NDA techniques for spent fuel verification and radiation monitoring · 2004. 8. 3. · 6 CONCLUSIONS OF EXISTING NDA TECHNIQUES 17 6.1 Gross defect verification of assemblies 17 6.2

STUK-YTO-TR 133 FI9700145AUGUST 1997

NDA techniques for spent fuelverification and radiationmonitoringReport on Activities 6a and 6b of Task JNTC799 (SAGOR)Finnish Support Programme to the IAEASafeguards

Matti TarvainenRadiation and Nuclear Safety Authority, Helsinki, FinlandFerenc LevaiTechnical University, Budapest, HungaryTimothy £. ValentineOak Ridge National Laboratory, Oak Ridge, TN, USAMark AbholdLos Alamos National Laboratory, Los Alamos, NM, USABruce MoranUSNRC, Washington D.C., USA

The conclusions presented in the report are those of the authorsand do not represent the official position of the Radiation andNuclear Safety Authority.

RADIATION AND NUCLEAR SAFETY AUTHORITY (STUK)P.O.BOX 14, FIN-00881 HELSINKI, FINLANDTel. +358-9-759881Fax +358-9-75988382

V9 - 0 1

ISBN 951-712-222-5ISSN 0785-9325

Oy Edita AbHelsinki 1997

STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

TARVAINEN, Matti (STUK), LEVAI, Ferenc (Budapest Technical University), VALENTINE,Timothy E (Oak Ridge National Laboratory), ABHOLD, Mark (Los Alamos National Laboratory),MORAN, Bruce (USNRC). NDA techniques for spent fuel verification and radiation monitoring.Report on Activities 6a and 6b of Task JNT C799 (SAGOR). Finnish Support Programme to theIAEA Safeguards. STUK-YTO-TR 133. Helsinki 1997. 23 pp. + Annexes 24 pp.

ISBN 951-712-222-5ISSN 0785-9325

Keywords: safeguards, spent fuel, NDA

ABSTRACT

A variety of NDA methods exist for measurement of spent fuel at various stages of the dispositionprocess. Each of the methods has weaknesses and strengths that make them applicable to one ormore stages in disposition. Both passive and active methods are, under favorable conditions, ca-pable of providing either a mapping of an assembly to identify missing fuel pins or a measurementof the fissile content and some are capable of providing a mapping of a canister to identify missingassemblies or a measurement of the fissile content. However, a spent fuel measurement system ca-pable of making routine partial defect tests of spent fuel assemblies is missing.

The active NDA methods, in particular, the active neutron methods, hold the most promise for pro-viding quantitative measurements on fuel assemblies and canisters. Application of NDA methodsto shielded casks may not be practical or even possible due to the extent of radiation attenuation bythe shielding materials, and none of these methods are considered to have potential for quantitativemeasurements once the spent fuel cask has been placed in a repository.

The most practical approach to spent fuel verification is to confirm the characteristics of the spentfuel prior to loading in a canister or cask at the conditioning facility. Fissile material tracking sys-tems in addition to containment and surveillance methods have the capability to assure continuityof the verified knowledge of the sample from loading of the canisters to final disposal and closingof the repository.

RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

CONTENTS

ABSTRACT Page

1 INTRODUCTION 5

2 ROLE OF VERIFICATION MEASUREMENTS 6

3 POTENTIAL SPENT FUEL MEASUREMENT POINTS 8

3.1 IAEA spent fuel verification criteria 83.2 Spent fuel flow chart 83.3 Measurement point categorization 103.4 Conditioning of spent fuel 10

4 PROPERTIES OF SPENT FUEL AND CONTAINERS 114.1 Spent fuel parameters 114.2 Disposal containers 124.3 Mixed emplacement 12

5 APPLICABILITY OF NDA-METHODS TO SPENT FUEL VERIFICATION 145.1 Measurement principles 145.2 Instruments for spent fuel verification 155.3 Fissile material tracking 16

6 CONCLUSIONS OF EXISTING NDA TECHNIQUES 176.1 Gross defect verification of assemblies 176.2 Partial defect verification of assemblies 176.3 Radiation monitoring of casks 186.4 Unsolved NDA problems 19

REFERENCES 20

ANNEX 1 Description of NDA Methods 24A1.1 Cerenkov 24A1.2 Passive gamma-ray techniques 24A1.3 Active gamma-ray techniques 25A1.4 Passi ve neutron techniques 26Al .5 Active neutron techniques 27A1.6 Other techniques 29A1.7 Table A1.1, NDA methods for spent HEU fuel 30A1.8 Table A1 .II, NDA methods for spent LEU fuel 31

ANNEX 2 List of NDA instruments 32

ANNEX 3 Relevant R&D projects 41A3.1 Fissile material measurement through the cask wall 41A3.2 Pu measurement of spent fuel 42A3.3 Partial defect testing of multi-assembly configurations 44A3.4 Verification of casks by passive neutron detection 46

ANNEX 4 Summary of work plan for SAGOR (Task C 799) 47


1 INTRODUCTION

A joint task of the Member State SupportProgrammes called "Programme for Develop-ment of Safeguards for Final Disposal of SpentFuel in Geological Repositories (SAGOR)" hasbeen carried out under Task JNT C799. Forpractical reasons, SAGOR has been dividedinto 10 separate Activities (see Annex 4), thelast being compiling of an integrated finalreport Activity 6 "NDA techniques for spentfuel verification and radiation monitoring" hasbeen carried out in co-operation with theSupport Programmes of Finland, Hungary andUSA.

Developing safeguards for final disposal issuesa challenge to the safeguards community andsafeguards experts. Conventional safeguards isdirected to nuclear material that can be handled,measured and monitored. All of the safeguardsmeasures developed, so far, are based on thisassumption. Final disposal of spent fuel ingeological repositories is a challenge becausenone of the conventional features are valid oncethe container has been placed into the repositoryand the access backfilled. Nuclear material canno more be handled, measured or even moni-tored. The safeguards measures should focus onindirect objects like surface of the closedrepository [1].

Final disposal in geologic repositories is meantto be an irreversible process. In this light,verification measurements of the declared para-meters of the nuclear material have to beperformed in a way to give all answers to thequestions made now and also in the future. The

main concerns include the amount' of nuclearmaterial deposited and its isotopic composition.The ultimate objectives of safeguards, i.e. "thetimely detection of diversion of significantquantities of nuclear material from peacefulnuclear activities to the manufacture of nuclearweapons or of other nuclear explosive devicesor for purposes unknown, and deterrence ofsuch diversion by the risk of early detection",can be achieved using conventional safeguardsmeasures while the material is in the con-ditioning facility or even in an open repository.How this goal can be reached in conditioningfacilities and open repositories using non-destructive assay (NDA) is the main topic ofthis report. In the case of a closed repository,the NDA techniques measuring radioactivity ofspent fuel are no more applicable and this phaseof the final deposition is not handled in thisreport.

The report covers a wide range of measurementactivities. Section 5 deals with the applicabilityof different measurement principles for NDA ofspent fuel. Conclusions of the present NDAtechniques are drawn in section 6. Annex 1includes a summary of the principles andmethods and Annex 2 lists existing NDAinstruments that are available or under devel-opment. These methods can be used in planningthe safeguards verification activities of the finaldeposition. Annex 3 lists R&D projects whichare considered to have the highest relevance forthe future final deposition safeguards measure-ments.


2 ROLE OF VERIFICATIONMEASUREMENTS

Independent verification of spent nuclear mate-rial by the IAEA makes use of NDA measure-ments. The material to be deposited has beenunder IAEA safeguards during the front-end ofthe fuel cycle prior to entering the conditioningfacility and the repository. Also verificationmeasurements have been performed as a part ofthe routine safeguards activities. Prior to load-ing into the storage containers to be deposited,spent fuel assemblies may loose their integrityunder consolidation and conditioning. Verifi-cation measurements need to be optimized togive full and continuing assurance that theoperator declared data are complete and correct.The logical order of safeguards measures isschematically shown in Figure 1.

The triangle of Figure 1 applies also to thesafeguards of the conditioning facility. In de-signing the safeguards measures, verification ofdeclared data is logically followed by con-tainment and surveillance (C/S) to maintaincontinuity of the verified knowledge. In an openrepository, use of C/S methods for monitoringof cask movements is foreseen.

Spent fuel arriving at the conditioning facilityhas operator declared mass, isotopic com-position, burnup and cooling time data Theparameters that can be directly verified by NDAmethods include mass, isotopic compositionand cooling time. Cooling time verificationmeasurements are based on the isotopic corre-

lation techniques of the fission product gamma-rays. Burnup (exposure), i.e. energy releasedper unit mass, can be directly verified onlyduring power production. Fission productgamma-rays can, however, be used for indirectverification of the burnup and the contents oftransuranic isotopes. In addition to the gamma-ray methods, passive and active neutron assaycan be used for verification of fissile content ofspent fuel.

One of the main parameters to be verified priorto final deposition is the completeness of thedeclared data. The optimum amount of infor-mation a verification measurement can producegives assurance that the material declarationrelates to the material measured, that theisotopic composition and mass data are correctand that no material is missing or has beenreplaced by dummies. Because spent fuelverified at the conditioning facility has beencooled and stored for decades, no details ofearlier verification measurements may be avail-able anymore or can be verified by re-measure-ments. In this light, one of the main goals of theverification measurements at the conditioningfacility should be to give assurance that spentfuel assemblies and rods are really irradiatednuclear material. If absence of diversion can beconfirmed, details of the composition of thematerial to be deposited is of minor interestfrom the proliferation point of view.


C/S(continuity

-of-knowledge)

Verification- correctness (isotopic, mass)- completeness (no diversion)

Accounting and reporting

Initial inventory/inventory changes

Figure 1. Logical order of safeguards measures applied during the fuel cycle of nuclear material.


3 POTENTIAL SPENT FUELMEASUREMENT POINTS

3.1 IAEA spent fuel verificationcriteria

IAEA has incorporated in the Safeguards Cri-teria the activities that are necessary to carryout the IAEA safeguards. In addition to regu-lations concerning inspection frequencies atdifferent types of facilities for verification ofdifferent types, forms and amounts of material,the Criteria define verification activities to beperformed. Different levels of verification arealso specified. The Criteria have been updatedannually during 1992-1996 [2]. Each of the tenfacility types sections has also verificationmeasurement tables. These tables list the re-commended instruments to be used to performthe required verification measurements. TheSafeguards Criteria obviously do not includecriteria for final deposition or conditioningfacilities. From the point of view of safeguardsverification measurements of spent fuel, e.g.those parts of sections 1 "LWRs" and 9"Storage Facilities" related to irradiated direct-use material are, however, relevant.

In conditioning facilities, spent fuel NDAverification measurements can be performed onthe gross defect level or on the partial defectlevel. Gross defect means that the wholeassembly (all nuclear material) is replaced witha dummy or the material is missing. Partialdefect means that one half or more of theirradiated rods of an assembly are replaced ormissing. The two categories include all methodsavailable.

The target of diversion is at least 1 SQ ofnuclear material. Within the reference reposi-tory a typical container will have about 14 kg ofPu plus about 23 kg of 235U. Thus, in order todivert at least 1 SQ, only one container of fuelwould have to be diverted and opened.

3.2 Spent fuel flow chart

According to the IAEA glossary [3], the keymeasurement point (KMP) is defined as "alocation where nuclear material appears in sucha form that it may be measured to determinematerial flow or inventory". The KMP conceptapplies to the final deposition as long asmaterial is available for measurements, i.e. atthe conditioning facility. Figure 2 shows sche-matically the flow of spent fuel assembliesfrom the point of view of verification measure-ments using NDA techniques.

The flow chart begins from a reactor pool (topleft) which receives spent fuel from the reactor.After several loading, transport and reloadingoperations as well as possible reconstruction,consolidation and conditioning operations theflow ends in a closed repository (bottom right).

After possible reconstruction operations (topmiddle), spent fuel assemblies are moved fromthe reactor pool to an interim storage. Whenseparate assemblies are stored in a storage rack,NDA verification measurements can be per-formed rather easily. This is indicated in Figure2 by placing the squares in the left-handcolumn. If assemblies or cans of rods are storedin closed containers, verification measurementsare more complicated, if possible at all. This isindicated by placing the squares in the right-hand column, respectively. The column in themiddle includes those parts of the spent fuelflow where fuel is moved or handled. For NDA,this may offer a potential measurement possi-bility because the fuel is moved, anyway.

The upper half of Figure 2 shows, for complete-ness, the storage history of spent fuel before itis transported into the conditioning facility. The

8


need for NDA verification depends on theinformation available from the fuel and thecontinuity-of-knowledge. The lower part under

the SAGOR limit in Figure 2 includes that partof the fuel cycle which is covered by theSAGOR task.

Access to separateassemblies

Cask handling orfuel operation

Cans / assembliesin closed containers

reactor pool > reconstruction2b

reactor pool

interim storage pool k-

») loading (<--

T 22 I transport j —

Tdry storage

vault (type 1)

unloading

loading

dry storage(type 2)

SAGOR LIMIT

conditioning facilityreception pool

conditioning facilityreception vault

transport

\ unloadingconditioning facility

buffer storage

consolidation

conditioning

conditioning facility

4,

ground surface

4, |oa6b

ding

' |transport /

transfer

4bJ repository> receipt storage

• >j open repository | S

j closed repository

Potential measurement points:1 Separate assemblies stored in storage rack under water, access from above without

fuel movement, from the side, if fuel moved.2 Separate assemblies (or closed cans) moved under water, access also from the side.2b Separate rods moved under water access also from the side.3 Separate assemblies stored in a dry storage vault, access from above in air.4 Separate or restructured/conditioned assemblies moved or handled in air, access

also from the side.4b Separate rods moved in air, access also from the side.5 A shielded cask or unshielded canister stored (including cans or assemblies).6 Rods stored in a closed can.6b Rods in a closed can moved in air.

Figure 2. Potential measurement points of spent fuel during the back-end of the fuel cycle ending in thefinal deposition. Methods under the SAGOR limit are discussed in this report.


3.3 Measurement pointcategorization

Figure 2 includes tags with numbers l-6b.Numbers indicate possible measurement points.Storage measurement points are shown in theleft-hand and in the right-hand column, flowmeasurement points in the mid-column, respec-tively. Another division is made into measure-ment points under water or in the air.

The number of potential measurement methodsapplicable for verification under the SAGORlimit in Figure 2 is large. Similar instrumentsand methods may have facility specific featureswhich still increases the total number of appli-cations. This is why similar measurementconditions at different points of the flow charthave been tagged with the same number. Forexample, the measurement point number " 1 "(Figure 2) is attached to all those places whereseparate spent fuel assemblies are stored inracks under water. In these points measurementscan, in principle, be performed from abovewithout moving the fuel. If the fuel assembly ismoved or lifted e.g. using a fuel handlingmachine or crane, measurements can be per-formed without interference from the sur-rounding assemblies and also from differentvertical positions. Different equipment andmeasurement methods are obviously needed formeasurements from above and from the side.Categorization of the measurement points ismeant to simplify the handling of the veri-fication question and the search for suitablemethods for different conditions.

At different stages of the flow (Figure 2) spentfuel will be transported and stored in closedcontainers. If a closed container is supposed tobe re-opened, verification of separate assem-blies or cans may be possible. Such verificationactivities are related to the conditioning facility.

In the SAGOR task it has been assumed thatwhen entering a geological repository, all spent

fuel will be in closed welded and/or sealeddisposal containers on which adequate veri-fication has been performed at the time ofloading and the continuity-of-knowledge hasbeen maintained since the time of the containerloading. Measurement methods applicable forverification of closed containers are few innumber. However, they are also dealt with inthis report.

3.4 Conditioning of spent fuel

At the conditioning facility spent fuel will betransferred into a hot cell for further processing.The following types of fuel will be received:• intact fuel assemblies• defective fuel assemblies• containers of damaged/defective fuel rods.

In the hot cell the fuel will be processed in oneof the following manners:• Encapsulation of intact fuel assemblies into

the disposal cask. This involves no dis-assembling.

• Encapsulation of disassembled fuel compo-nents into canisters and subsequent en-capsulation of canisters into the disposalcask. Depending on the operational require-ments, fuel and non-fuel bearingcomponents will be separated.

• Disassembled fuel components will bedivided into smaller pieces before placinginto the disposal cask.

• Encapsulation of waste generated by theprocess into the disposal cask together withother material with same specifications.

During conditioning, transition of items fromone type (e.g. fuel assemblies or rods) intoanother type (i.e. final disposal container) willoccur. The contents of the new item may beverified in respect to the nuclear materialcontents.

10


4 PROPERTffiS OF SPENT FUEL ANDCONTAINERS

4.1 Spent fuel parameters

Typical spent fuel parameters for 5 differentfuel types are shown in Table I. The Table iscompiled by D. M. Wuschke for Activity 1 ofthe SAGOR project The Table shows thoseparameters of fuel assemblies that are relevantfor safeguards, i.e. nuclear material contents as

well as physical dimensions of fuel assembliesthat emit and also absorb the radiation used forverification measurements.

Existing safeguards measurement techniqueshave been challenged by recent technologicaldevelopment for better management of thereactor fuel. New fuel types are designed for

Table I. Characteristics of spent fuel intended for direct disposal.

Characteristics

Reactor size (MWe net)

Approximate fuel assemblydimensions (cm)

LengthCross sectionSide (square)Diameter (cylinder or sphere)

Mass per assembly (kg)TotalHeavy metal

Rods per assemblyPower density (MW/m3)Design burnup (GWd/Mg)Total activity (Ci/kg)

after 10 aDecay heat (W/kg)

after 10 aCalculated fuel discharge rate

MgU/GWe.yrFuel enrichment

initial % 235Uafter irradiation % 235U

Plutonium content after irradiationkg/assemblykg/fuel rod

No. of items per SQ of PuFuel assemblyFuel rod

PWR

1000

320-483

19-23

480-840322-548126-331

8526-50

320

2,3

32-38

3,0-4,40,8-1,26

3-50,014

2-3560

BWR I HWR-I CANDU

1000

447

14-15,3

250-307172-194

100-900

49,5

8,1-10,3

16,6-24,713,4-19,8

47-64 ! 19-3750 ! 10

THTR/HTR-500300/500

N/A

6(sphere)

N/A11

_J^A__ j

RBMK

1000/1500

1006

7,9

18513018

32/36 | ?27,5-40 \ 6,5-8,1 | 80-160 | 25

290 | 84|

? ?]

2,2 | 0,22 I ?

38-40 150

2,5-3,5 | 0,710,8-1,0 | 0,205-0,282

1,2-2,00,022

4-7360

?

93/9,180/8,1

0,06-0,08 I N/A0,002-0,004 I N/A

100-1204000

N/AN/A

?

60

2,0-221

??

??

D.M. Wuschke: A reference repository for the development of safeguards for disposal of spent fuel

11


easy disassembling. During rod exchange, fuelassemblies lose their original constituents andidentity. In addition to accountancy, this maycause problems to C/S and verification meas-urements. Advanced fuel designs incorporatingdifferent initial enrichments, axial variations inenrichment, use of MOX and burnable poisonrods introduce additional difficulties for NDAmeasurements. Use of special storage basketsor multielement bottles (MEB) or multipurposecanisters (MPC) to accommodate an increasingnumber of spent fuel assemblies have addedfurther complications also to verification meas-urements.

4.2 Disposal containers

A geological repository for final disposal maybe designed for emplacement of disposal con-tainers within the rooms or in bore holes drilledin the walls or floors of the rooms. Some statesintend to dispose spent fuel, high level waste(HLW) and possibly low and intermediate levelwaste (LILW) in the same repository. Thedisposal containers for the spent fuel and thetwo types of waste would be designed to bedifferent in appearance.

Characteristics of disposal containers for fivecountries are summarized in Table II. The basicunit handled in the repository will be thedisposal container. When not within a shieldedfacility, the disposal container is enclosed in are-usable disposal container cask. Spent fuel isnormally not present in the repository exceptwithin a disposal container, and the repositorydesign and operation are independent of mostspent fuel characteristics. However, the size andmass of the fuel affect the design of thecontainers and container casks. Individual fuelassemblies or rods would be present only if acontainer were damaged or deliberately opened.

4.3 Mixed emplacement

In those countries where both direct disposaland reprocessing of spent fuel is considered,simultaneous disposal of both high level waste(HLW) and spent fuel might occur in the samegeological repository. For safeguards purposesit will be required to distinguish canisterscontaining HLW from canisters containingspent fuel. This will be important for thematerial flow into the repository as well as forthe possible material flow out of the repository.

Table II. Characteristics of currently considered disposal containers.

Type of fuel

Disposal Container• Material

• Height (m)• Diameter (m)• Wall thickness (mm)• Mass with Fuel (Mg)• Self-ShieldedMass of Fuel/Container

Assemblies/Container

(max).Surface Dose Rate

jmSv/h)Pu content (kg)

(approx.)

Canada

HWR

Titanium

2,250,636,352,8no

1,65

72

5x10"

5

Germany

LWR

Steel

5,351,5735064

yes

3,0

8 PWR or24 BWR

<0,2

30

HTRpebbles

Steel

5,501,4535055,0yes

1,7

8400

?

?

Sweden

LWR

Copper/Steel4,500,8100

18,5-22no

1,5

8-12 BWRor 4 PWR

?

14

UnitedStatesLWR

PWR/BWR

Steel

5,31,65/1,6

12050/46

no

8

44 BWR21 PWR120-290/

L_ 80-270100-150/80/144

Finland

BWR

Copper/Steel4,50,8100

14-19no

1,5

9 BWR

100

14

D.M. Wuschke: A reference repository for the development of safeguards for disposal of spentfuel (Report of SAGOR Activity 1B)

12


Table HI. Characteristics of disposal geologies and repository design.

Package Type

Mixed Emplace-ment with HLWCapacity (MgU)Disposal Rate• Mg U/a• Assemblies/d• Containers/d

Belgium

unshielded

unknown

3,500

120

?

Canada

unshielded

no

191,000

4,7001,080

15

Germany

selfshieldedprobably

10,000

200??

Sweden

unshielded

no

8,000

300

<1

Finland

unshielded

no

1,840

140?

<1

UnitedStates

unshielded

yes

87,000

3,000

3-8

D.M. Wuschke: A reference repository for the development of safeguards for disposal of spent fuel

Table III includes selected parameters related to the mixed emplacement of HLW and spent fuelfor the final disposal concept of 6 countries.

13


5 APPLICABILITY OF NDA-METHODSTO SPENT FUEL VERIFICATION

5.1 Measurement principles

This section provides a summary of the prin-ciples of nondestructive assay (NDA) tech-niques that can be considered for use in one ormore phases of spent fuel disposition. A moredetailed description on existing NDA methodsis included in Annex 1.

Passive NDA measurement methods rely on thedetection of radiation emitted from materials tocharacterize the material The measurementsystem may measure the primary radiationemitted directly from the material or secondaryradiation emitted because of interactions ofprimary radiation. Passive NDA methods aregenerally categorized into two areas: gamma-ray and neutron measurements, but calorimetricand Cerenkov measurements can also be con-sidered as passive methods. Passive meas-urements require reference measurements toobtain quantitative information about the spentfuel burnup and a destructive analysis couldthen be performed on the reference assembly toobtain the absolute burnup. Calibration curveswould have to account for differences in theinitial fuel enrichment, irradiation history, mod-erator conditions, and poison concentrations atdifferent reactors. Both neutron and gamma-raymeasurements also require axial profiles of thefuel assemblies to correct for burnup variations.Passive NDA methods are simpler than activemethods and may be more appropriate de-pending on the application.

Passive gamma-ray methods can be categorizedinto two areas: gross gamma counting andgamma spectroscopy. There are many ways inwhich gamma-ray spectroscopy has been

applied to identify fissile materials. Thesemethods rely on measuring spectral signaturesfrom fissile materials.

Passive neutron measurement methods can beused to estimate burnup or to provide anindication of the fissile mass in spent fuel. Thetwo main passive neutron measurements aretotal neutron counting and coincidence coun-ting.

Active NDA methods rely on gamma-ray orneutron sources to measure the attenuation ofradiation through spent fuel, to obtain a map-ping of the spent fuel in an assembly orcanister, and/or to obtain quantitative estimatesof the fissile content in spent fuel. Some activeNDA methods utilize the prompt radiationresulting from fission while others utilize theemission of delayed radiation from fission. Awide variety of sources exist for active NDAsuch as 252Cf , AmLi, PuBe, DT generators,electron accelerators, 57Co, etc. One benefit ofactive NDA methods is that the energy andintensity of the interrogating radiation can bechanged to emphasize the desired properties ofspent fuel.

Active gamma-ray methods employ photonsources that interact with the fissile materialand subsequent radiation measurements areperformed. Densitometry and X-ray fluores-cence are the two prominent active photonNDA methods. Photo-fission sources are an-other active photon NDA measurement alter-native.

Active neutron NDA methods utilize neutronsources to induce fission in fissile materials and

14


require the measurement of the subsequentprompt and delayed neutrons emitted from thefissile material. The fissile content of a materialcan readily be determined from many of thesemethods if calibration standards are available.These methods are all related and vary de-pending on what radiation is measured and howthe data is processed. The measurement methodmay analyze prompt neutrons, prompt gamma-rays, delayed neutrons, delayed gamma-rays, orthe neutron transmission in a fuel assembly.

Other NDA methods that can be considered forspent fuel assay include calorimetry and Ceren-kov radiation.

The 239Pu content of LEU Light Water Reactor(LWR) spent fuel is of primary importance tosafeguards, with the residual M5U content beingmuch less significant. An ideal NDA methodfor LEU fuel would then directly measure the239Pu content, separating it from all of the otherconstituents. Unfortunately, LEU spent fuelassemblies pose measurement difficulties un-matched by any other nuclear material. Fissionproducts in the spent fuel emit gamma-rays atan extremely high rate, making all but a fewgamma emission lines invisible in the Comptonbackground. It is not possible to measure theemission lines from Pu in spent fuel withpassive techniques. Similarly, the neutron emis-sion rate of LEU spent fuel is very large, and isdominated by spontaneous fission from Curiumisotopes, making direct passive neutron meas-urements of the spontaneous fission of pluto-nium isotopes impossible. Passive methods arethen limited to measuring attributes of the fuelthat can only be associated with the plutoniumcontent by reference to calculations. However,given acceptable knowledge of the history ofthe assembly, its initial enrichment, irradiationhistory, cooling time, and physical integrity,these indirect measurements can provide accu-rate knowledge of the Pu content. Activemeasurement on LEU fuel are also very chal-lenging, with the radiation emission of theassembly not only creating difficulty inachieving good signal-to-noise ratios, but alsomaking physical access to the fuel difficult.

While the 239Pu content in LEU fuel is thecomponent of most importance, the 235U contentbecomes more important in HEU spent fuel.HEU spent fuel will comprise at most a smallpart of the geologic repository total inventory,but in some ways may be a more attractivetarget for diversion, being smaller and lighterthan LWR fuel and typically having lowerradiation levels. NDA methods for measuringHEU fuel will be addressed separately fromLEU fuel. HEU spent fuel presents somewhatless difficulties for NDA measurements thanLEU spent fuel.

Measurement of spent fuel inside canisterscontaining multiple assemblies is also ad-dressed. Canister measurements pose a numberof additional difficulties above and beyond themeasurement of single assemblies. The burnup,age, and cooling times may vary amongst theassemblies in a single canister, in fact, it may bebeneficial in meeting the canister total heat loadand criticality limits to purposely mix low andhigh burnup fuel within a single canister.Proposed canisters vary widely in construction,materials, criticality control methods, and capa-city. The largest canisters may have 21 or morePWR assemblies, 40 or more BWR assemblies.Assemblies located in the interior of these largecanisters are effectively shielded by the exteriorassemblies.

5.2 Instruments for spent fuelverification

Section 5.1 above deals with methods andprinciples that under certain conditions can beused for verification of declared data of spentfuel assemblies or fuel rods. Those instrumentsand methods that are already in use or that areunder development on prototype level are listedin Annex 2. Annex 2 is meant to include onlymethods that have potential for verification ofspent fuel before or during conditioning endingin final deposition. Annex 2 includes also ameasurement point no. that relates to the flowchart of Figure 2.

15


Verification of closed disposal containers is amuch more demanding task than verification ofindividual assemblies. Even though all possiblemeasurement techniques are considered, onecannot find any method or technique that couldbe used as such for practical verification ofdeclared data of a closed disposal container.There are, however, a few R&D projects thatare related to verification of closed spent fuelcontainers including closed disposal containers.Such projects are described in Annex 3.

5.3 Fissile material tracking

Perimeter radiation monitors are used to moni-tor the transport of nuclear materials within acontrolled area. Perimeter monitors can be usedto monitor the regular path of nuclear materialswithin a controlled area or to detect un-authorized removal of nuclear materials from acontrolled area This section highlights thecurrently available perimeter monitors.

Continuous site tracking system

The emission of gamma radiation from spentfuel can be used as a signature to track andmonitor the movement of spent fuel within acontrolled area This system was originallydesigned to serve as a criticality accident alarmsystem [34] but could also serve as a fuel

tracking and monitoring system. The proposedsystem would be comprised of a set of gammasensitive detectors located on the ceiling of thearea to provide a two dimensional mapping ofthe spent fuel movement. This system has notbeen applied to spent fuel monitoring but itslow cost makes it an attractive method.

Unattended monitoring systems

Unattended continuous radiation monitoringsystems are in IAEA use at reactors worldwide[35]. These systems monitor fresh-fuel re-charge, the reactor core power level, spent fueldischarge, and spent fuel storage. The radiationsensors are operated in the continuous modewith data collection at intervals of one to twoseconds. The data is collected at inspectionperiods of one to three months, with datafiltering reducing the total amount of datastored to just those periods when nuclearmaterial is moving. In conjunction with con-tainment/surveillance systems and burnup veri-fication measurements, unattended monitoringsystems have the potential to provide conti-nuity-of-knowledge from fresh fuel loadingthrough spent fuel canister loading and storageat reactors and conditioning facilities. Effortsare currently underway to allow remote moni-toring of these systems to reduce inspectortravel and workload.

16


6 CONCLUSIONS OF EXISTINGNDA TECHNIQUES

6.1 Gross defect verification ofassemblies

Most of the NDA measurement methods avail-able are capable of performing gross defectverification of spent fuel. Gross defect test isalso the requirement most often needed by theIAEA for verification of spent fuel assembliesduring different phases of the fuel cycle.

Annex 2 includes detailed information of 18different measurement methods. All of themsatisfy the gross defect requirement i.e. theycan be used to verify that the assembly is notreplaced with a dummy or missing. The varietyof NDA methods for gross defect verification iswide and there is no need for major R&Defforts.

6.2 Partial defect verification ofassemblies

According to the IAEA Safeguards Criteria,partial defect testing may be needed before theidentity of fuel assemblies is lost (by cutting,dissolving, etc.). It is foreseen that partialdefect verification needs to be performed foreach fuel assembly arriving at the conditioningfacility in an active way, i.e. the measurementneeds to prove that all rods in each assemblyare present and that all rods include irradiatednuclear fuel. A schematic potential measure-ment system is shown in Figure 3 below.

In the input stage shown in Figure 3, eachassembly still maintaining its integrity, is meas-ured using a partial defect method. In essence,

Input

Presence of irradiated rodsverified by NDA

Process(rod dismantling)

Output

Can

JulILulu

Waste

Continuity-of-knowledge maintained by C/S

Figure 3. Verification of spent fuel assemblies in the conditioning facility.

17


the presence of all irradiated fuel rods isverified. If based on the measurements thedeclared data are questioned, the assembly canbe moved aside and clarification can be re-quested from the operator or shipper. If declareddata is confirmed to be correct and complete bythe measurement, the assembly can be pro-cessed further and its integrity may be lost.Continuity of the verified knowledge is main-tained by C/S methods.

Those NDA methods existing now or underdevelopment that are considered to have thepower needed for partial defect verification arediscussed briefly below.

Fork detector + HRGS

Passive gamma and neutron detection, if com-bined with high resolution gamma-ray spectro-metry, can be used for partial defect verificationif the method is calibrated properly. Themeasurement principle is based on the fact thatthe neutron signal depends linearly on theamount of material present. The neutron signaldepends strongly also on the burnup (power 3 -4). If material is missing, operator may declarea lower burnup yielding in a lower neutron rate.The problem can be solved by correlating themeasured neutron rate with the measured Cs-137 activity which depends linearly on theburnup and not on the amount of the materialpresent in an infinitely thick object like a fuelassembly. Different calibration curves are, how-ever, needed for different fuel types and initialenrichments.

High energy gamma emission tomography

Passive gamma detection both under wet anddry conditions can provide a sensitivity level of1 missing rod for BWR assemblies and anestimated sensitivity level of 1-5 missing rodsfor large PWR assemblies, respectively. Thetomographic methods is superior to any othermethod in the respect that no reference infor-mation is needed. Any assembly can be inde-pendently verified on partial defect level with-out knowing any declared data.

Californium source driven noisemeasurement

The active neutron interrogation technique (Cf-252) can be used for measurement of the totalfissile content with a high precision if appro-priate reference information is available. Onlydry assemblies can, however, be measured.

Python

The active mode of the method allows verifi-cation of the residual fissile contents to beverified. Calibration for different fuel designsand measurement conditions is, however,needed.

6.3 Radiation monitoring ofcasks

Radiation monitoring systems are used as asupporting measure, backing up camera surveil-lance. Such radiation detectors will use simpleneutron or gamma measurement, as appropriate,to facilitate the distinction between full andempty casks as well as indicating the directionof movement of casks. No NDA methods areavailable that can be used to quantitativelydistinguish between spent fuel and high levelwaste. In addition, radiation detection systemscan be applied, which are suitable to dis-tinguish, at least qualitatively, casks containingonly high level waste from casks containingspent fuel.

There are no readily available non-destructiveassay technique that can be used for quantitativeverification of casks at the input of an operatingrepository. However, for unshielded and/orshielded containers, there are research anddevelopment projects that could help resolvethis issue (Annex 3).

The availability of quantitative verificationtechniques for casks would have a great impacton the safeguards approach. In practice, nomeasurements can easily be made under theground level. Only the surface facilities mayhave locations for such activities.

18


6.4 Unsolved NDA problems * N ° A for measurement of fissile materialthrough the cask wall.

The following verification needs include un- • NDA for measurement of Pu in spent fuelsolved NDA problems and require more R&D assemblies,efforts in the future: • NDA for partial defect testing of (un-• Measurement system for routine partial shielded) multi-assembly configurations,

defect tests of spent fuel assemblies. • Verification of shielded/unshielded casks.• Distinguishing canisters containing HLW

from canisters containing spent fuel (incase of mixed emplacement).

19


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[2] Larrimore J. IAEA Safeguards Criteria1991-1995: Where do they stand in1996?. In: Proc. 39th Annual Meeting,INMM, 1996.

[3] IAEA Safeguards Glossary, 1987 Edition,Report IAEA/SG/INF/1 (Rev. 1), Vienna1987.

[4] Levai F, Desi S, Tarvainen M, Arlt R.Detection of missing rods in a spent fuelassembly by computed gamma emissiontomography. In: Proceedings of the 32ndINMM Annual Meeting, New Orleans,USA, 1991: 1012-1017.

[5] Levai F, Desi S, Tarvainen M, Arlt R. Useof an underwater multidetector systemfor gamma emission tomography of spentfuel assemblies. In: Proceedings of the15h ESARDA Symposium, Rome, Italy,1993:387-392.

[6] Levai F, Desi S, Tarvainen M, Arlt R. Useof high energy gamma emission tomo-graphy for partial defect verification ofspent fuel assemblies. Final Report onthe Task FIN A98 of The Finnish SupportProgramme to IAEA Safeguards. Novem-ber 1993, Report STUK-YTO-TR 56.

[7] Levai F, Tarvainen M. Partial defecttesting of spent fuel for final disposalusing tomography. In: ANS Transactions,Washington 1995

[8] Mihalczo JT, Valentine TE, MattinglyJK. Feasibility of Subcriticality and NDAMeasurements for Spent Fuel by Fre-quency Analysis Techniques with 252Cf.In: Proc. Int. Top. Meeting on Nucl. PlantInstrumentation, Control, and Human-Machine Interface Tech., May 6-9, 1996,Pennsylvania State University, USA.

[9] Mattingly JK, Valentine TE, MihalczoJT. Feasibility of Fissile Mass Assay ofSpent Nuclear Fuel Using 252Cf-Source-Driven Frequency Analysis. In: Proc. ofInstitute of Nucl. Materials ManagementAnnual Meeting, July 1996, Naples,Florida.

[10] Valentine TE, Mattingly JK, MihalczoJT. Dry Spent Fuel Cask Monitoring by252Cf-Source-Driven Frequency AnalysisMeasurements. In: Proc. of Institute ofNucl. Materials Management AnnualMeeting, July 1996, Naples, Florida.

[11] Abdurrahman NM, Block RC, HarrisDR, Slovacek RE, Lee YD, Rodriguez-Vera F. "Spent-Fuel Assay Performanceand Monte Carlo Analysis of the Rensse-laer Slowing-Down-Time Spectrometer,"Nucl. Science and Engineering (115)1993:279-296.

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[12] Yong-Deok Lee et al. Neutron emissiontomography for nuclear fissile materialsafeguards.

[13] Abdurrahnan NM et al. Detection sensi-tivity of an LS DTS for Spent-fuel FissileAssay Application, ANS Transactions(68) 1993: 65-66.

[14] Yong-Deok Lee et al. A Practical Spent-fuel Assay Device Using the Lead Spec-trometer. ANS Transactions (68) 1993:66-67.

[15] Nilsson A, Danielsson N, Chen JD,Young GJ, Vodrazka P, Nakaoka. "Ceren-kov Light Images of LWR Fuel Assem-blies," Proceedings of the 30th annualMeeting of the Institute of Nuclear Mate-rials Management, July 1989, Orlando,Florida, USA, vol. XVHI: 902-909.

[16] Chen JD, Attas EM, Young GJ, BurtonGR, Keeffe R, Ward-Whate HildingssonP, Nilsson A, Trepte O. "Spent FuelVerification Using an Ultraviolet Sensi-tive Charge Coupled Device," Proc. ofthe 37th annual Meeting of the Instituteof Nuclear Materials Management, July1996, Naples, Florida, USA.

[17] Rinard PM, Bosler GE. "SafeguardingLWR Spent Fuel with the Fork Detector",Los Alamos National Laboratory, LA-11096-MS, (March 1988).

[18] Reilly D, Ensslin N, Smith H Jr, KreinerS. Passive Nondestructive Assay of Nu-clear Materials, U. S. Nuclear Regulato-ry Commission, NUREG/CR-5550, LA-UR-90-732, 1991.

[19] Sher R, Untermyer II S. The Detection ofFissionable Materials by Non-destructiveMeans, American Nuclear Society, LaGrange, Illinois, 1980.

[20] Hsue ST, Crane TW, Talbert WL Jr, LeeJG Nondestructive Assay Methods forIrradiated Nuclear Fuels, LA-6923,1978.

[21] Prettyman TH, Betts SE, Taggart DP,Estep RJ, Harlan RA. "Field Experiencewith a Mobile Tomographic Nondestruc-tive Assay System," Proc. of the 4th

Nondestructive Assay and Nondestruc-tive Examination Waste CharacterizationConference, Los Alamos National Lab-oratory Report LA-UR-95-3501, (Oct.1995).

[22] Gozani T. Active Nondestructive Assay ofNuclear Materials. NUREG/CR-0602,Jan. 1981.

[23] Untermyer II S. Development and Test ofMethods for the Nondestructive Assay ofSpent Fuel Assemblies, EPRI NP-2812,1983.

[24] Bosler GE, Halbig JK, Klosterbuer SF,Menlove HO, Phillips JR. "Passive Neu-tron Measurement Applications for Ir-radiated Fuel Assemblies," Trans. Am.Nucl. Soc. (39) 198: 348.

[25] Menlove HO, Reilly TD, Siebelist R."The Verification of Reactor OperatingHistory Using the FORK Detector", Proc.of Institute of Nucl. Materials Manage-ment Annual Meeting, (July 1996).

[26] Menlove HO, Stewart JE, Qioa SZ, WenzTR, Verrecchia GPD. "Neutron CollarCalibration and Evaluation for Assay ofLWR Fuel Assemblies Containing Burn-able Neutrcn Absorbers," Los AlamosNational Laboratory, LA-11965-MS,(November 1990).

[27] Menlove HO, Tesche CD, Thorpe MM,Walton RB. "A Resonance Self-Indi-cation Technique for Isotopic Assay ofFissile Materials", Nuclear Applications,Vol. 6, (April 1969).

21


[28] Wuerz H. "A Nondestructive Measure-ment System for Spent LWR Fuel As-semblies", in Proc. of the InternationalSymposium on Recent Advances in Nu-clear Material Safeguards, IAEA reportIAEA-SM-260, (November 1982).

[29] Abhold ME, Hsue ST, Menlove HO,Walton G, Holt S. "The Design andPerformance of the Research ReactorFuel Counter", Proc. of Institute of Nucl.Materials Management Annual Meeting,(July 1996).

[30] Pare VK, Mihalczo JT. "Reactivity fromPower Spectral Density Measurementswith Californium-252," Nucl. Sci. Eng.(56) 1975: 213.

[31 ] Pickrell MM, Kendall PK. "The Synchro-nous Active Neutron Detection Systemfor Spent Fuel Assay," presented at theI&EC Special Symposium, AmericanChemical Society, (Sept. 1994).

[32] Eccleston GW, Menlove HO, Echo MW."A Measurement Technique for HighEnrichment Spent fuel Assemblies andWaste Solids," Nuclear Materials Man-agement 8, (1979).

[33] Cole JD, Aryaeinejad R, Drigert MW."Gamma Neutron Assay Technique", pre-sentation at the National Spent FuelProgram, Nondestructive Assay Meeting,February 7-8, 1996 (unpublished).

[34] Mihalczo JT, Hutchinson DP, WilliamsJA, Thacker LH. "Criticality EvacuationDetectors That Locate the Accident,"Trans. Am. Nucl. Soc. (75) 1996: 184-186.

[35] Klosterbuer SF et. al. "Continuous Re-mote Unattended Monitoring for Safe-guards Data Collection Systems," inProceedings IAEA Symposium on Inter-national Safeguards 1994: Vision for theFuture, IAEA, Vienna, (1994).

[36] Bosler GE, Phillips JR, Wilson WB,LaBauve RJ, England TR. "Production ofActinide Isotopes in Simulated PWRFuel and Their Influence on InherentNeutron Emission", LANL Report LA-9343, July 1982.

[37] Phillips JR, Bement TR, Hatcher CR,Hsue ST, Lee DM. "Nondestructive Veri-fication of the Exposure of Heavy-Waterreactor Fuel Elements", LANL ReportLA-9432-MS, June 1982.

[38] Rinard P. "A Spent-Fuel Cooling Curvefor Safeguards Applications of Gross-Gamma Measurements", LANL ReportLA-9757-MS, April 1983.

[39] Bosler GE, Rinard PM, Klosterbuer SF,Painter J. "Automated Methods for Real-Time Analysis of Spent-Fuel measure-ment Data", In: Proceedings of the 29thAnnual INMM Meeting, Vol. XVII, June26-29, 1988.

[40] Nelson AJ, Bosler GE, Augustson RH,Cowder LR. "Underwater Measurementof a 15x15 MOX PWR-Type Fuel As-sembly", LANL Report LA-11850-MS,December 1990.

[41] Hsue S-T, Menlove HO, Bacca J, BoslerGE, Dye HR, Walton G, Halbig J,Siebelist R. Research reactor Fork UsersManual, LANL Report LA-12666-M.

[42] Chesterman AS, Clark PA. Spent fuel andresidue measurement instrumentation atthe Sellafield nuclear fuel reprocessingfacility, In: INMM 36th Annual MeetingProceedings, Palm Desert, California,July 9-12, 1995: 185-192.

[43] Nicolau G, Koch L, "Characterization ofSpent Nuclear Fuel by Non-destructiveAssay", In: Proceedings of the 5th Int.Conf. on Radioactive Waste Measure-ment and Env. Rem., Berlin 3-7 Sep-tember, 1995.

22


[44] Tarvainen M, Paakkunainen M, Tiitta A, [48]Sarparanta K. "BWR SEAT, gross-defectverification of spent BWR fuel", ReportSTUK-YTO-TR 72, STUK, Finland,April 1994.

[45] Tarvainen M, Backlin A, Hakansson A."Calibration of the TVO spent BWRreference fuel assembly", Report STUK-YTO-TR 37, STUK, Finland, February1992.

[46] af Ekenstam G, Tarvainen M. "Indepen- [49]dent burnup verification of BWR-typenuclear fuel by means of the 137-Csactivity", Report STUK-A52, STUK,Finland, June 1987.

[47] Hildingsson L, af Ekenstam G, TarvainenM, Tiitta A. "Feasibility of Gamma RayVerification of Non-standard Fuel Itemsat CLAB", Report SKI 96:23, SKI,Sweden, March 1996.

Eccleston GW, Johnson SS, MenloveHO, Van Lyssel TV, Black D, Carlson B,Decker L, Echo MW. "FAST FacilitySpent-Fuel and Waste Assay Instrument",In: Proc. of the Conference on SafeguardsTechnology. The Process-safeguards In-terface, Hilton Head Island, South Caro-lina, November 28-December 2, 1983,USDOE, New Brunswick Laboratory,Argonne, Illinois, CONF-831106, August1984: 253-264.

Piper TC, Kirkham RJ, Eccleston GW,Menlove HO. "Analysis of Spent, HighlyEnriched Reactor Fuel by delayed Neu-tron Interrogation", In: Proc. of Int.Topical Meeting on Safety Margins inCriticality Safety, San Francisco, Cali-fornia, November 26-30, 1989, ANS,1989:168-175.

23


ANNEX 1

Some of the techniques included in this reviewdo not have any potential for application tospent fuel measurement, but nevertheless havebeen included for completeness. Measurementmethods have been identified that are underdevelopment that could further aid in thecontrol and safeguards of spent nuclear fuel,and where additional development is necessaryfor practical application, it will be so noted. TheNDA methods have been categorized intoseveral distinct areas: passive and active gam-ma, passive and active neutron, and other NDAtechniques. In addition, fissile material trackingmethods are discussed. Passive and active NDAtechniques are those methods which can verifythe presence of spent fuel or provide a quanti-tative measure of the uranium and/or plutoniumcontent in spent fuel. Passive and active NDAmethods are both useful depending on the typeof information needed. Fissile material trackingmethods employ measurement systems that canaid in tracking spent fuel in a conditioningfacility or repository site. This article summa-rizes many NDA methods and specifies theirlimitations and the specific spent fuel meas-urement areas in which these methods may beapplied.

Al.l Cerenkov

Measurements of Cerenkov light emanatingfrom a spent fuel assembly in water has beensuccessfully used to verify spent fuel as-semblies in storage pools [15]. High energyelectrons, gamma-rays, and neutrons from spentnuclear fuel are capable of producing Cerenkovlight directly or indirectly in storage pools.Cerenkov viewing devices (CVD) are used toidentify physical characteristics of spent fuelassemblies and to provide an indication of theburnup of an assembly in storage pools. Thesedevices are not capable of identifying missingfuel rods or the substitution of dummy fuel rodsin a spent fuel assembly. The absolute Cerenkovlight level and its decay with time is related tospent fuel burnup and it may provide a possi-bility for quantitative measurements of theglow from spent fuel [16]. In practice, however,CVD verification is a single measurement, soonly consistency of the declared burnup and

DESCRIPTION OF NDA METHODS

cooling time with the Cerenkov radiation can bechecked. The effectiveness of this method isreduced for verification of long cooled and/orlow burnup assemblies. This technique is onlyapplicable for wet storage of individual rods orassemblies, and does not have potential formeasurement of canisters.

A 1.2 Passive gamma-raytechniques

Total counting

The total gamma-ray activity of spent fuelassemblies can be measured with scintillators,thermoluminescent dosimeters, or ionizationchambers as is done by the FORK [17, 24, 25,38, 45] detector which the IAEA commonlyuses for spent fuel verification. For coolingtimes greater than 1 year, the total gamma-rayactivity is roughly proportional to the burnup.Consistency of operator declared values forburnup and cooling time can be determined byrepeat measurements over several months time.These measurements provide an indirect signa-ture of the fuel burnup and age, which can berelated to Pu content by calculation.

Total gamma counting has limited potential forquantitative assay of fuel canisters. The gammaattenuation of the canister and contents wouldhave to be accurately calculated, and mixturesof low and high burnup fuel inside the samecanister would result in a non-unique solutionfor the average calculated Pu content. However,total gamma counting would provide a valuablecomponent of a radiation signature, and couldbe used to help verify the integrity of a canisterwith prior knowledge of its content.

Spectroscopy

Gamma-ray spectroscopy is a common tech-nique used to characterize radioactive materials[18-20, 44—47]. The decay of radioactive iso-topes results in the emission of gamma-rays andx-rays that are characteristic of the radioactiveisotope. In spent fuel applications the spectrallines are used to verify the presence of fissionproduct actinides. For example, absolute ac-

24



tivity measurements of 137Cs [45] can beperformed on spent fuel rods and assemblies toget a qualitative measure of the spent fuelburnup; however, this requires the assemblyattenuation and measurement geometry to beknown or precisely calculated. Subsequentcalculations can then be performed to obtainfissile mass content. The burnup of spent fuelcan also be obtained from the ratio of activitiesfor some fission product isotopes. The ratio of134Cs to 137Cs is nearly linear with burnup;however, corrections must be made for thedecay of the isotopes for long burned fuels. Thegamma activity ratio method must also becorrected for gamma-ray attenuation in as-semblies. Fission product activity ratios areeasier to determine than absolute activities forfield measurements because only ratios of thedetector efficiencies, which are a function ofgamma-ray energy, need to be known. Due toattenuation affects these methods are mostlikely not applicable for canisters or casks ofspent fuels.

High energy gamma emission tomography(passive gamma)

A tomographic measurement system has beendeveloped and built for cross sectional viewingof spent fuel assemblies and multi-elementbottles (MEB) [4-7]. The method can reveal therod structure of spent fuel assemblies and thepresence of fuel assemblies in closed MEBs.The hardware developed allows measurementof the emission of fission product gamma-rays(Eu-154 and Pr-144), in the air or under water.The cross sectional image of a spent fuelassembly gives a rod-to-rod distribution ofgamma emitter concentration in the section ofspent fuel mapped. The high sensitivity to theremoval of irradiated rods is explained by thefollowing facts: (i) No need for a reference dataset, because the activity map provides aninherent rod-to-rod comparison of fission pro-duct activities, (ii) the inherent feature of theimage reconstruction method provides noisereduction by averaging measured data obtainedat different views.

ANNEX 1

The method has additional capabilities ofmeasuring assemblies of new design, e.g. axialenrichment variations and MOX. Gamma-raymeasurement can be made on the side of anirradiated fuel assembly that is partially raisedfrom the storage rack or moved to a meas-urement position.

A1.3 Active gamma-ray techniques

Densitometry

Densitometry entails the use of a photon sourceand a detector to measure the attenuation of thephotons from the source [18,19]. The photonsource may emit particles at a single energy ormultiple energies. These measurements areuseful for scanning rods and assemblies toidentify burnup variations and holes in as-semblies. This method may not be useful forquantitative measures because it measures thebulk attenuation of all materials in the rod andis best used as a means to map out holes in fuelassemblies. The applicability of this method islimited mainly to individual fuel rods andsingle fuel assemblies because canisters orcasks will shield the assembly from the source.

X-ray fluorescence

X-ray fluorescence can be used to get aqualitative and quantitative assay of fissilematerial [18, 22]. X-rays emitted from anionized atom have energies that are cha-racteristic of the element. The X-ray intensity isdependent on the concentration of the elementand the intensity of the ionizing source. Thismethod cannot discriminate between isotopesof the same elements because X-rays originatefrom electron transitions and not as a result of anuclear process. Because of the low energy ofX-rays this method may only be practicallyapplied to assaying a single fuel rod. The highbackground radiation from a spent fuel assem-bly or canister and the complicated attenuationfor X-rays would make measurements in suchgeometry impossible.

25

RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR133

ANNEX 1

Photon induced fission

This method requires the use of high energyphotons from an accelerator to induce fission inuranium and plutonium materials and meas-uring the subsequent emission of prompt anddelayed neutrons from the fissile material [19].However, the method is limited in that it isdifficult to discriminate between different fissilematerials and there is interference from (y,n)reactions in the materials. This method has notbeen applied to a spent fuel assay.

A 1.4 Passive neutron techniques

Total counting

The total neutron emission rate from spent fuelcan serve as an indicator of the burnup of thespent fuel [18, 19, 23, 24]. The total neutroncounting method has some advantages anddisadvantages when compared to gamma-rayspectroscopy. Neutrons emitted from the spentfuel are not attenuated as much as gamma-rays,and neutron counting measurements can bemade soon after the fuel is removed from thecore whereas gamma-ray measurements cannotbecause short lived decay products dominatethe gamma-ray emissions.

Two primary neutron-emitting isotopes are244Cm and 242Cm. To use total neutron countingmeasurements the cooling time of the fuel mustbe precisely known because of the short half-lives of these isotopes. The initial 235U enrich-ment and the irradiation history are factors thatcan affect the interpretation of the results oftotal neutron counting measurements. Veri-fication of reactor operating history can beaccomplished by total neutron counting inconjunction with total gamma counting as inthe modified FORK detector [25].

Coincidence counting

Passive neutron coincidence counting relies onthe presence of spontaneous fissile isotopes infissile materials to produce time correlated


neutrons [18-20]. A number of passive co-incidence measurement have been developedthat rely on the time correlation of radiationfrom spontaneous fission. Most spontaneousfissile isotopes yield on average more than twoprompt neutrons and more than five promptgamma-rays emitted essentially simultaneously.These measurements can be made in thepresence of background radiation because thebackground radiation is random in nature.Passive coincidence measurements are suscep-tible to self multiplication effects that obscurethe actual amount of material This self-multiplication effect can be significant for spentfuel systems with high (a,n) rates because the(cc,n) neutrons may induce fission in the fissilematerials. This method has been applied tofresh fuel [26], but has not been applied to spentfuel.

Neutron albedo

Spent fuel assemblies emit spontaneous fissionand (a,n) neutrons at a high rate. Some of theseneutrons moderate in the water surrounding theassembly and return to further induce fission inthe assembly. If cadmium is placed around theassembly, these returning neutrons, or albedoneutrons, are absorbed in the cadmium. Meas-urements made with and without a cadmiumliner can be used to determine the multiplicationin the assembly resulting from the albedoneutrons, therefore the fissile mass can beestimated. This method does not have potentialto assay canisters or casks.

Multiplicity

Multiplicity analysis is an extension of co-incidence counting, but in addition to doubles,two or more neutrons detected coincident intime, three or more coincident neutrons areobserved. This higher order correlation providesadditional information that can be useful inseparating the fission neutron source, (a,n)neutron sources, and the effect from the assem-bly's intrinsic multiplication. Since spent fuelassemblies in water have relatively high re-

26



activity, there is a substantial probability of aneutron causing a fission chain, or super-fission, which can release large numbers ofnearly simultaneous neutrons. These super-fissions can be seen in the neutron multiplicitydistribution, and yield valuable fissile contentinformation. This method does not have poten-tial to assay dry canisters, which have relativelylow reactivity and complicated geometry.

Neutron resonance absorption

All of the passive neutron techniques discussedso far only have the capability to indirectlyindicate the Pu or total fissile content in spentfuel by correlation with calculations. Neutronresonance absorption techniques have the po-tential to passively separate the Pu content inLEU fuel from other fissile isotopes. Thismethod utilizes the resonance structure in theneutron fission cross sections and relies ondetectors that are sensitive to the resonanceabsorption lines [27]. This method has thepotential to verify Pu content in the outer pinsof an assembly underwater, with further devel-opment, however does not have potential forassay of canisters.

A1.5 Active neutron techniques

Total counting

Active total neutron counting uses a moderatedneutron source to induce fission in a fissilematerial and a detection system to measure theemission of prompt neutrons from the material[18, 19]. Prompt neutrons can be distinguishedfrom source neutrons using energy discrimi-nation. Moderated neutrons sources have beenused to scan individual fuel rods and measurethe prompt neutrons emitting from the rod withfast neutron detectors. This methodology maybe applied to spent fuel assemblies but caremust be taken to discriminate the promptneutrons from background radiation. An ad-ditional measurement would be required toestimate the contribution of the backgroundradiation to the detector response.

ANNEX 1

An assembly monitor based on combined activeand passive neutron counting has been used todetermine the burnup and initial enrichment ofspent LWR fuel [28]. This method can providean indirect indication of Pu content throughisotopic correlation.

Active coincidence counting

Active coincidence counting systems employ aneutron source to induce fission in fissilematerial and measure the subsequent emissionof prompt radiation from the material [18, 19,22]. The emission of prompt radiation occursessentially simultaneously and provides a me-chanism to time correlate the detector response.Coincidence counting measurements are typi-cally not affected by the neutron source and areindependent of the background radiation from(a, n) sources but are affected by the inherentspontaneous fission rate in a spent fuel assem-bly. By removing the neutron source, a passivecoincidence count can be made to estimate thecontribution of the inherent spontaneous fissionsources. A profile of the fissile mass content ofa spent fuel assembly could be obtained byperforming a vertical scan of the assembly. Thismethod has been applied to the measurement ofHEU spent fuel assemblies [29].

Noise analysis

The noise analysis method [30] uses a 252Cfsource in an ionization chamber or a neutrongenerator to induce fissions in a fissile materialand two or more detectors to measure theemission of prompt neutrons and gamma-raysfrom the fissile material. The source ionizationchamber or neutron generator provides a timedsignal for each source event (99.9% of thespontaneous fissions are typically counted for1 u.g 252Cf sources). The source and detectorsignals are correlated with each other to obtainsignatures that depend only on the inducedfission rate in the fissile material. The detectorsignals are also correlated with each other toobtain signatures that depend on both theinherent and induced fission rates in the fissile

27


ANNEX 1

material. The method has an advantage in thatthe correlation between the source and thedetector is independent of the inherent sourcesand background radiation and depends only ondetector efficiency to the first power. Othercorrelation and coincidence techniques dependon detection efficiency to the second powerresulting in longer measurement times andrequire a passive measurement to estimate thecontribution of inherent spontaneous fissionsources to the detector response. Feasibilitystudies have been performed to assess theapplication of this technique to spent fuel assay.These feasibility studies have shown that thecorrelation between the source and detector isnearly linear with fissile mass content for asingle fuel assembly [8, 9, 34] and for anunshielded canister of spent fuel [10] and thatthis method could be used to scan a fuelassembly to obtain the profile of the fissilemass.

Pulse neutron—prompt

Pulsed neutron methods have been used toinduce fission in fissile materials to measure thesubsequent emission of delayed neutrons afterthe pulse [19, 22]. These methods require theuse of a pulsed accelerator, a movable source,or a shutter device to pulse the fissile material.The detection system is triggered on after thepulse to reduce the contribution of sourceneutrons to the detector response. Repetitiveburst of neutrons would be required to obtainestimates of the total fissile mass in the spentfuel assembly. Neutron source strengths mustbe on the order of 108 or more neutrons persecond to induce a fission signal that iscomparable in size to the passive neutronbackground from a spent assembly. Novelmethods have been investigated to improve thesignal to background ratio using pulsed promptsources. A factor of 4 improvement has beenshown by a synchronous active technique usinglock-in amplifiers [31]. These pulsed neutronmethods can potentially determine the fissilecontent in an assembly and, with a large enoughsource, can be used to assay canisters of fuel.


Pulse neutron—prompt/delayed

Methods have been employed to assay HEUspent fuel based on Cf shuffler techniques [32].The technique employs fast neutron irradiationof the fuel by a Cf source followed by delayedneutron counting after the source is transferredto storage. Since the delayed neutron yield issignificantly different for U and Pu fission, thismethod has the potential to separate the fissileisotopes. There are presently no pulsed systemsin operation measuring LWR fuel assemblies.Delayed neutron yields are only a small fractionof the prompt yield, therefore, measurement ofLWR fuel would require extremely large neu-tron sources, on the order of 10" neutrons persecond, or neutron sources capable of producingextremely large bursts of neutrons. Develop-ment of such sources is underway, with proto-type sources expected within two years. Thismethod has the potential to determine the Pucontent in assemblies and canisters.

Neutron radiography

Neutron radiography uses a source of thermal,epithermal or fast neutrons, and the penetratingqualities of neutrons are used in the study ofdistribution of material variations. The image oftransmitted neutrons can provide information ofthe neutron attenuation property of materials.Special features of the techniques are that (i)uranium and each plutonium isotopes havedifferent absorption properties for thermal neu-trons, (ii) epithermal and fast neutrons arehighly penetrating radiations capable of imag-ing thick objects and (iii) pure neutron imagemay be obtained even in case of gamma activespent fuels. The method is mainly used forimaging single spent fuel rods, especially forpost irradiation examination purposes.

Neutron resonance absorption

Neutron resonance absorption is a neutrontransmission measurement that uses a neutrontime of flight spectrometer to identify reso-nances in fissile materials [20]. There is little

28



interference from fission product resonances inthe time of flight spectra This method may beapplied to a single fuel rod but is not practicalfor a fuel assembly because of the assemblythickness and geometry.

Another active method for assay of spent fuelthat relies on resonance absorption is theslowing down time spectrometer [11]. Thismethod currently relies on a Linac to provide aburst of neutrons that slow down by scatter withlead. The burst neutrons as they slow downpasses through distinctive resonances in thefuel, producing fast fission neutrons in theprocess. The fast neutrons are detected bythreshold detectors and the time history of thedetected neutrons is related to the isotopiccontent This method has the potential toseparate Pu from U, and may be able to assayLWR fuel assemblies. The development of apractical burst neutron source is essential to theapplication of the concept. Efforts are underwayto develop such sources in the next severalyears. This method does not have potential toassay canisters of fuel.

A1.6 Other techniques

Gamma neutron assay technique (GNAT)

This technique is a combination of activeneutron interrogation with the measurement ofthe prompt gamma emission from fission reac-

ANNEX1

tions [33]. The gamma emission from thefission fragments are coincidence counted toreduce the gamma background from the other(non-correlated) gamma sources. A time gate isopened by triggering on the arrival of a specificgamma-ray from one of the two fission frag-ments. Gamma spectra are then obtained in anarrow time window, which enables manygamma lines to be discriminated from thebackground radiation. Among the gamma linesthat can be seen in this window are the linesfrom the partner fragment. The analysis of thespectra yields the isotope of the fissile material.Count rate limitations in the high resolutiongamma spectroscopy system may effectivelylimit the efficiency of this technique, howeverquantitative assay of the Pu content in spentfuel may be possible with further development.

Calorimetry

Calorimetry is the quantitative measurement ofthe heat produced by a sample. Calorimetry canbe used to measure the heat production rate ofradionuclides, and is in common use measuringPu and tritium. The total heat produced by aspent fuel assembly or canister is a function ofthe initial fissile mass and the burnup. Multiplecalorimetric measurements of the spent fuelover time can be used to confirm characteristicsof spent fuel, and can be used on canisters andcasks.

29


ANNEX 1 DESCRIPTION OF NDA METHODS

A1.7 Table Al.I, NDA methods for spent HEU fuelTable ALL Nondestructive measurement methods applicability to spent fuel in a conditioning facility,high-enriched uranium fuel.

Technique

Cerenkov

Passive gamma

Total Counting

Spectroscopy

Tomography

Active gamma

Densitometry

X-ray fluorescence

Photon inducedfission

Passive neutron

Total Counting

Coincidence Counting

Neutron Albedo

Multiplicity

Resonance Absorption

Active neutron

Total Counting

Active Coincidence

Noise Analysis

Pulse NeutronPrompt

Pulse NeutronPrompt/Delayed

Neutron Radiography

Neutron ResonanceAbsorption

Gamma NeutronActive (GNAT)

Calorimetry

Fuel Rod

Indirect

Indirect

Indirect

Indirect

Qualitative

Qualitative

N/A

N/A

N/A

N/A

N/A

N/A

Direct TotalFissile

Direct TotalFissile

Direct TotalFissile

Direct TotalFissile

N/A

Qualitative

N/A

Direct

Indirect

FuelAssembly

Indirect (D)

Indirect

Indirect

Indirect

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

Direct TotalFissile

Direct TotalFissile

Direct TotalFissile

Direct TotalFissile (D)

N/A

Qualitative

N/A

Indirect (D)

Indirect

Low CapacityCanister

Qualitative

Qualitative

Qualitative

Indirect

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A





N/A

Qualitative (D)

N/A

N/A

Qualitative

High CapacityCanister

Qualitative

Qualitative

Qualitative

Qualitative

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

Qualitative

Indirect



N/A

Qualitative (D)

N/A

N/A

Qualitative

Cask Limiting(Shielded) Conditions

N/A Wet only

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

Qualitative(D)

Qualitative(D)

N/A

N/A *)

N/A

N/A

Qualitative

*) = State of art uses accelerator facilitiesQualitative = Able to measure indications of characteristics consistent with spent nuclear fuel.Indirect = Able to measure an attribute(s) of the item that permits calculation of the plutonium or uranium based onassumptions.Direct total fissile = Direct quantitative measurement of the total fissile content.Direct = Direct quantitative measurement of the uranium or plutonium in the item.(D) = Technique requires development or testing

30


DESCRIPTION OF NDA METHODS ANNEX 1

A1.8 Table Al.II, NDA methods for spent LEU fuelTable A1.II. Nondestructive measurement methods applicability to spent fuel in a conditioning facility,low-enriched uranium fuel.

Technique

Cerenkov

Passive gamma

Total Counting

Spectroscopy

Tomography

Active gamma

Densitometry

X-ray fluorescence

Photon inducedfission

Passive neutron

Total Counting

Coincidence Counting

Neutron Albedo

Multiplicity


Active neutron

Total Counting

Active Coincidence

Noise Analysis

Pulse NeutronPrompt

Pulse NeutronPrompt/Delayed

Neutron Radiography


Gamma NeutronActive (GNAT)

Calorimetry

Fuel Rod

Indirect

Indirect

Indirect

Indirect

Qualitative

Qualitative

N/A

Indirect

Indirect


Indirect

Direct (D)

Direct TotalFissile

Direct TotalFissile


Direct TotalFissile

Direct (D)

Qualitative

Direct (D)

Direct

Indirect

Fuel Assembly

Indirect (D)

Indirect

Indirect

Indirect

N/A

N/A

N/A

Indirect

Indirect


Direct TotalFissile (Wet)Indirect (Dry)

N/A (Dry)Qualitative (Wet)

Direct TotalFissile




Direct (D)

Qualitative

Direct (D)

Indirect (D)

Indirect

Low CapacityCanister

Qualitative

Qualitative

Qualitative

Indirect

N/A

N/A

N/A

Indirect

Indirect

N/A

Indirect

N/A


?



Direct (D)

Qualitative

N/A

N/A

Qualitative

High CapacityCanister

Qualitative

Qualitative

Qualitative

Qualitative

N/A

N/A

N/A

Indirect

Qualitative

N/A

Qualitative

N/A

Qualitative

N/A




Qualitative

N/A

N/A

Qualitative

Cask Limiting(Shielded) Condition

N/A Wet only

N/A

N/A

N/A

N/A

N/A

N/A

Qualitative

NA/

N/A

N/A

N/A

N/A

N/A

Qualitative(D)

Qualitative(D)

Qualitative(D)

N/A

N/A

N/A *)

Qualitative

*) = State of art uses accelerator facilitiesOther markings as in Table A1.1.

31

U I No. Meas.Meas. principlePoint # Instn/method name

Description(Techniques, handling, attribute verified, application)

Possibilities and limitations(level of verification)

JO

1 1 Passive gamma

LWR SFAT Spent FuelAttribute Tester

Gamma Taucher(Euratom)

Passive neutron,passive gamma

GRAND/ForkDetector

Shielded gamma spectrometer attached to air-filledcollimator pipe suspended over fuel element underwater.Manipulation using fuel handling machine (ormanually).Collects gamma-ray spectrum of fission products usingNa(I) or CdZnTe detectors and standard measurementelectronics. Used normally in point-wise mode,scanning mode also possible.Measurement time in point-wise mode 30-200 sec,20-40 assemblies per hour depending on measurementconditions.Attribute verified: Fission product gamma-rays e.g.Cs-137,Pr-144.Application: Verification of spent fuel specificgamma-ray emissions, raw estimation of burnup andcooling time.

Gross detect verification of spent LWR fuel.Distinguishes irradiated fuel from irradiatedmetal and dummies by detecting spent fuelspecific radiation (finger prints).No fuel movements needed, verification basedon the top part of the assembly.No practical limitations of cooling time orburnup.Cheap to construct, maintain and use andmodify for different facility conditions.Assemblies with very short cooling time maybe difficult to verify if located next toassemblies with long cooling time and/or lowburnup.

U-shaped polyethylene fork with two fission chambers(neutrons) and one ionization chamber (gamma-rays) ineach tine. Cd-wrapped fission chamber for fast and barefission chamber for thermal neutron detection. Fork isplaced around assembly to obtain neutron and gammaemission data.Fork is used from mobile bridge or fixed to a wall forunderwater measurements. Form a bridge, assembliesdo not need to be totally removed from the storage rack.Measurement time 30-60 sec, 6-12 assemblies perhour.Attribute verified: Gross neutron emissions, mainlyfrom even Cm isotopes, gross gamma detection offission and activation products.

Gross defect verification of spent fuel.Consistency measurement of burnup andcooling time based on gross neutron and grossgamma detection.Neutron emission rate depends on initialenrichment, irradiation history and amount ofmaterial.Potential for partial defect verification bycombining fission product gamma rate(HRGS) with neutron rate.Distinguishes between irradiated MOX andLEU fuel.

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No. IVIeas. Meas. principle DescriptionPoint # Instn/method name (Techniques, handling, attribute verified, application)


3Application: Verification of declared burnup andcooling time based on gamma and neutron emissionlevels versus expected values for quantity and initialenrichment.

Fuel lifting or movement needed, any verticalposition verifiable.Determine irradiation history (number ofcycles); already been applied as well asunattended mode with camera surveillance.

Visual

CVD CerenkovViewing Device

Detection of Cerenkov glow using a CVD devicewith increased sensitivity to characteristicwavelengths (UV).Positioned directly above the assembly, visualdetection of Cerenkov glow used to verify thepresence of irradiated fuel.Visual pattern of the top structure of the assemblyused to verify the type of the object (assembly,dummy, etc.). No recording of the observed pattern.Attribute verified: Gamma-ray emissions of theassembly. Light emission caused mainly bysecondary energetic electrons originating fromCompton scattering of gamma-rays in water.Application: Verification that fuel assembly has beenirradiated.

Easy, fast and non-intrusive to use.Cannot distinguish irradiated fuel fromirradiated metal structures. Detected glowcaused by any highly radioactive materialunder water.Requires transparent, non-rippling pool water.Long cooling time (> 10 a), low burnup (< 10MWd/kgU) and storage pattern limit usability.Verifies only the top part of the assembly.

O

o

>

Passive neutron,passive gamma

BUD Burnup Device/CONSULHAContainment andSurveillance for HighActivities

Passive gamma and neutron detectors provide nuclearsignatures to permit verification of burnup of spentfuel assemblies.Gamma and neutron detectors coupled to unattendedvideo surveillance of events.Two units together form the verification device BUD/CONSULHAAttribute verified: Neutron and gamma emissions.Application: Unattended monitoring of spent fuelsignatures to verify nuclear material movements andinventories.

BUD measurement provides an estimate of theburnup.Neutron sensitivity depends on water thick-ness between assembly and fission detector.Typically lower detection limit ~ 8 GWd/tUfor neutronsBUD is more process oriented device whileCONSULHA is more surveillance oriented.

No. Meas. Meas. principlePoint # Instn/method name

"description(Techniques, handling, attribute verified, application)


Passive gamma,passive neutron,active neutron

FPFM Feed PondFuel Monitor

In passive mode, a 15 % HPGe detector and 5 fissionchambers are used in a monitor station for gamma-rayand neutron detection.In active mode, a 252Cf source is transferred next to theassembly and neutrons are measured from fissionsinduced by the source.Each assembly is measured up to four measurementheights during rotation in a monitor station. Localvalues of cooling time, irradiation, initial and finalenrichments are calculated in addition to an average foreach assembly.Attribute verified: Burnup, cooling time, initial andresidual enrichment 235U.Application: Unattended monitoring of spent fuelsignatures to verify nuclear material movements andinventories

Cooling time verification is based on theactivity ratios 134Cs/154Eu and IO6Ru/137Cs.Three separate values of irradiation aredetermined using absolute I37Cs rate, iso-topicratios of including 10*Ru, l37Cs and 134Cs ratesas well as passive neutron rate originatingmainly from 244Cm.In operation at THORP, Sellafield.Used for verification of burnup and coolingtime without using operator declared dataother than fuel type.

X! oz

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o50

Passive/activeneutron/gamma

PYTHON SpentFuel Detector,Cadarache

In passive mode, two detector heads under water withcollimated gamma and neutron detectors to measureemitted neutron and gamma radiation from the assemblybrought inside the device.Device is positioned on the top of the storage rack,assemblies lifted inside the device using fuel handlingmachine.In active mode, an external 2S2Cf source can be movedadjacent to an assembly using a cable and neutronscounted from fissions induced in the residual fissilematerial.Applicable both for criticality safety and safeguardsapplications. Measurement time 600 sec per assembly inpassive mode.

Gross defect verification of spent fuel.Burnup verification, burnup evaluationcode used for interpretation.Consistency measurement of burnup andcooling time based on gross neutron andgross gamma detection.Neutron emission rate depends on initialenrichment, irradiation history and amountof material.In active mode potential for partial defectverification by measuring amount ofresidual fissile content. Calibration needed.Fuel lifting or movement needed, anyvertical position verifiable.

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No. Meas. Meas. principlePoint # Instn/method name


Possibilities and limitations(level of verification) 1

Attribute verified: Gross neutron emissions, mainlyfrom even Cm isotopes, gross gamma detection of fissionand activation products (passive mode). Neutronsoriginating from induced fissions in the residual fissilematerial (active mode).Application: Verification of declared burnup and coolingtime based on gamma and neutron emission levels versusexpected values for quantity and initial enrichment(passive mode). Verification of residual fissile contents(active mode).

Commercially available.

Passive gamma

GBUV GammaBurnup Verification

Detection of fission product gamma-rays (mainly 137Cs)using standard high resolution gamma spectrometry(HRGS) with HPGe detector and facility specific fuelhandling.Assembly moved and/or rotated in front of a horizontalgamma collimator.Point-wise measurement allows any vertical position tobe measured. Scanning mode allows verification ofmost of the fuel volume.Calibration source (l37Cs) or reference assembly usedfor normalization.Attribute verified: Fission product gamma-rays.Application: Verification of declared burnup based onlinear relationship between 137Cs activity and burnup.

Quantitative gross defect verification ofburnup.Limited potential to reveal missing of fuelrods. Potential for partial defect verification ifcombined with passive neutron counting usinge.g. Fork detector.Fuel movement is needed, verification of anyvertical position possible.Verification based on 137Cs has no dependenceon initial enrichment or irradiation history.No limitations of burnup or cooling time.Cooling time verification possible based onfission product isotopic ratios.

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8 2,4 Passive gamma

TOMOGRAPHYPassive High EnergyGamma EmissionTomography

Reconstruction of 2-D activity cross section map ofassembly from measured activity profiles. Measurementtime for a BWR assembly to reveal missing of a singlerod around 1 h using an array of 10 CdTe detectors.Activity profiles of emitted gamma-rays of l44Pr (2186keV and 1489 keV) and 154Eu (1275 keV) are measured

Real partial defect verification on rod level.No operator declared information needed ofburnup, cooling time or irradiation history.Possible to detect missing of individual LWRfuel rods, not a direct measurement of Pucontent.

ao23

ON No. Meas. Meas. principle DescriptionPoint # Xnstn/method name (Techniques, handling, attribute verified, application)


using a well collimated array of detectors (CdTe,CdZnTe). Systematic scanning across the detectorarray at specified angles.Mathematical program used to calculate the crosssection activity map from measured profiles.Attribute verified: Fission product gamma-rays withlocations of emitting fuel rods.Application: Verification of integrity of spent fuelcontent in spent fuel assembly.

Fuel movement and rotation needed.

Passive neutron

Passive NeutronMultiplicity Counting

Use of passive neutron multiplicity counting todetermine content directly and to measure the 240Pu/244Cm ratio for the indirect verification of plutonium.Neutron multiplicity of singles, doubles and triplesneutrons for measuring 240Pu, 244Cm.Attribute verified: Coincidence neutrons from 240Pu.Application: Verification of plutonium content inspent fuel.

Impractical for LWR spent fuel becauseneutron rate from Cm » rate from Pu; poorPu/Cm accuracy.

10 4 Active neutron

Cf-252-Source-DrivenNoise measurements

A method measuring sub-critical neutron multiplicationfactor k of LWR fuel arrays. 252Cf source providesneutrons to initiate the fission chain multiplicationprocess in the spent fuel array.Fission counters measure the frequency-dependentcross-power spectral densities (CPSDs) between a pairof detectors located in or near the fissile material aswell as measurements of CPSDs between these samedetectors and a source of neutrons emanating from the252Cf source ionization chamber positioned in or nearthe fissile material.

Large neutron source is required (5 x 109 n/sec) to overcome background from 244Cm.Total fissile content is measured.Cannot detect natural uranium substitution inirradiated fuel.Very long measurement time needed (order ofdays is possible).

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No. Meas. IVIeas. principlePoint # Instn/method name


Possibilities and limitations(level of verification) H

OAttribute verified: Quantity of fissile material versusreference condition.Application: Verification of fissile content ofassemblies and casks.

11 2 Active neutron,passive neutron

Spent fuelIdentification System

Measurement performed in one axial position inapproximately 15 min/assembly. Neutron source (252Cfand neutron detector are positioned on opposite sides offuel assembly. Passive neutron emissions determined byfour neutron detectors positioned on four sides ofassembly.Attribute verified: Passive and induced neutronemissions from higher actinides and fissile materials.Application: Verification of burnup, total fissile, initialenrichment and type of fuel thus permitting calculationof amount of plutonium contained in uranium spent fuelassemblies.

of fuel and initial enrichment can bedetermined in cases where 244Cm is the mainneutron emitter (cooling time > 2 a, burnup>15 GWd/tU) by correlating the measuredneutron emission and multiplication.Active and passive neutron methods are used.Both uranium and MOX fuel assemblies canbe measured.

O

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12 4, 4b Active neutron

Active l24Sb-BeEpithermal NeutronMethod

124Sb-Be neutron source activation of spent fuelassembly. Penetration of assembly with epithermalneutrons from the source and the subsequent detectionof the prompt and delayed neutrons produced by theresulting fissions. Four 124Sb-Be sources penetrate theassembly and provide a nearly uniform interrogation ofthe fuel. Assembly's high neutron and gamma back-grounds are overridden by intense interrogationsources, thus ensuring a fission neutron production ratehigher than the background neutron emission rate.Attribute verified: Total fissile content.Application: Verification of fissile actinide content ofspent fuel assemblies.

X

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3

00 No. Meas. Meas. principle DescriptionPoint # Instn/method name (Techniques, handling, attribute verified, application)


13 2 Active neutron

Californium Shuffler

Delayed neutrons, counted immediately after theneutron source is removed, come from neutron richfission products resulting from fissions induced by theneutron source presence. Number delayed neutrons isproportional to number of induced fissions which isproportional to the fissile material present.Attribute verified: Total fissionable content.Application: Verification of uranium and plutoniumcontent of spent fuel assemblies.

Large background rate from Cm in LWR, soyou need a large 252Cf source (> 3 mg).Practical (and already used in Idaho ChemicalProcessing Plant, US) on HEU fuel.

14 4 Active neutron

Lead Slowing DownTime Spectrometer

Lead spectrometer consists of lead pile driven byneutron pulse from accelerator. As neutron pulse slowsdown in lead, neutrons are focused in energy about amean energy which is inversely proportional to theslowing down time. Amount of fissile material of thefuel assembly inside the lead pile is obtained bymeasuring prompt fissile neutrons emitted from fuelwhich reflects the effective spectrum averaged crosssection of fissile isotopes present.Attribute verified: Quantity of fissile isotopes.Application: Determination of quantity of uranium andplutonium (and other actinides) in spent fuelassemblies.

Direct assay of total mass of fissile material,spatial fissile material distribution usingtomography under development.Only method that can measure Pu and Useparately by detection of induced fast fissionneutrons as a function of time (betweeninterrogating pulses).Large, massive (heavy lead construction isneeded) not practical for wider use.Intense pulse neutron source (Linac) isneeded.

15 4 Active neutron

Synchronous ActiveNeutron Detection

14 MeV neutron generator and a novel detection system(digital lock-in amplifier) applied to direct measurementof spent fuel by detecting small signals in presence oflarge background. Interrogating neutrons are non-thermaland penetrating.Attribute verified: Fission neutrons from fissile content.Application: Verification of fissile material content inspent fuel.

Laboratory testing only so far(LANL, LA-UR-94-3135).

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No. Meas.Meas. principle DescriptionPoint # Instn/method name (Techniques, handling, attribute verified, application)


16 2 Passive gamma

CANDU Fuel Monitor

Multiple detectors fixed on reactor face, triggered bymovement of spent fuel out of reactor.Identify reactor's region where fuel is expelled.Secure network for information flow and analysis.Attribute verified: Gross gamma emissions from spentfuel fission products.Application: Unattended monitoring of spent fuelsignatures to verify nuclear material movements andinventories.

Installed in Canada.Only shows movement and position ofexpelled fuel; no burnup analysis.

17 4 Passive neutron,passive gamma

Underwater FuelMovement

lonization chambers used as gamma radiationdetectors for monitoring fuel movement and directionof movement through channels.Secure network for information flow and analysis.Attribute verified: Gross gamma emissions fromspent fuel fission products used to trigger cameras tophotograph IDs on fuel or "skip" with fuel.Application: Unattended monitoring of spent fuelsignatures to verify nuclear material movements andinventories.

Installed in Joyo, Japan.Installed in THORP (Transfer ChannelMonitor).Used by Euratom and IAEA at Vandellos,Spain (SPEFAC, 2 gamma detectors)No burnup analysis.

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18 4 Passive neutron

NDA of ResearchReactor Fuel #1 & #2(LANL)

"Can", GRAND-I and JSR-11 (#i) for low-powerresearch reactor fuel.3He neutron detectors and ionization chambers for rods •measured in groups of 4-5 at a time inside a canisterunder water.Neutron rates correlated to exposure, ion chambers recordexposure profile."Fork", GRAND-I and HRGS (#2) for low-powerresearch reactor fuel.

For low burnups, measure Pu directly withpassive count.Used at Savannah River, US.


ANNEX 2 LIST OF NDA INSTRUMENTS

'a.

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40


RELEVANT R&D PROJECTS ANNEX 3

The need for new safeguards verification techniques is based on the needs of the final deposition.In planning the safeguards measures, the existing NDA techniques form the basis. Because of theirreversible nature of the final deposition, higher level verification of the correctness andcompleteness of the operator declared spent fuel data is needed. Even though some special NDAmethods are known or have been shown feasible, additional research and development efforts areneeded to further improve the techniques and optimize the instrumentation especially for partialdefect verification of spent fuel prior to final deposition.

A3.1 Fissile material measurement through the cask wall

Technique used:Californium Source Driven Noise measurement

Features:• active neutron interrogation technique (Cf-252 source),• measured parameters: cross power spectral density (CPSD) and auto power spectral density

functions,• sensitive only to correlated fission events,• insensitive to any other neutron/gamma activities,• not sensitive to inner parts of the cask,• measured quantity: total fissile content,• good reference is needed.

Status:• laboratory experiments,• limited feasibility study for a dry cask with identical assemblies, for details see ref. [8, 9, 10].

1" thick steel cask

poly reflector

He detector

poly moderator

Figure A3.1 a) Arrangement (Source detector cask configuration).

41


ANNEX 3 RELEVANT R&D PROJECTS

0,12

_ 0,11CM

o

0,10

0,09

0,08

0,07

fresh

32 GWdMTU

0,10 0,15 0,20 0,25 0,30

Fissile material content, gcc

0,35

Figure A3.1 b) Results: Magnitude ofCPSD G as a Junction of fissile mass.

A3.2 Pu measurement of spent fuel

Technique used:Lead Slowing Down Time (LSDT) spectrometer

Features:• active neutron interrogation technique,• separate measurement of Pu-239 and U-235,• high intensity Linac source is needed,• can be used only for dry conditions,• heavy lead construction is needed.

Status:• experimental result for small samples,• feasibility for spent fuel assembly,• simulation for mapping Pu content by tomographic arrangement,• for details see ref. [11, 12, 13].

42



ICM

Transfer cask with a drying mechanism/(16ton lead shield)

Spent fuel assembly

Optimized LSDTS(36 ton lead slowing down medium)

Electron linear acceleratori (linac)

Target at 20cm from the center(Ta, D or Be using(e r) (r n) reaction)

60 surrounding threshold detectors(U238 and TH232) in each layer(each detector 2,5 cm dia and30 cm long)

15 ton lead shield

25cm ^14,6cm

160 cm

Figure A3.2 Arrangement, geometric configuration of optimally designed LSDTS with a shield.

43



A3.3 Partial defect testing of multi-assembly configurations

Technique used:High Energy Gamma Emission tomography

Features:• passive gamma technique,• high energy fission products (Eu-154, Pr-144) are measured by an array of semiconductor

detectors,• can be used for both wet and dry conditions,• application limited to small size MEBs (max. 7 BWR or 5 PWR assemblies).

Status:• experimental proof for BWR assemblies (see fig.),• simulations for PWR and small size MEB configurations,• for details see ref. [4-7].

MEB wall

assemblies

f en::::::

Scanningdirection

\Collimator Detector

Figure A3.3 Arrangement, diametrical scan of a MEBfor cross sectional imaging.

AA



ISSCaoo"

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Figwre A3.4A measured tomograph (activity map) of an 8x8 BWR assembly showing the position of thewater filled (not fuel containing) inner fuel rod.

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"llilitiFigure A3.5 A simulated tomograph revealingmissing of one inner BWR assembly in a MEBfor 7BWR assemblies.

45



A3.4 Verification of casks by passive neutron detection

Techniques used:a) Measuring passive total neutron signal outside the canister when the contents are known

(baseline measurement) and later whenever verification is necessary. Cm decay correction canbe applied.

b) Calculational based passive approach, when no baseline measurement is possible.

Features:• baseline measurements/or calculations are necessary for all arrangements and conditions (air,

underwater, shielded, etc.),• technique b) has no potential to see missing assemblies in large casks and it cannot be used for

large self-shielded casks.

Status/reference:• See tasks of the US Safeguards Support programme to the IAEA.

46


SUMMARY OF WORK PLAN FOR SAGOR (TASK C 799)

Revised: 15 February 15 1997

ANNEX 4

ActivityNo.

0

1

2

3

4

5

6

7

8

9(formerlyactivities)

9-13

10 (for-merly 14)

Activity Description

Introduction

Describe model facility

Identify diversion paths and detection points

Identify events and conditions requiring DIVand examination procedures

Evaluate IAEA use of operator safeguards,safety, and process system outputs

Identify potentially applicable geophysicaltechniques

Evaluate NDA techniques for spent fuelverification and radiation monitoring

Evaluate C/S techniques and integratedverification systems for spent fuel monitoring

Determine guidelines for acceptablesafeguards approaches

• Design safeguards approach and evaluatecandidate approaches• Develop redundancy and reliabilityrequirements for verification systems• Select safeguards approach and identify R&Dneeds• Develop QA program for detection systems• Specify system design requirements formodel safeguards approach

Integrated final report

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BEL = Belgium; CAN = Canada; FIN = Finland; FRA = France; HUN = Hungary; SWE = Sweden;UK = United Kingdom; US = United States; * = Lead responsibility.

47

Documents

NDA techniques for spent fuel verification and radiation monitoring · 2004. 8. 3. · 6 CONCLUSIONS OF EXISTING NDA TECHNIQUES 17 6.1 Gross defect verification of assemblies 17 6.2