling cell as

well as canisters prepared-for loading into storage casks for the

presence,'of gamma radiation.

" Design the cask handling and spent fuel handling operations to be

performed remotely from a central control room provided with I.AEA

controlled CCTV monitors and video recorders.

" Subdivide the storage area into sub-areas by barriers that cannot be

crossed by the storage cask transporter except at gateways that can

be sealed using IAEA seals after the sub-area is filled.

" Design the perimeter of the storage area'so that the storage cask

transporter can only enter through the storage cask loading room in,

the receiving and handling building or through infrequently used

port'als. that can be sealed with JAFA seals.

*Design casks to allow use of fibre optic seals with remote electronic

verification capability.

" Design storage casks to be sealed to the concrete pads on which they

are placed in the storage area.

" Provide temperature sensors that can monitor the storage cask

interiors.

" Incorporate a collimator in the storage cask shielding wall'to permit

gamma energy analysis of the contained spent fuel.

In addition to the specific design features listed above, some general

characteristics should be incorporated in the facility. These i~nclude, *for

example, the following:

*adequate lighting, in the operating areas for the IAEA surveillance

cameras

*structural designs and plant layout that minimize the number of

access portals from which spent fuel could be removed from the

facility, thereby reducing the demand on IAEA C/S systems,

" designs that minimize interferences with the surveillance cameras and

reduce the 'frequency of anomalies resulting from restricting the view

of the cameras

" design of shipping and storage casks that allow the placement of IAEA

.seals with minimum effort and ensure that violations of containment

are detectable.

viii
1.0 INTRODUCTION

The objective of the work described in this report was to identify and

analyze design features that can influence the effectiveness and efficiency of

IAEA safeguards inspections at away-from-reactor (AFR) dry storage facilities.

The goals of the work were to present specific examples of design features that

facilitate IAEA inspections, and to provide some perspectives on ,solutions to

anticipated problems in verifying nuclear material flows and inve'ntories at AFR

dry-storage facilities.

The level of effort required and the degree of effectiveness of IAEA

safeguards in detecting diversion of nuclear material can be affected by the

facility designs. These designs include material handling procedures, layout

of equipment, process And storage areas, physical and radiation barriers, and

degree of automation of the process. The U.S. Interagency Action Plan Working

Group for Strengthening IAEA Safeguards recognized this need by recomme 'nding

development of facility design features that could improve the efficiency

and/or the effectiveness of the IAEA's inspection and verification efforts.

Dry storage of spent LWR fuel away from reactors was recognized to have spe-'

cial safeguards problems and was included in the list of facility types to be

investigated.

Because of the worldwide delays in reprocessing, it is becoming increasingly important to expand the capacity for storing spent. LWR fuel. Options

available for increasing at-reactor (AR) capacity include installing highdensity storage racks, double tiering, disassembling and consolidating fuel

rods, expanding the size of the storage pools, or using dry storage at the

reactor site.

Separate AFR storage is a viable alternative to AR storage. Delicated

storage facilities such as the Morris Operation in the U.S. and CLAB in Sweden

represent wet storage facilities that are in use today. Dry storage is a technically attractive approach to interim storage of LWR spent fuel. The monitored retrievable storage (MRS) approach being developed in the U.S. is based

on dry, passively cooled storage modes. One operating AFR dry storage for LWR

spent fuels is the 1,500-metric-ton facility at Gorleben in Lower Saxony, FRG.

This facility and a similar one being constructed at Ahaus in North Rhine-.* Westphalia,' FRG, 2store LWR fuel assemblies in steel casks that are also used as

transport casks.2 Other storage modes are at various stages of devel'opment for

LWR fuels as well as for gas-cooled'reactor and heavy water reactor fuels.

These include, for example, emplacing spent fuel in air-cooled vaults,, in

surface dry wells, and in silos or casks placed on the surface in open fields.

The methodology used in evaluating design 4features for the selected

facility type was developed in earlier work. I The methodology is summarized

only briefly here. The principal steps comprising the methodology are the

following: 1) define the baseline facility and safeguards system, 2) evaluate

the baseline system to identify safeguards problem areas, and 3) identify

1
potential design features for alleviating the safeguards problems, and after

screening out those that are impractical, determine the impact of incorporating

the remaining design features in the facility design.

2
2.-0! SUMMARY AND CONCLUSIONS

Thirteen design features were identified that have potential for enhancing

IAEA capability for verifying nuclear material flow and inventory at future AFR

spent fuel storage facilities. In some cases, the design features represent

different approaches for resolving the same safeguards issue, and it is not

expected that they all would be deployed. The design features. address 1) ways

to minimize demands onIAEA manpower and equipment and 2) ways to resolve those

verification issues peculiar to AFR storage facilities that perform fuel handling and packing operations, e.g., frequent shipments to and from the facility, fuel bundle disassembly, and long-term storage of large inventories.

It was concluded that incorporation of design features can substantially

reduce the cost of flow verification by providing ways to~reduce or eliminate

inspector presence without increasing the level of intrusiveness on the facility operator. Alternatives for enhancing inventory verification efficiency and

methods for resolving anomalies in the event of failures of containment or surveillance measures were identified.

The issues addressed in this paper concern facilities that may be built in

the future. By taking the features that effect safeguards into consideration

early in the design stages, it is possible to circumvent problems in verifying

nuclear material at AFR storage sites.

3


3.0 BASELINE ASSUMPTIONS

The baseline assumptions used in this study include a set of technical

objectives for safeguarding an AFR LWR spent fuel dry storage facility,

descriptions of the baseline facility including the nuclear material, buildings

and equipment, and the safeguards approach needed to attain the safeguards

technical objectives.

3.1 SAFEGUARDS OBJECTIVES

The safeguards technical objectives can be translated into detection goals

by quantitatively defining significant quantity, detection timeligess, and

probability of detecting diversion. The IAEA Safeguards Glossary provides

guidelines for the detection goals. Table 1 lists the values that are assumed

for this study.

The values suggested in the IAEA Safeguards Glossary 5 for detection

probabilities are 90% to 95%, With a false alarm probability of 5% or less.

The values assumed for this study are 90% and 5%, respectively. Because no

capability exists in the baseline facility for uranium-plutonium separation,

the detection goal quantity is fixed by the plutonium content in a fuel assembly or package of consolidated fuel rods. On this basis, a goal quantity is

2 PWR assemblies or 5 BWR assemblies for intact fuel assemblies. For consolidated fuel rods, a canister typically contains the equivalent of 2 assemblies,

thus 1 canister of PWR rods or 2 canisters of BWR rods constitute a goal

quantity.

3.2 SPENT FUEL RECEIVING

ý.The facility receives spent fuel as intact assemblies or as canister ed

fuel rods from rod consolidation processes tarried out at the reactor site.

Capability for retrieving from storage and shipping from the facility is provided. Spent fuel is received during the first 20 years of operation and shipments are carried out during the last 5 years of operation.

TABLE 1. Detection Goals

Si gni f icant

Quantity Detection Time

Plutonium in 8 kg (total element) 1 to 3 months

irradiated

fuel

U-235 in uranium 75 kg (total isotope) 1 year

enriched <20%

5
The spent,' fuel will be transported in massive, heavily shi'elded casks from

the reactor sites. Transportation technology for spent fuel is well developed,

and a number of rail, sea, or truck shipping casks have been fabricated and are

in use today, and others are in the design stage. Characteristics of some of

these casks are listed in Table 2 to illustrate typical capacities and physical

sizes and masses of the casks that might be used for shipping spent fuel to and

from AFR storage sites.

Only LWR U02 fuels (not mixed-oxide fuels) are considered in the study.

Characteristics of the LWR fuels vary widely in dimensions, mass, burnup, cooling times and other factors. Table 3 describes-the spent fuel characteristics

that are assumed for' th is study.

TABLE 2. Phys-ical Characteristics of Some Spent Fuel Shipping Casks

Cask

IF-300

NLI-1/24

TN- 12

NFS-4

TN-S.

TN-9

Transport

Mode

Rail

Rail

RailI.

Truck

,Truck/rail

Truck/rail

Capacity

Assemblies

WR BWR

7 18

10 24

12 32

SWei ght,

metric tons

63

88

106

23

36

36

OverallI

Dimensions, m

Dia. Length

1.63 5.28

ý2.63 5.69

2.49 5.41

1.27 5.44

1.73 5.59

1.73 5.84

1

3

2

7

.TABLE 3. Characteristics of Typical LWR

Characteristic

Fuel assembly dimens~ions:

length, m

cross section, mm

Assembly weight:

total , kg

uranium, kg

No. of fuel rods per

assembly

Final uranium enrichment:

%U-235

Plutonium content:

kg/assembly

kg/rod

Decay heat,,W/assembly

(10 yr after discharge)

PWR

4.06

210

665

461

264

0.8-1.2

3.8

0.0144

1000

U02 Fuels

BWR

4.47

139

276

180

63

0.8-1.0

1.*4

0.0222.

400

6
3.3 MATERIAL FLOW FOR BASELINE_ FACILITY

Simplified overall material flow diagrams for the baseline facility are

shown in Figures 1 and 2 for the receiving and handling operations. The

receiving and handling operations take place in a single building. The spent

fuel is received in a heavily shielded transport cask, either by rail or truck,

in an enclosed receiving and inspection room where the vehicle and the external

parts of the cask are washed down to remove road dirt. The cask and vehicle

are visually inspected for damage, and surveyed and smear-tested for radioactive contamination. If no damage or contamination are found, the cask covers

and protective devices are~ removed, and the cask is placed on a cart and moved

into the cask handling and decontamina 'tion room. The IAEA seal, if one has

been applied at the reactor, can be checked while in this room.

When the transport cask has been cooled and decontaminated, 'it is moved

into the cask unloading cell. The cask unloading cell is a remotely operated

concrete-shielded cell equipped with viewing windows and closed circuit television (CCTV). This cell is connected to the fuel handling hot cell by a port

to which the shipping cask can be connected and sealed. This port allows the

fuel transfer to be made without spreading radioactive contaminants and with

minimum radiation to the workers. Fuel tran sfers are made by positioning the

transport cask under the transfer port, removing the shielding plug in the

port, removing the inner cask cover, *and lifting the fuel into the cell.

After the fuel is transferred, the interior of the cask is decontaminated

and the inner cask cover is replaced. The shielding plug is replaced in th 'e

port, and the cask is transferred to the cask hand~ling and decontamination room

where the exterior is decontaminated and the outer lid is reattached. The cask

is then moved to the receiving and inspection room where it is placed on the

transport vehicle and prepared for shipment offsite.

All operations required to prepare the spent fuel for dry storage take

place in the fuel handling hot cell. This cell is a heavily shielded, remotely

operated room equipped with viewing windows and CCTV. Fuel handling is accomplished with remotely operated cranes, and master-slave manipulators are used

to perform process and maintenance activities. There is a port for transferring the prepared fuel to the storage cask. This cell is connected via a port

to a maintenance cell overhead where the cranes and process equipment can be

moved for repairs or for decontamination prior to removal from the facility.

3.4 OPERATIONS IN FUEL HANDLING HOT CELL

In the fuel handling hot cell the bare spent fuel assemblies and canisters

are inspected to confirm their identity by the serial numbers placed on the

assemblies by the fuel manufacturer or unique identity markings applied to

canisters by the reactor operator. The bare assemblies are examined visually

for damaged fuel rods, and any assemblies suspected of containing leaking rods

are placed in special canisters. For fuel received in canisters, the canisters

are examined for integrity and placed in overpacks if damaged. The special

can~isters or overpacks are welded shut, and unique identifying numbers are

applied to them.

7
To Spent Fuel Handling

and Storage Operations

Transfer Spent Fuel to

Fuel Handling Hot Cell

Remove Cell Shielding

Plug and Cask Inner Cover

Engage Cask to Transfer

Part of Fuel Handling Hot Cell

Remove Outer Cask Cover

Move Cask into Unloading Area

Cask and

Cart

Load Cask on Transfer Cart

Inspect IAEA Seal

Remove Cask Protective

Covers and Impact

Limiters

Survey for Radioactive

Contamination and

Decontaminate if Needed

Move Cask and Vehicle

into Cask Preparation Area

t Cask and

Vehicle

Cask

Clean Cask Interior

Replace Cask Inner Cover

and Cell Shielding Plug

Dis engage Cask from

Cell Port

Decontaminate Cask

Replace Cask Outer Cover

ilask

Vehicle

Move Cask into Cask

Preparation Area

Load Cask on Vehicle

Replace Protective Covers

and Impact Limiters

Remove Road Dirt from

Cask and Vehicle

Move Cask and Vehicle

into Receiving Area

Receive Spent Fuel

From Reactor

Cask and

Move Cask and Vehicle

into Receiving Area

Prepare Cask .for Shipping

to Reactor

To Reactor

.FIGURE 1. Material Flow Diagram: Transport Cask Handling

Operations

8
Move Canisters to Temporary

Storage Location

Apply Unique Identifying

Number to Canister

Place Fuel Rods in Canister

and Weld Lid

Place Non-Fuel Components

in Canister and Weld Lid

t

Disassemble and Remove

Fuel Rods

Observe and Record Fuel

Assembly Number

Open Canister (if Necessary)

Move Fuel Assemblies to

Rod Consolidation Area

Lmoad Out Non-Fuel

component Canister

Disposal

Opioa

Move Spent Fuel Canister

to Load Out Port

Position Storage Cask in

Load Out Area

Transfer Canisters to

Cask

Weld Cask Cover in Place

Apply Unique Identifying

Number to Cask

Load Storage Cask on

Transporter

Transport Cask to Storage

Location

Lift Cask from Transporter and Place on

Storage Pad

HWeld Canister Cover

Place Bare Intact

Assemblies in Canister

I I

.Test Assemblies and

Canisters for Damage

Verify and Record Assembly

or Canister Identify

Number

Transfer Spent Fuel

Assemblies and Canisters

of Spent Fuel Rods to

Fuel Handling Cell

Spent Fuel

FIGURE 2. Material Flow Diagram: Spent Fuel Handling

9
Capability for fuel rod consolidation is provided in the cell. Equipment

is installed for removal of the assemblies from the canisters, if packaged by

the reactor operator, and for disassembly and removal of the fuel rods. The

fuel rods are placed in canisters, welded shut, and given a unique identifying

number. The consolidation process is performed so that rods from a single

assembly will all be placed in one canister, but rods from more than one assembly may be packed in one canister. The stainless steel and Zircaloy hardware

from the assemblies is compacted, placed in canisters, and removed from the

fuel handling cell for separate storage.

3.5 LOADOUT AND STORAGE OF FUEL

The canistered spent fuel is stored in concrete casks placed on concrete

pads in open fields at the site. Each storage cask holds .10 to 12 PWR assemblies or 25 to 30 BWR assemblies. After packaging operations are completed in

the spent fuel handling cell, the canisters are loaded into 'the storage cask

through the loadout port in the fuel preparation hot cell. The storage cask is

moved to the storage field by the transporter and placed on a storage pad.

Retrieval of the spent fuel and shipment to the reprocessing plant or to

the ultimate disposal site is assumed to involve operations essentially the

reverse of the above procedure.

10
4.0 -BASELINE SAFEGUARDS APPROACH

The s~afeguards approach for the baseline facility is based on a system of

material accountancy heavily complemented by cgntainment and 7surveillance (C/S)

measures, and on descriptions given by Tkharev and Ermakov.7

The material accountancy system is based on item control. The items are

either intact fuel assemblies or canisters of fuel rods. The. nuclear material

content is based on the fuel manufacturer's-recorded uranium element and U-235

content corrected for 1) calculated plutonium production and uranium depletion

in the reactor, and 2) radioactive decay subsequent to discharge from t~he

reactor.

Material, flow verification is further complicated if-fuel :rod consolidation is carried out at the facility because batch identity is lost when the

fuel rods are removed, from the assemblies and consolidated.

4.1 RECORDS:AND REPORTS.

When used in conjunction with reports' from the reactor facility, the AFR

facility records and reports can enable the inspector to establish continuity

of knowledge for each spent fuel assembly or canister of fuel. rods from the

fuel fabrication facility, through the reactor, to the storage facility. The

operating records plan an especially important role in a dry storage facility

because of the heavy reliance on C/S measures for verifying flow and inventory. Investigating anomalies resulting from CCTV failures or seal breakage

may be possible only by examining the operator records and comparing them to

the facility accounting records and the reports sent to the IAEA. Also,

receipts or shipments of spent fuel, loading of fuel into canisters, and

loading canisters into storage casks when an inspector is not present may have

to be indirectly verified by comparing containment and surveillance records

with operator records. Indirect measures are not entirely satisfactory, but

events are likely to occur that provide no other alternative.

4.2 VERIFYING.FLOW AND INVENTORY

The massive building structures and the high radiation levels associated

with the spent fuel restricts the movement of the spent fuel to well-defined

paths inside the facility. This offers possible approaches to help verify flow

and inventory by the use of tamper-resisting seals and video surveillance

equipment.

The entire facility is encompassed by one material balance area (MBA)

having the following key measurement points (KMP) for flow verification:

KMP-1 - Spent fuel receipts

KMP-2 - Placement of fuel in canisters

KMP-3 - Placement of canistered fuel in storage casks

KMP-4 - Shipments of canistered fuel away from the facility.

11
Spent fuel inventory at the facility is located primarily in the storage

casks in the storage area, but there will be additional inventory in the lag

storage area in the fuel handling cell and in unopened receipts in transport

casks in the rece~iving and inspection room. The KMPs for inventory verification are the storage area, KMP-A, the lag storage, KMP-B, and the fuel handling

cell, KMP-C.

The material flow is summarized in Table 4. The rates are based on the

assumption that this facility will receive fuel for 20 years followed by

5 years of shipments. The ratio of PWR to BWR fuel in tons is three to one.

About 5% of the fuel is received already in canisters. The canisters filled

at the facility contain one PWR or BWR assembly, but consolidated fuel rod

canisters may hold rods from up to three PWR or seven BWR assemblies.

The strategic points including the KMPs and locations of IAEA C/S devices

are depicted in Figure 3. Surveillance cameras are located in the transport

cask handling and decontamination room, the storage cask handling and decontamination room, and the cold maintenance area. Because of the possible'

removal of spent fuel via the hot decontamination and maintenance cell, cameras

in the cold maintenance area view the exit from that area. The other two

cameras view the cask movement pathways. In this way, all access to areas in

which unshielded spent fuel can be handled are under surveillance. An inspector is presumed to be present when the fuel is brought into the facility and

when the canisters are filled and placed into storage casks.

TABLE 4. Material Flow at the Baseline Facility

Measu rement

Point Annual Material Flow

KMP-1 30 receipts per year of 3 TN-12 rail

transport casks per shipment

including:

-660 bare intact PWR

assembl ies - 35 canistered PWR assemblies

or equivalent number of rods

-925 bare intact BWR assemblies - 46 canistered BWR assemblies or

equivalent number of rods

KMP-2 660 PWR assemblies

925 BWR assemblies

KMP-3 1666 canisters

50 storage casks

KMP-4 120 shipments of 3 TN-12 rail casks per

shipment (6670 canisters)

12
1. IAEA Seal Location

2. IAEA Surveillance Camera Location

3. KMP-1

4. KMVP-2

5. KMVP-3

6. KM IP-A

7. KM P-B

8. KMVP-C

2-. 1' .- " (T (Thý)

Lag

Storage

Area

crr'0

Transport Cask Transport

Fuel Receiving & Handling & Cask Storage Cask Storage Cask Storage Area

Inspection Room Decon Room Unloading Loading Cell Handling and

Cell Decon Room

FIGURE 3. Safeguards Strategic Points at the Baseline AFR Dry Storage Facility
If an inspector is not present during these operations, the C/S system

must be relied on for verification. No satisfactory methods are available for

verifying receipts after the spent fuel has been canistered and placed in storage casks. This stresses the importance of defining the procedure to hold the

spent fuel in lag storage until the inspector is present.

14
5.0 BASELINE SAFEGUARDS SYSTEM EVALUATION

.In assessing the effectiveness of safeguards at an AFR dry storage facil-ý

ity, consideration needs to be given to the capability of the safeguards

approach to achieve the technical objective with respect to the detection of

diversion of a goal quantity, the timeliness, and the probabilities of detection and false alarms. In assessing the efficiency, the demand for IAEA

resources, including inspection manpower both in the field and at headquarters

as well as the cost of IAEA equipment, should be considered. In addition to

effectiveness and efficiency, which have an impact primarily on the IAEA, the

-degree of intrusiveness on the facility operator must be considered.. Effectiveness, efficiency, and intrusiveness are the principal criteria that can be

used for judging the efficacy of a proposed design feature for enhancing the

implementation of IAEA safeguards.

Other factors for consideration are the possible effects on inspectors'

health and safety; e.g., the expected radiation dose that may be accumulated in

performing an inspection activity, or the physical hazards encountered in verifying seals on top of the concrete storage casks in the open fields of the

storage area.

5.1 CONSTRAINTS

Two quantitative constraints placed on the inspection effort are the

availability of IAEA resources and the limitations specified in the relevant

safeguards agreement. 1he basis for the latter is found in INFCIRC/153 (Corrected), Paragraph 80b. It is doubtful that AFR spent fuel storage facilities

were considered when the criteria for maximum routine inspection effort (MRIE)

were developed, and it is not clear from INFCIRC/153 (Corrected) which criteria

apply. It can be argued that the strategic importance of the spent fuel stored

at the AFR is not different from that at reactor facilities, in which case MRIE

could be based on Paragraph 80a. On the other hand, the large quantities of

plutonium emplaced in the facility each year and the large inventory after a

few years of operation, coupled with the long storage period during which verification by direct measurements cannot be made, may place a different perspective on the importance of safeguarding spent fuel. One-sixth of a man-year

inspection effort as allowed under Paragraph 80a may not be sufficient for

the baseline facility; therefo~re, it may be more reasonable to base MRIE on

Paragraph 80b.

The routine inspection effort at the baseline facility was estimated using

the material flows from Table 4 and the inspector manpower assumptions in the

following paragraphs. For the baseline case, inspectors are present during the

unloading of receipts and during the loading of the storage casks. The estimate is for the period when the facility is receiving fuel, i.e., the first

20 years of operation. No shipments from the baseline facility take place

during this time.

15
5.2 INSPECTION MANPOWER FOR FLOW VERIFICATION

The following inspections are provided as needed:

* Up to 12 hours to observe the unloading of each receipt into the fuel

handling cell.

* 4 hours to observe the loading of the spent fuel canisters into each

storage cask and to apply an IAEA seal to the outer cover of the cask

after it is welded in place.

* The following activities are performed at three-month intervals:

- 8 hours to review video surveillance records - 8 hours to examine facility accounting records - 4 hours to verify seals on storage casks placed in the storage

area during the three month period - 2 hours to count all the storage casks in the storage area - 4 hours to count and identify by serial number the assemblies and

canisters of spent fuel in the fuel handling cell lag storage.

5.3 INSPECTION MANPOWER FOR PHYSICAL INVENIORY VERIFICATION

The following inspections are performed once a year:

* 26 hours to perform activities in item 3 in the previous paragraph

* 150 hours *to inspect the seals on the storage casks in the storage

area on a statistical sampling basis. (The man-hours required for

this activity depend on how long the facility has been in operation;

e.g., after 1 year there are 50 casks and after 20 years there are

1000 casks. We assume an average of'500 casks in the inventory and a

manpower requirement of one-third hour per cask for verifying and

replacing the seal, using a probability of detection of 90% for

detecting 1 seal violation.)

5.4 TOTAL INSPECTION EFFORT

The total inspection time, based on the foregoing assumed activities,

equals 100 inspection days per year. This value includes only routine inspection activities and does not provide time for anomaly resolution that could be

needed if C/S failures occur or when accounting mistakes are detected. It is

reasonable to expect that an additional 10 to 15 inspection man-days may be

needed for anomaly resolution, and for the baseline case, 15 additional days

are included.

16
6.0 DESIGN FEATURES FOR AFR STORAGE _FACILITIES

Specific design features that may facilitate implementing IAEA safeguards

at an AFR spent fuel dry storage facility are identified and discussed below.'

In general, each design feature addresses one or more of the following IAEA

safeguards target areas:

" Reduce effort needed to verify spent fuel receipts, spent fuel

loading into canisters,*and physical inventory in the storage area.

" Enhance capabilities to verify spent fuel receipts, spent fu.eil

transfers in the fuel handling cell, and physical inventories.

The design features considered in this study were analyzed on the basis of

their impact on effectiveness and efficiency of IAEA safeguards implementation.

The effect that each design feature can have on the degree of intrusiveness of

IAEA safeguards on the facility operator was also assessed.

In assessing the value of a design feature, we also considered the practicality of implementation. This deals with the state of development of the

particular safeguards technology that is intended to be used. It is pointless,

for example, to propose adapting the facility design to permit the use of a

safeguards method that is unlikely to be developed in the time frame the

facility is to be operated.

6.1 SPECIFIC DESIGN FEATURES

Quarantine Area for Receipts. This design feature addresses the capability to verify receipts. In the absence of an IAEA inspector during unloading of the receipts into the fuel handling cell, it is not possible to confirm

that the spent fuel was actual'ly unloaded from the tr 'ansport cask. By holding

the receipts in a dedicated area that allows the inspector to count and identify the fuel assemblies or canistered spent fuel, verifi~cation 'can be per-.

formed any time after the cask is unloaded and before the spent fuel receipts

undergo further processing. The reduction in inspection effort th'at can be

realized depends on the capacity of the quarantine area. A quarantine area

with a capacity of 15 metric-tons, i.e., one shipment, will. s'av e 30 mandays/yr. The design permits the quarantine area to be sealed off from the

-operating areas of the fuel handling cell. This is a moderate increase in the

intrusiveness on the facility operator. No undeveloped safeguards technology

is involved.

CCTV Surveillance of Fuel Loading Process. Designing the fuel loading

process including equipment and equipment location so the fuel rod consolidation step and the loading of fuel assemblies into canisters can be viewed and

recorded by CCTV enables verification of material flows without direct visual

observation by the IAEA inspector. This feature can detect undeclared fuel

transfers or attempts to substitute dummy assemblies or canisters. The effect

17
on inspection manpower is the additional time required to service the CCTV systems and to review the recordings. This is estimated to add about one inspection man-day per inspection or 4 inspection man-days/yr. The development of

the CCTV systems is well advanced, but remotizing and radiation hardening (.or,

development of shielding designs) for the in-cell units are-yet to be done.

The level of intrusiveness for this design feature will be less than in the

baseline case.

Capability to Photograph Weld Beads on Canister Covers. Weld beads have

unique appearances that are suitable for identifyiTn~g fillTed canisters to verify

the inventory of filled canisters awaiting loading in the storage casks or to

resolve anomalies. It addresses the problem of substituting dummy canisters

with falsified serial numbers. Implementing the inspec 'tion activities, associated with this design feature will increase the demand on inspector man-hours

by several inspector man-days/yr. No additional IAEA equipment is involved.

The technique for photographing weld beads is well developed, but doing it

remotely in a hot cell will require additional engineering development.. The

level of intrusiveness will be significantly greater than the baseline case

because of the need to move canisters to a viewing location.

Weighi ng Station for Filled Canisters. This design feature provides an

attribu-tes test for verifying the inventory of canisters in the fuel handling

cell. This test can be done on a statistical sample of the items in the inventory' during the annual physical inventory verification inspection. The,

increase in inspection manpower depends on the size of the inventory, but it

is reasonable to expect it to increase by about two man-days/yr. The technology for remote in-cell weighing using load cells is well developed. The

load cell can be attached to an in-cell crane, thus allowing the canisters to'

be weighed without moving them from the storage location.

Quantitative Measurement of Receipts. This design feature would enable

verification of receipts ar 'riving at the facility, in transport casks that were

not sealed at the reactor facility. It provides protecti~on against substitution with dummy fuel'assemblies or unirradiated fuel assemblies having falsified serial numbers. The impact on inspection manpower would depend on the

number of-transport casks that arrive without IAEA seals. Two possible systems

are envisioned for the NDA measurement: a portable system simi-lar to the one

being developed by Los Alamos National Laboratory that measures the gross gamma

and neutron emisvions,9 or the gamma energy analysis system described by Nilsson, et al., with a collimator built into a shielding wall of the hot cell.

and a mechanism for scanning the fuel assemblies one at a time. The portable

system was designed for use in water storage pools at reactors and has been

field tested. Additional development is required before the system can be used

in a hot-cell. The gamma energy analysis system using a built-in collimator

has been designed for use in the CLAB facility in Sweden.

Qualitative Measurement of Spent Fuel in the Fuel Handling Cell. This.

design7feature permits a gross gamma emission measurement of the spent fuel

assemblies and canistered fuel. It permits attributes testing for inventory

verification. Because gamma ray emissi~on is more difficult to falsely simulate

than mass, it would provide greater assurance than a weighing system. The

18
measurements are made with high-level radiation detectors such as'ion chambers.

The test would be performed on a statistical sample of the assemblies and canisters in the fuel handling cell during the annual physical inventory verification. The additional inspection manpower over the baseline case depends on

the size of the inventory, but it is reasonable to expect it will add about

10 man-days/yr of inspection effort. The measurement technology needed for

this design feature is developed, but engineering development is needed for

each specific application. The level of intrusiveness will increase over the

baseline case.

Remote Operation and Monitoring of the Cask and Fuel Handling-Operations.

Operating the process from a central control room using CCTV to 'view the spent

fuel movements permits the IAEA inspector to observe the movements as they

occur. Video recording of the process during the time the inspector is absent,

i.e., during the second or third shift, will allow detection of undeclared

movements around the clock. The reduction in inspection man-hours associated

with this design feature is estimated to be 4 inspection man-days/yr. CCTV

systems for viewing and recording operations in hot cells are widely used in

the nuclear industry today. However, adaptation of designs to give the

tamper-resistance needed for independent verification will be required. This

design feature will reduce intrusiveness.

Compartmentalization of the Storage Area. This design feature requires

the storage area to be subdivided into sub-areas that can be sealed off after

all the space for storage casks has been~filled. In this way, the testing of a

single seal on the access gate to the sub-area can supplant the need for testing the seals on every cask in the sub-area. This will reduce the effort for

the annual physical inventory inspection. The baseline facility-has space for

1000 storage casks, of which 500 will be filled at the end of 10 years of

operations. Assuming each sub-area contains 100 storage casks, the physical

inventory verification of the storage area will require the testing of 5 seals.

Under the baseline case, 450 seals on the storage cask outer covers are

inspected annually to achieve 90% probability of detection. The net reduction

in inspection manpower from this design feature is about 18 inspection mandays. The only IAEA equipment needed are the seals and possibly surveillance

cameras, both of which are widely used by the IAEA inspectorate. The level of

intrusiveness is reduced by this design feature.

Limit Access to the Storage Area by the Storage Cask Transporter to a

Single Pathway. This design feature requires the perimeter of the storage area

to be surrounded by a barrier that is impenetrable by the storage cask transporter except through a single access point connected to the fuel receiving and

handling building. By viewing the access point with a surveillance camera, the

cask movements can be monitored, permitting the detection of any undeclared

storage cask movements. This provides a possibility for resolving anomalies

caused by inadvertent loss of containment of a storage cask, e.g., damaged

seals or indications of cask movement. The main objective is to improve safeguards effectiveness, and this design feature has little impact on safeguards

efficiency. *The implementation of this design feature relies only on surveillance cameras that are currently~avail~able to the IAEA. The level of intrusiveness is reduced by this design feature.

19
Storage Cask -Designs,-That AllIow the Use of Fi bre Opti c Seal s wi th Remote

Readout of the'Status of the Seals. This design feature requires cask designs

that permit attachment of elTectr-onic fibre optic seals with provisions for protection of the seals from the weather, and allow interrogation of the status of

the seals'by the IAEA from a remote location, i.e., outside thel~tyjage area

and possibly from IAEA headquarters, as proposed by the RECOVER system.

Each year during the first 20 years of operation of the baseline facility, 50

storage casks are added to the storage area, requiring 50 fibre optic seals to

be attached and included i.n the electronic readout network. A totally reliable

system could reduce the inspection effort for physical inventory verification

of the storage area by the time required to verify-the conventional seals used

in the baseline safeguards approach, less the time required to affix the

50 fibre optic seals added each year,, plus the time for maintaining previously

deployed seals and the readout system. There is a potential for reducing the

inspection effort by 10 to 15 man-days/yr. The technology for fibre optic

seals is under active development. The level of intrusiveness is reduced by

this design feature.

Designs to Allow Sealing of Storage Casks to the Concrete Pads. This

design feature also addresses the problem of verifying the inventory in the

storage area. Considerable inspection time can be saved by eliminating the

need to climb to the top of the storage casks to test and replace seals. If

this design feature is implemented, the IAEA inspector will affix seals on all

casks that have been emplaced in the storage area since-the last annual or

interim inspection. He first confirms the cask identity and verifies the

containment by checking the.IAEA seal placed on the outer cover of the cask

while in the receiving and handling building. Reduction in inspection effort

is estimated to be 4 man-days/yr. This value was reached by estimating a time

for testing a seal at ground level of one-sixth hour versus one-third hour for

one on top of a cask, and using 500 casks as the average number in the'storage

area. To this was added hours for affixing seals on the 50 casks emplaced in

the storage area each year. This design feature uses no IAEA equipment other

than conventional seals. The level of intrusiveness is reduced by this design

feature.

Provide Temperature Sensors for the Storage Casks. This design feature

involves inclusion of one or more thermocouple wells leading to the cavity of

the storage casks. This will permit inspectors to observe the temperature

differential between the cask interior and exterior, providing a measure of the

heat of decay associated with the spent fuel. This is a straightforward test

to verify the contents on an attributes basis. Failure of seals or other indications of loss of containment of the storage casks creates a difficult verification problem. It will be costly to move a storage cask back to the fuel

handling and receiving building to permit direct verification of a cask's

contents, but a temperature measurement is a satisfactory alternative. It is

reasonable to believe that 3 to 5 casks/hr can be tested in this manner. No

estimate of the impact on inspection effort can be made in the absence of data

on the frequency of sea~l failures over long periods. Use of the test requires

a calibrated thermocouple and a millivolt meter. The level of intrusiveness is

not affected by this design feature.

20
.Provisions for Gamma, Energy Analysis of Storage Cask Contents. A gamma

energy spectrum obtained at the time the storage cask is place'd-in the storage

area will provide an initial measure of the gamma emitting fission products in

the cask. Subsequent measurements can-*be made to verify the contents using the

characteristic gamma spectra of long-lived fission prod 'uct isotopes 'such-as

Cs-137 or Eu-152. One.,approach is to measure the ratio of counting rates of

two different isotopes. After performing the measurement once, a subsequent

measurement made at a known time in the future can be used to verify the cask's

contents by comparing the two results, after correcting for radioactive . P

decay. The main effect of this design feature on inspection effort is time. -

required to perform the initial measurement. A single measurement can be made

in about one-half hour, and 50 measurements per year are required. The

increased inspection effort is about 3 man-days/yr. The technology for this

design feature is developed, but specific designs for the collimator in the

shielding wall of the storage casks are needed. Detectors and multichannel

analyzers are available. This design feature will increase the-level of

intrusiveness.

6.2 GENERAL FACILITY CHARACTERISTICS

General characteristics of an AFR facility that would permit adoption of

one or 'more design features include adequate lighting in the operating areas

for the IAEA surveillance cameras, structural designs and plant layout to mini-*

mize the entry and exit portals that accommodate spent fuel, plant designs that

are compatible with the placement, operation, and servicing of surveillance

cameras, and storage cask design that utilizes placement of external sealing

techni ques.

21

REFERENCES

1. A. Nilsson, B. Gustafsson, "Safeguards at an Away-From-Reactor Storage

Facility," In Proceedings of Nuclear Safeguards Technology 1982,

International Aomic Energy Agency, Vienna, Austria (1982).

2. Nuclear News, "Construction Start on German AFR," p. 67 (September 1984).

3. G. Borgonovi, et al., Design Features to Facilitate IAEA Safeguards at

Light Water Reactors, NUREG/CR-2265, prepared for the Nuclear Regulatory

Commission by Science Applications, Inc., La Jolla, California (1981).

4. M. F. Mullen, et al., Design Features to Facilitate IAEA Safeguards at

Reprocessing Plants, NUREG/CR-2312, prepared for the Nuclear Regulatory

Commission by the Pacific Northwest Laboratory, Richland, Washington

5. International Atomic Energy Agency, IAEA Safeguards Glossary. IAEA/SG/

INF/1, International Atomic Energy Agency, Vienna, Austria- (980).

6. E. Tkharev, Safeguarding of Long-Term Spent Fuel Storage, STR-1499

International Atomic Energy Agency, Vienna, Austria (1984).

7. S. Ermakov, Safeguards Approach for Away from Reactor (AFR) Spent-Fuel

Storage, Part 1-Wet Type of Storage, STR-119, International Atomic Energy

Agency,- Vienna, Austria (1982).

8. International Atomic Energy Agency, The Structure and Content of Agreements Between the Agency and States Required in Connection with the Treaty

on the Non-Pr-oliferation of Nuclear Weapons, INFCIRC/153 (Corrected),

International Atomic Energy Agency, Vienna, Austria (1972).

9. J. H. Phillips, et al., "Experience Using a Spent-Fuel Measurement

System," J. Nuclear Mater. Manag. Proceedings Issue, Vol. XII,

pp. 175-181, Instit-u-te of Nuclear Materials management, Chicago,

Illinois (1983).

10. E. V. Weinstock and J. B. Sanborn, An Evaluation of a Remote Continual

Verification System, RECOVER, for International Safeguards, ISPO-198,

BrookWhaven National Laboratory, Upton, New York (1983).

11. 5. J. Crutzen, et al., "Remote Controlled and Long Distance Unique Identification of Reactor Fuel Elements or Assemblies," in Poednsof Nuclear Safeguards Technology, pp. 561-571, International Atomicf&Energy

Agency, Vienna (1979).

23

NUREG/CR -4517

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22 Pacific Northwest Laboratory

N. L. Ha rms(1:

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F. P. Roberts

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Publishing Coordination (2)

Technical Information (5)

Distr-1

NFIC FORM 335 U.S. NUCLEAR REGULATORY COMMAISSIONd I REPORT NUMBER lA,,,gneadbv TIDC, add V&i No, il 0v0

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NF1CM 1102.

3201.3202 BIBLIOGRAPHIC DATA SHEET NUREG/CR-4517

SEE INSTRUCTIONS ON THE REVERSE. . PNL-5321

2. TITLE AND SUBTITLE 3. LEAVE BLANK

Design Features for Enhancing.International

Safeguards of Away-From-Reactor Dry Storage 4. DATE REPORTCOPED for Spent LWR Fuel MNHY EAR

5. AUTHOR(S) February 1986

F. P. Roberts MNH6. DATE REPORT ISSUED EA

N L.HrsOctober 1986 7. PERFORMING ORGANIZATION NAME AND MAI LING ADDRESS (IneI~de Zlo Code) 8. PROJECT/TASKIWORK UNIT NUMBER

Pacific Northwest Laboratory 9. FIN OR GRANT NUMBER

P.O. Box 999

Richland, WA 99352 FIN B2475

10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS fIncIudeZiP Code) Ila~. TYPE OP REPORT

Division of Safeguards Technical

Office of Nuclear Material Safety and Safeguards b. PERIOD COVERED (nclusivedates)

U.S. Nuclear Regulatory Commission

Washing ton, DC 20555

12. SUPPLEMENTARY NOTES

13. ABSTRACT (200 words o' (ass)

The Pacific Northwest Laboratory has performed a study for the Nuclear Regulatory

Commission to identify and analyze design features that can facilitate the implementation of IAEA safeguards at away-from-reactor facilities for dry storage of light

water reactor spent fuels. Thirteen specific design features are identified that can

enhance verification of nuclear material flow and. inventory. These are assessed from

the viewpoint of safeguards effectiveness and possible impacts on the IAEA and the

operator of the AFR facility.

14. DOCUMENT ANALYSIS -. KEYWORDS/DESCRIPTORS 15. AVAILABILITY

STATEMENT

Design Features, Dry Storage, Away from Reactor Dry Storage, Sp~ent Fuel, Uni te

Spent Fuel Storage, AFR Dry StorageUni te

16. SECURITY CLASSIFICATION

Is, (DENT IF IE RS/OPEN.ENDEO TERMS Unclassified

/Ths,s reo~'d

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