NDA Techniques for Spent Fuel Verification and Radiation Monitoring

STUK-YTO-TR 133 FI9700145 AUGUST 1997

NDA techniques for spent fuel verification and radiation monitoring Report on Activities 6a and 6b of Task JNT C799 (SAGOR) Finnish Support Programme to the IAEA Safeguards

Matti Tarvainen Radiation and Nuclear Safety Authority, Helsinki, Finland Ferenc Levai Technical University, Budapest, Hungary Timothy £. Valentine Oak Ridge National Laboratory, Oak Ridge, TN, USA Mark Abhold Los Alamos National Laboratory, Los Alamos, NM, USA Bruce Moran USNRC, Washington D.C., USA

The conclusions presented in the report are those of the authors and do not represent the official position of the Radiation and Nuclear Safety Authority.

RADIATION AND NUCLEAR SAFETY AUTHORITY (STUK) P.O.BOX 14, FIN-00881 HELSINKI, FINLAND Tel. +358-9-759881 Fax +358-9-75988382 V 9-0 1 ISBN 951-712-222-5 ISSN 0785-9325

Oy Edita Ab Helsinki 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 the IAEA Safeguards. STUK-YTO-TR 133. Helsinki 1997. 23 pp. + Annexes 24 pp.

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

Keywords: safeguards, spent fuel, NDA

ABSTRACT

A variety of NDA methods exist for measurement of spent fuel at various stages of the disposition process. Each of the methods has weaknesses and strengths that make them applicable to one or more stages in disposition. Both passive and active methods are, under favorable conditions, capable of providing either a mapping of an assembly to identify missing fuel pins or a measurement of the fissile content and some are capable of providing a mapping of a canister to identify missing assemblies or a measurement of the fissile content. However, a spent fuel measurement system capable 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 providing quantitative measurements on fuel assemblies and canisters. Application of NDA methods to shielded casks may not be practical or even possible due to the extent of radiation attenuation by the shielding materials, and none of these methods are considered to have potential for quantitative measurements 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 spent fuel prior to loading in a canister or cask at the conditioning facility. Fissile material tracking systems in addition to containment and surveillance methods have the capability to assure continuity of the verified knowledge of the sample from loading of the canisters to final disposal and closing of 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 8 3.2 Spent fuel flow chart 8 3.3 Measurement point categorization 10 3.4 Conditioning of spent fuel 10

4 PROPERTIES OF SPENT FUEL AND CONTAINERS 11 4.1 Spent fuel parameters 11 4.2 Disposal containers 12 4.3 Mixed emplacement 12

5 APPLICABILITY OF NDA-METHODS TO SPENT FUEL VERIFICATION 14 5.1 Measurement principles 14 5.2 Instruments for spent fuel verification 15 5.3 Fissile material tracking 16

6 CONCLUSIONS OF EXISTING NDA TECHNIQUES 17 6.1 Gross defect verification of assemblies 17 6.2 Partial defect verification of assemblies 17 6.3 Radiation monitoring of casks 18 6.4 Unsolved NDA problems 19

REFERENCES 20

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

ANNEX 2 List of NDA instruments 32

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

ANNEX 4 Summary of work plan for SAGOR (Task C 799) 47 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

1 INTRODUCTION

A joint task of the Member State Support main concerns include the amount' of nuclear Programmes called "Programme for Develop- material deposited and its isotopic composition. ment of Safeguards for Final Disposal of Spent The ultimate objectives of safeguards, i.e. "the Fuel in Geological Repositories (SAGOR)" has timely detection of diversion of significant been carried out under Task JNT C799. For quantities of nuclear material from peaceful practical reasons, SAGOR has been divided nuclear activities to the manufacture of nuclear into 10 separate Activities (see Annex 4), the weapons or of other nuclear explosive devices last being compiling of an integrated final or for purposes unknown, and deterrence of report Activity 6 "NDA techniques for spent such diversion by the risk of early detection", fuel verification and radiation monitoring" has can be achieved using conventional safeguards been carried out in co-operation with the measures while the material is in the con- Support Programmes of Finland, Hungary and ditioning facility or even in an open repository. USA. How this goal can be reached in conditioning facilities and open repositories using non- Developing safeguards for final disposal issues destructive assay (NDA) is the main topic of a challenge to the safeguards community and this report. In the case of a closed repository, safeguards experts. Conventional safeguards is the NDA techniques measuring radioactivity of directed to nuclear material that can be handled, spent fuel are no more applicable and this phase measured and monitored. All of the safeguards of the final deposition is not handled in this measures developed, so far, are based on this report. assumption. Final disposal of spent fuel in geological repositories is a challenge because The report covers a wide range of measurement none of the conventional features are valid once activities. Section 5 deals with the applicability the container has been placed into the repository of different measurement principles for NDA of and the access backfilled. Nuclear material can spent fuel. Conclusions of the present NDA no more be handled, measured or even moni- techniques are drawn in section 6. Annex 1 tored. The safeguards measures should focus on includes a summary of the principles and indirect objects like surface of the closed methods and Annex 2 lists existing NDA repository [1]. instruments that are available or under development. These methods can be used in planning Final disposal in geologic repositories is meant the safeguards verification activities of the final to be an irreversible process. In this light, deposition. Annex 3 lists R&D projects which verification measurements of the declared para- are considered to have the highest relevance for meters of the nuclear material have to be the future final deposition safeguards measure- performed in a way to give all answers to the ments. questions made now and also in the future. The RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

2 ROLE OF VERIFICATION MEASUREMENTS

Independent verification of spent nuclear mate- lation techniques of the fission product gamma- rial by the IAEA makes use of NDA measure- rays. Burnup (exposure), i.e. energy released ments. The material to be deposited has been per unit mass, can be directly verified only under IAEA safeguards during the front-end of during power production. Fission product the fuel cycle prior to entering the conditioning gamma-rays can, however, be used for indirect facility and the repository. Also verification verification of the burnup and the contents of measurements have been performed as a part of transuranic isotopes. In addition to the gamma- the routine safeguards activities. Prior to load- ray methods, passive and active neutron assay ing into the storage containers to be deposited, can be used for verification of fissile content of spent fuel assemblies may loose their integrity spent fuel. under consolidation and conditioning. Verifi- cation measurements need to be optimized to One of the main parameters to be verified prior give full and continuing assurance that the to final deposition is the completeness of the operator declared data are complete and correct. declared data. The optimum amount of infor- The logical order of safeguards measures is mation a verification measurement can produce schematically shown in Figure 1. gives assurance that the material declaration relates to the material measured, that the The triangle of Figure 1 applies also to the isotopic composition and mass data are correct safeguards of the conditioning facility. In de- and that no material is missing or has been signing the safeguards measures, verification of replaced by dummies. Because spent fuel declared data is logically followed by con- verified at the conditioning facility has been tainment and surveillance (C/S) to maintain cooled and stored for decades, no details of continuity of the verified knowledge. In an open earlier verification measurements may be avail- repository, use of C/S methods for monitoring able anymore or can be verified by re-measure- of cask movements is foreseen. ments. In this light, one of the main goals of the verification measurements at the conditioning Spent fuel arriving at the conditioning facility facility should be to give assurance that spent has operator declared mass, isotopic com- fuel assemblies and rods are really irradiated position, burnup and cooling time data The nuclear material. If absence of diversion can be parameters that can be directly verified by NDA confirmed, details of the composition of the methods include mass, isotopic composition material to be deposited is of minor interest and cooling time. Cooling time verification from the proliferation point of view. measurements are based on the isotopic corre- STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

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. RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

3 POTENTIAL SPENT FUEL MEASUREMENT POINTS

3.1 IAEA spent fuel verification 3.2 Spent fuel flow chart

criteria According to the IAEA glossary [3], the key IAEA has incorporated in the Safeguards Cri- measurement point (KMP) is defined as "a teria the activities that are necessary to carry location where nuclear material appears in such out the IAEA safeguards. In addition to regu- a form that it may be measured to determine lations concerning inspection frequencies at material flow or inventory". The KMP concept different types of facilities for verification of applies to the final deposition as long as different types, forms and amounts of material, material is available for measurements, i.e. at the Criteria define verification activities to be the conditioning facility. Figure 2 shows sche- performed. Different levels of verification are matically the flow of spent fuel assemblies also specified. The Criteria have been updated from the point of view of verification measure- annually during 1992-1996 [2]. Each of the ten ments using NDA techniques. facility types sections has also verification measurement tables. These tables list the re- The flow chart begins from a reactor pool (top commended instruments to be used to perform left) which receives spent fuel from the reactor. the required verification measurements. The After several loading, transport and reloading Safeguards Criteria obviously do not include operations as well as possible reconstruction, criteria for final deposition or conditioning consolidation and conditioning operations the facilities. From the point of view of safeguards flow ends in a closed repository (bottom right). verification measurements of spent fuel, e.g. those parts of sections 1 "LWRs" and 9 After possible reconstruction operations (top "Storage Facilities" related to irradiated direct- middle), spent fuel assemblies are moved from use material are, however, relevant. the reactor pool to an interim storage. When separate assemblies are stored in a storage rack, In conditioning facilities, spent fuel NDA NDA verification measurements can be per- verification measurements can be performed on formed rather easily. This is indicated in Figure the gross defect level or on the partial defect 2 by placing the squares in the left-hand level. Gross defect means that the whole column. If assemblies or cans of rods are stored assembly (all nuclear material) is replaced with in closed containers, verification measurements a dummy or the material is missing. Partial are more complicated, if possible at all. This is defect means that one half or more of the indicated by placing the squares in the right- irradiated rods of an assembly are replaced or hand column, respectively. The column in the missing. The two categories include all methods middle includes those parts of the spent fuel available. flow where fuel is moved or handled. For NDA, this may offer a potential measurement possi- The target of diversion is at least 1 SQ of bility because the fuel is moved, anyway. nuclear material. Within the reference repository a typical container will have about 14 kg of The upper half of Figure 2 shows, for complete- Pu plus about 23 kg of 235U. Thus, in order to ness, the storage history of spent fuel before it divert at least 1 SQ, only one container of fuel is transported into the conditioning facility. The would have to be diverted and opened.

8 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

need for NDA verification depends on the the SAGOR limit in Figure 2 includes that part information available from the fuel and the of the fuel cycle which is covered by the continuity-of-knowledge. The lower part under SAGOR task.

Access to separate Cask handling or Cans / assemblies assemblies fuel operation in closed containers

reactor pool > reconstruction reactor pool 2b ») loading (<-- T 2 2 I transport j— dry storage interim storage pool k- (type 2) unloadinT g dry storage vault (type 1) loading

SAGOR LIMIT transport conditioning facility reception pool conditioning facility \ unloading buffer storage conditioning facility reception vault consolidation conditioning facility conditioning 4, 4b J repository 4, |oading > 6b receipt storage ' | transport / transfer ground surface

• >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 the final deposition. Methods under the SAGOR limit are discussed in this report. RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

3.3 Measurement point fuel will be in closed welded and/or sealed disposal containers on which adequate veri- categorization fication has been performed at the time of Figure 2 includes tags with numbers l-6b. loading and the continuity-of-knowledge has Numbers indicate possible measurement points. been maintained since the time of the container Storage measurement points are shown in the loading. Measurement methods applicable for left-hand and in the right-hand column, flow verification of closed containers are few in measurement points in the mid-column, respec- number. However, they are also dealt with in tively. Another division is made into measure- this report. ment points under water or in the air.

The number of potential measurement methods applicable for verification under the SAGOR 3.4 Conditioning of spent fuel limit in Figure 2 is large. Similar instruments and methods may have facility specific features At the conditioning facility spent fuel will be which still increases the total number of appli- transferred into a hot cell for further processing. cations. This is why similar measurement The following types of fuel will be received: conditions at different points of the flow chart • intact fuel assemblies have been tagged with the same number. For • defective fuel assemblies example, the measurement point number "1" • containers of damaged/defective fuel rods. (Figure 2) is attached to all those places where separate spent fuel assemblies are stored in In the hot cell the fuel will be processed in one racks under water. In these points measurements of the following manners: can, in principle, be performed from above • Encapsulation of intact fuel assemblies into without moving the fuel. If the fuel assembly is the disposal cask. This involves no dis- moved or lifted e.g. using a fuel handling assembling. machine or crane, measurements can be per- • Encapsulation of disassembled fuel compo- formed without interference from the sur- nents into canisters and subsequent en- rounding assemblies and also from different capsulation of canisters into the disposal vertical positions. Different equipment and cask. Depending on the operational require- measurement methods are obviously needed for ments, fuel and non-fuel bearing measurements from above and from the side. components will be separated. Categorization of the measurement points is • Disassembled fuel components will be meant to simplify the handling of the veri- divided into smaller pieces before placing fication question and the search for suitable into the disposal cask. methods for different conditions. • Encapsulation of waste generated by the process into the disposal cask together with At different stages of the flow (Figure 2) spent other material with same specifications. fuel will be transported and stored in closed containers. If a closed container is supposed to During conditioning, transition of items from be re-opened, verification of separate assem- one type (e.g. fuel assemblies or rods) into blies or cans may be possible. Such verification another type (i.e. final disposal container) will activities are related to the conditioning facility. occur. The contents of the new item may be verified in respect to the nuclear material In the SAGOR task it has been assumed that contents. when entering a geological repository, all spent

10 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

4 PROPERTffiS OF SPENT FUEL AND CONTAINERS

4.1 Spent fuel parameters well as physical dimensions of fuel assemblies that emit and also absorb the radiation used for Typical spent fuel parameters for 5 different verification measurements. fuel types are shown in Table I. The Table is compiled by D. M. Wuschke for Activity 1 of Existing safeguards measurement techniques the SAGOR project The Table shows those have been challenged by recent technological parameters of fuel assemblies that are relevant development for better management of the for safeguards, i.e. nuclear material contents as reactor fuel. New fuel types are designed for

Table I. Characteristics of spent fuel intended for direct disposal. Characteristics PWR BWR I HWR- THTR/ RBMK I CANDU HTR-500 Reactor size (MWe net) 1000 1000 100-900 300/500 1000/ 1500 Approximate fuel assembly dimensions (cm) Length 320-483 447 49,5 N/A 1006 Cross section Side (square) 19-23 14-15,3 Diameter (cylinder or sphere) 8,1-10,3 6 7,9 (sphere) Mass per assembly (kg) Total 480-840 250-307 16,6-24,7 N/A 185 Heavy metal 322-548 172-194 13,4-19,8 11 130 Rods per assembly 126-331 47-64 ! 19-37 _J^A__j 18 Power density (MW/m3) 85 50 ! 10 32/36 | ? Design burnup (GWd/Mg) 26-50 27,5-40 \ 6,5-8,1 | 80-160 | 25 Total activity (Ci/kg) 320 290 | 84 ? ? after 10 a | ] Decay heat (W/kg) 2,3 2,2 | 0,22 I ? ? after 10 a Calculated fuel discharge rate 32-38 38-40 150 ? 60 MgU/GWe.yr Fuel enrichment initial %235 U 3,0-4,4 2,5-3,5 | 0,71 93/9,1 2,0-22 after irradiation %235 U 0,8-1,26 0,8-1,0 | 0,205-0,282 80/8,1 1 Plutonium content after irradiation kg/assembly 3-5 1,2-2,0 0,06-0,08 I N/A ? kg/fuel rod 0,014 0,022 0,002-0,004 I N/A ? No. of items per SQ of Pu Fuel assembly 2-3 4-7 100-120 N/A ? Fuel rod 560 360 4000 N/A ? D.M. Wuschke: A reference repository for the development of safeguards for disposal of spent fuel

11 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133 easy disassembling. During rod exchange, fuel Characteristics of disposal containers for five assemblies lose their original constituents and countries are summarized in Table II. The basic identity. In addition to accountancy, this may unit handled in the repository will be the cause problems to C/S and verification meas- disposal container. When not within a shielded urements. Advanced fuel designs incorporating facility, the disposal container is enclosed in a different initial enrichments, axial variations in re-usable disposal container cask. Spent fuel is enrichment, use of MOX and burnable poison normally not present in the repository except rods introduce additional difficulties for NDA within a disposal container, and the repository measurements. Use of special storage baskets design and operation are independent of most or multielement bottles (MEB) or multipurpose spent fuel characteristics. However, the size and canisters (MPC) to accommodate an increasing mass of the fuel affect the design of the number of spent fuel assemblies have added containers and container casks. Individual fuel further complications also to verification meas- assemblies or rods would be present only if a urements. container were damaged or deliberately opened.

4.2 Disposal containers 4.3 Mixed emplacement

A geological repository for final disposal may In those countries where both direct disposal be designed for emplacement of disposal con- and reprocessing of spent fuel is considered, tainers within the rooms or in bore holes drilled simultaneous disposal of both high level waste in the walls or floors of the rooms. Some states (HLW) and spent fuel might occur in the same intend to dispose spent fuel, high level waste geological repository. For safeguards purposes (HLW) and possibly low and intermediate level it will be required to distinguish canisters waste (LILW) in the same repository. The containing HLW from canisters containing disposal containers for the spent fuel and the spent fuel. This will be important for the two types of waste would be designed to be material flow into the repository as well as for different in appearance. the possible material flow out of the repository.

Table II. Characteristics of currently considered disposal containers. Canada Germany Sweden United Finland States Type of fuel HWR LWR HTR LWR LWR BWR pebbles PWR/BWR Disposal Container • Material Titanium Steel Steel Copper/ Steel Copper/ Steel Steel • Height (m) 2,25 5,35 5,50 4,50 5,3 4,5 • Diameter (m) 0,63 1,57 1,45 0,8 1,65/1,6 0,8 • Wall thickness (mm) 6,35 350 350 100 120 100 • Mass with Fuel (Mg) 2,8 64 55,0 18,5-22 50/46 14-19 • Self-Shielded no yes yes no no no Mass of Fuel/Container 1,65 3,0 1,7 1,5 8 1,5

Assemblies/Container 72 8 PWR or 8400 8-12 BWR 44 BWR 9 BWR (max). 24 BWR or 4 PWR 21 PWR Surface Dose Rate 5x10" <0,2 ? ? 120-290/ 100 jmSv/h) L_ 80-270 Pu content (kg) 5 30 ? 14 100-150/ 14 (approx.) 80/144 D.M. Wuschke: A reference repository for the development of safeguards for disposal of spent fuel (Report of SAGOR Activity 1B)

12 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

Table HI. Characteristics of disposal geologies and repository design. Belgium Canada Germany Sweden Finland United States Package Type unshielded unshielded self unshielded unshielded unshielded shielded Mixed Emplace- unknown no probably no no yes ment with HLW Capacity (MgU) 3,500 191,000 10,000 8,000 1,840 87,000 Disposal Rate • Mg U/a 120 4,700 200 300 140 3,000 • Assemblies/d 1,080 ? ? • Containers/d ? 15 ? <1 <1 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 fuel for the final disposal concept of 6 countries.

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5 APPLICABILITY OF NDA-METHODS TO SPENT FUEL VERIFICATION

5.1 Measurement principles applied to identify fissile materials. These methods rely on measuring spectral signatures This section provides a summary of the prin- from fissile materials. ciples of nondestructive assay (NDA) techniques that can be considered for use in one or Passive neutron measurement methods can be more phases of spent fuel disposition. A more used to estimate burnup or to provide an detailed description on existing NDA methods indication of the fissile mass in spent fuel. The is included in Annex 1. two main passive neutron measurements are total neutron counting and coincidence coun- Passive NDA measurement methods rely on the ting. detection of radiation emitted from materials to characterize the material The measurement Active NDA methods rely on gamma-ray or system may measure the primary radiation neutron sources to measure the attenuation of emitted directly from the material or secondary radiation through spent fuel, to obtain a map- radiation emitted because of interactions of ping of the spent fuel in an assembly or primary radiation. Passive NDA methods are canister, and/or to obtain quantitative estimates generally categorized into two areas: gamma- of the fissile content in spent fuel. Some active ray and neutron measurements, but calorimetric NDA methods utilize the prompt radiation and Cerenkov measurements can also be con- resulting from fission while others utilize the sidered as passive methods. Passive meas- emission of delayed radiation from fission. A urements require reference measurements to wide variety of sources exist for active NDA obtain quantitative information about the spent such as 252Cf , AmLi, PuBe, DT generators, fuel burnup and a destructive analysis could electron accelerators, 57Co, etc. One benefit of then be performed on the reference assembly to active NDA methods is that the energy and obtain the absolute burnup. Calibration curves intensity of the interrogating radiation can be would have to account for differences in the changed to emphasize the desired properties of initial fuel enrichment, irradiation history, mod- spent fuel. erator conditions, and poison concentrations at different reactors. Both neutron and gamma-ray Active gamma-ray methods employ photon measurements also require axial profiles of the sources that interact with the fissile material fuel assemblies to correct for burnup variations. and subsequent radiation measurements are Passive NDA methods are simpler than active performed. Densitometry and X-ray fluores- methods and may be more appropriate de- cence are the two prominent active photon pending on the application. NDA methods. Photo-fission sources are another active photon NDA measurement alter- Passive gamma-ray methods can be categorized native. into two areas: gross gamma counting and gamma spectroscopy. There are many ways in Active neutron NDA methods utilize neutron which gamma-ray spectroscopy has been sources to induce fission in fissile materials and

14 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY require the measurement of the subsequent While the 239Pu content in LEU fuel is the prompt and delayed neutrons emitted from the component of most importance, the 235U content fissile material. The fissile content of a material becomes more important in HEU spent fuel. can readily be determined from many of these HEU spent fuel will comprise at most a small methods if calibration standards are available. part of the geologic repository total inventory, These methods are all related and vary de- but in some ways may be a more attractive pending on what radiation is measured and how target for diversion, being smaller and lighter the data is processed. The measurement method than LWR fuel and typically having lower may analyze prompt neutrons, prompt gamma- radiation levels. NDA methods for measuring rays, delayed neutrons, delayed gamma-rays, or HEU fuel will be addressed separately from the neutron transmission in a fuel assembly. LEU fuel. HEU spent fuel presents somewhat less difficulties for NDA measurements than Other NDA methods that can be considered for LEU spent fuel. spent fuel assay include calorimetry and Ceren- kov radiation. Measurement of spent fuel inside canisters containing multiple assemblies is also ad- The 239Pu content of LEU Light Water Reactor dressed. Canister measurements pose a number (LWR) spent fuel is of primary importance to of additional difficulties above and beyond the safeguards, with the residual M5U content being measurement of single assemblies. The burnup, much less significant. An ideal NDA method age, and cooling times may vary amongst the for LEU fuel would then directly measure the assemblies in a single canister, in fact, it may be 239Pu content, separating it from all of the other beneficial in meeting the canister total heat load constituents. Unfortunately, LEU spent fuel and criticality limits to purposely mix low and assemblies pose measurement difficulties un- high burnup fuel within a single canister. matched by any other nuclear material. Fission Proposed canisters vary widely in construction, products in the spent fuel emit gamma-rays at materials, criticality control methods, and capa- an extremely high rate, making all but a few city. The largest canisters may have 21 or more gamma emission lines invisible in the Compton PWR assemblies, 40 or more BWR assemblies. background. It is not possible to measure the Assemblies located in the interior of these large emission lines from Pu in spent fuel with canisters are effectively shielded by the exterior passive techniques. Similarly, the neutron emis- assemblies. sion rate of LEU spent fuel is very large, and is dominated by spontaneous fission from Curium 5.2 Instruments for spent fuel isotopes, making direct passive neutron measurements of the spontaneous fission of pluto- verification nium isotopes impossible. Passive methods are Section 5.1 above deals with methods and then limited to measuring attributes of the fuel principles that under certain conditions can be that can only be associated with the plutonium used for verification of declared data of spent content by reference to calculations. However, fuel assemblies or fuel rods. Those instruments given acceptable knowledge of the history of and methods that are already in use or that are the assembly, its initial enrichment, irradiation under development on prototype level are listed history, cooling time, and physical integrity, in Annex 2. Annex 2 is meant to include only these indirect measurements can provide accu- methods that have potential for verification of rate knowledge of the Pu content. Active spent fuel before or during conditioning ending measurement on LEU fuel are also very chal- in final deposition. Annex 2 includes also a lenging, with the radiation emission of the measurement point no. that relates to the flow assembly not only creating difficulty in chart of Figure 2. achieving good signal-to-noise ratios, but also making physical access to the fuel difficult.

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Verification of closed disposal containers is a tracking and monitoring system. The proposed much more demanding task than verification of system would be comprised of a set of gamma individual assemblies. Even though all possible sensitive detectors located on the ceiling of the measurement techniques are considered, one area to provide a two dimensional mapping of cannot find any method or technique that could the spent fuel movement. This system has not be used as such for practical verification of been applied to spent fuel monitoring but its declared data of a closed disposal container. low cost makes it an attractive method. There are, however, a few R&D projects that are related to verification of closed spent fuel Unattended monitoring systems containers including closed disposal containers. Such projects are described in Annex 3. Unattended continuous radiation monitoring systems are in IAEA use at reactors worldwide 5.3 Fissile material tracking [35]. These systems monitor fresh-fuel re- charge, the reactor core power level, spent fuel Perimeter radiation monitors are used to moni- discharge, and spent fuel storage. The radiation tor the transport of nuclear materials within a sensors are operated in the continuous mode controlled area. Perimeter monitors can be used with data collection at intervals of one to two to monitor the regular path of nuclear materials seconds. The data is collected at inspection within a controlled area or to detect un- periods of one to three months, with data authorized removal of nuclear materials from a filtering reducing the total amount of data controlled area This section highlights the stored to just those periods when nuclear currently available perimeter monitors. material is moving. In conjunction with containment/surveillance systems and burnup veri- Continuous site tracking system fication measurements, unattended monitoring systems have the potential to provide conti- The emission of gamma radiation from spent nuity-of-knowledge from fresh fuel loading fuel can be used as a signature to track and through spent fuel canister loading and storage monitor the movement of spent fuel within a at reactors and conditioning facilities. Efforts controlled area This system was originally are currently underway to allow remote moni- designed to serve as a criticality accident alarm toring of these systems to reduce inspector system [34] but could also serve as a fuel travel and workload.

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6 CONCLUSIONS OF EXISTING NDA TECHNIQUES

6.1 Gross defect verification of 6.2 Partial defect verification of assemblies assemblies Most of the NDA measurement methods avail- According to the IAEA Safeguards Criteria, able are capable of performing gross defect partial defect testing may be needed before the verification of spent fuel. Gross defect test is identity of fuel assemblies is lost (by cutting, also the requirement most often needed by the dissolving, etc.). It is foreseen that partial IAEA for verification of spent fuel assemblies defect verification needs to be performed for during different phases of the fuel cycle. each fuel assembly arriving at the conditioning facility in an active way, i.e. the measurement Annex 2 includes detailed information of 18 needs to prove that all rods in each assembly different measurement methods. All of them are present and that all rods include irradiated satisfy the gross defect requirement i.e. they nuclear fuel. A schematic potential measure- can be used to verify that the assembly is not ment system is shown in Figure 3 below. replaced with a dummy or missing. The variety of NDA methods for gross defect verification is In the input stage shown in Figure 3, each wide and there is no need for major R&D assembly still maintaining its integrity, is meas- efforts. ured using a partial defect method. In essence,

Input Process Output (rod dismantling)

Can

JulILulu

Waste Presence of irradiated rods Continuity-of-knowledge maintained by C/S verified by NDA

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

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the presence of all irradiated fuel rods is Californium source driven noise verified. If based on the measurements the measurement declared data are questioned, the assembly can be moved aside and clarification can be re- The active neutron interrogation technique (Cf- quested from the operator or shipper. If declared 252) can be used for measurement of the total data is confirmed to be correct and complete by fissile content with a high precision if appro- the measurement, the assembly can be pro- priate reference information is available. Only cessed further and its integrity may be lost. dry assemblies can, however, be measured. Continuity of the verified knowledge is maintained by C/S methods. Python

Those NDA methods existing now or under The active mode of the method allows verifi- development that are considered to have the cation of the residual fissile contents to be power needed for partial defect verification are verified. Calibration for different fuel designs discussed briefly below. and measurement conditions is, however, needed. Fork detector + HRGS 6.3 Radiation monitoring of Passive gamma and neutron detection, if com- casks bined with high resolution gamma-ray spectrometry, can be used for partial defect verification Radiation monitoring systems are used as a if the method is calibrated properly. The supporting measure, backing up camera surveil- measurement principle is based on the fact that lance. Such radiation detectors will use simple the neutron signal depends linearly on the neutron or gamma measurement, as appropriate, amount of material present. The neutron signal to facilitate the distinction between full and depends strongly also on the burnup (power 3- empty casks as well as indicating the direction 4). If material is missing, operator may declare of movement of casks. No NDA methods are a lower burnup yielding in a lower neutron rate. available that can be used to quantitatively The problem can be solved by correlating the distinguish between spent fuel and high level measured neutron rate with the measured Cs- waste. In addition, radiation detection systems 137 activity which depends linearly on the can be applied, which are suitable to dis- burnup and not on the amount of the material tinguish, at least qualitatively, casks containing present in an infinitely thick object like a fuel only high level waste from casks containing assembly. Different calibration curves are, how- spent fuel. ever, needed for different fuel types and initial enrichments. There are no readily available non-destructive assay technique that can be used for quantitative High energy gamma emission tomography verification of casks at the input of an operating repository. However, for unshielded and/or Passive gamma detection both under wet and shielded containers, there are research and dry conditions can provide a sensitivity level of development projects that could help resolve 1 missing rod for BWR assemblies and an this issue (Annex 3). estimated sensitivity level of 1-5 missing rods for large PWR assemblies, respectively. The The availability of quantitative verification tomographic methods is superior to any other techniques for casks would have a great impact method in the respect that no reference infor- on the safeguards approach. In practice, no mation is needed. Any assembly can be inde- measurements can easily be made under the pendently verified on partial defect level with- ground level. Only the surface facilities may out knowing any declared data. have locations for such activities.

18 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

6.4 Unsolved NDA problems * N°A for measurement of fissile material through the cask wall. The following verification needs include un- • NDA for measurement of Pu in spent fuel solved 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 (in case of mixed emplacement).

19 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

REFERENCES

[1] Report on Consultants' Group Meeting [7] Levai F, Tarvainen M. Partial defect on Safeguards for the Direct Final Dis- testing of spent fuel for final disposal posal of Spent Fuel in Geological Repo- using tomography. In: ANS Transactions, sitories. Vienna, 27 November-1 Decem- Washington 1995 ber 1995, IAEA, Report STR-305. [8] Mihalczo JT, Valentine TE, Mattingly [2] Larrimore J. IAEA Safeguards Criteria JK. Feasibility of Subcriticality and NDA 1991-1995: Where do they stand in Measurements for Spent Fuel by Fre- 1996?. In: Proc. 39th Annual Meeting, quency Analysis Techniques with 252Cf. INMM, 1996. In: Proc. Int. Top. Meeting on Nucl. Plant Instrumentation, Control, and Human- [3] IAEA Safeguards Glossary, 1987 Edition, Machine Interface Tech., May 6-9, 1996, Report IAEA/SG/INF/1 (Rev. 1), Vienna Pennsylvania State University, USA. 1987. [9] Mattingly JK, Valentine TE, Mihalczo [4] Levai F, Desi S, Tarvainen M, Arlt R. JT. Feasibility of Fissile Mass Assay of Detection of missing rods in a spent fuel Spent Nuclear Fuel Using 252Cf-Source- assembly by computed gamma emission Driven Frequency Analysis. In: Proc. of tomography. In: Proceedings of the 32nd Institute of Nucl. Materials Management INMM Annual Meeting, New Orleans, Annual Meeting, July 1996, Naples, USA, 1991: 1012-1017. Florida.

[5] Levai F, Desi S, Tarvainen M, Arlt R. Use [10] Valentine TE, Mattingly JK, Mihalczo of an underwater multidetector system JT. Dry Spent Fuel Cask Monitoring by for gamma emission tomography of spent 252Cf-Source-Driven Frequency Analysis fuel assemblies. In: Proceedings of the Measurements. In: Proc. of Institute of 15h ESARDA Symposium, Rome, Italy, Nucl. Materials Management Annual 1993:387-392. Meeting, July 1996, Naples, Florida.

[6] Levai F, Desi S, Tarvainen M, Arlt R. Use [11] Abdurrahman NM, Block RC, Harris of high energy gamma emission tomo- DR, Slovacek RE, Lee YD, Rodriguez- graphy for partial defect verification of Vera F. "Spent-Fuel Assay Performance spent fuel assemblies. Final Report on and Monte Carlo Analysis of the Rensse- the Task FIN A98 of The Finnish Support laer Slowing-Down-Time Spectrometer," Programme to IAEA Safeguards. Novem- Nucl. Science and Engineering (115) ber 1993, Report STUK-YTO-TR 56. 1993:279-296.

20 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

[12] Yong-Deok Lee et al. Neutron emission [20] Hsue ST, Crane TW, Talbert WL Jr, Lee tomography for nuclear fissile material JG Nondestructive Assay Methods for safeguards. Irradiated Nuclear Fuels, LA-6923, 1978. [13] Abdurrahnan NM et al. Detection sensitivity of an LS DTS for Spent-fuel Fissile [21] Prettyman TH, Betts SE, Taggart DP, Assay Application, ANS Transactions Estep RJ, Harlan RA. "Field Experience (68) 1993: 65-66. with a Mobile Tomographic Nondestruc- tive Assay System," Proc. of the 4th [14] Yong-Deok Lee et al. A Practical Spent- Nondestructive Assay and Nondestruc- fuel Assay Device Using the Lead Spec- tive Examination Waste Characterization trometer. ANS Transactions (68) 1993: Conference, Los Alamos National Lab- 66-67. oratory Report LA-UR-95-3501, (Oct. 1995). [15] Nilsson A, Danielsson N, Chen JD, Young GJ, Vodrazka P, Nakaoka. "Ceren- [22] Gozani T. Active Nondestructive Assay of kov Light Images of LWR Fuel Assem- Nuclear Materials. NUREG/CR-0602, blies," Proceedings of the 30th annual Jan. 1981. Meeting of the Institute of Nuclear Mate- rials Management, July 1989, Orlando, [23] Untermyer II S. Development and Test of Florida, USA, vol. XVHI: 902-909. Methods for the Nondestructive Assay of Spent Fuel Assemblies, EPRI NP-2812, [16] Chen JD, Attas EM, Young GJ, Burton 1983. GR, Keeffe R, Ward-Whate Hildingsson P, Nilsson A, Trepte O. "Spent Fuel [24] Bosler GE, Halbig JK, Klosterbuer SF, Verification Using an Ultraviolet Sensi- Menlove HO, Phillips JR. "Passive Neu- tive Charge Coupled Device," Proc. of tron Measurement Applications for Ir- the 37th annual Meeting of the Institute radiated Fuel Assemblies," Trans. Am. of Nuclear Materials Management, July Nucl. Soc. (39) 198: 348. 1996, Naples, Florida, USA. [25] Menlove HO, Reilly TD, Siebelist R. [17] Rinard PM, Bosler GE. "Safeguarding "The Verification of Reactor Operating LWR Spent Fuel with the Fork Detector", History Using the FORK Detector", Proc. Los Alamos National Laboratory, LA- of Institute of Nucl. Materials Manage- 11096-MS, (March 1988). ment Annual Meeting, (July 1996).

[18] Reilly D, Ensslin N, Smith H Jr, Kreiner [26] Menlove HO, Stewart JE, Qioa SZ, Wenz S. Passive Nondestructive Assay of Nu- TR, Verrecchia GPD. "Neutron Collar clear Materials, U. S. Nuclear Regulato- Calibration and Evaluation for Assay of ry Commission, NUREG/CR-5550, LA- LWR Fuel Assemblies Containing Burn- UR-90-732, 1991. able Neutrcn Absorbers," Los Alamos National Laboratory, LA-11965-MS, [19] Sher R, Untermyer II S. The Detection of (November 1990). Fissionable Materials by Non-destructive Means, American Nuclear Society, La [27] Menlove HO, Tesche CD, Thorpe MM, Grange, Illinois, 1980. Walton RB. "A Resonance Self-Indi- cation Technique for Isotopic Assay of Fissile Materials", Nuclear Applications, Vol. 6, (April 1969).

21 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

[28] Wuerz H. "A Nondestructive Measure- [36] Bosler GE, Phillips JR, Wilson WB, ment System for Spent LWR Fuel As- LaBauve RJ, England TR. "Production of semblies", in Proc. of the International Actinide Isotopes in Simulated PWR Symposium on Recent Advances in Nu- Fuel and Their Influence on Inherent clear Material Safeguards, IAEA report Neutron Emission", LANL Report LA- IAEA-SM-260, (November 1982). 9343, July 1982.

[29] Abhold ME, Hsue ST, Menlove HO, [37] Phillips JR, Bement TR, Hatcher CR, Walton G, Holt S. "The Design and Hsue ST, Lee DM. "Nondestructive Veri- Performance of the Research Reactor fication of the Exposure of Heavy-Water Fuel Counter", Proc. of Institute of Nucl. reactor Fuel Elements", LANL Report Materials Management Annual Meeting, LA-9432-MS, June 1982. (July 1996). [38] Rinard P. "A Spent-Fuel Cooling Curve [30] Pare VK, Mihalczo JT. "Reactivity from for Safeguards Applications of Gross- Power Spectral Density Measurements Gamma Measurements", LANL Report with Californium-252," Nucl. Sci. Eng. LA-9757-MS, April 1983. (56) 1975: 213. [39] Bosler GE, Rinard PM, Klosterbuer SF, [31 ] Pickrell MM, Kendall PK. "The Synchro- Painter J. "Automated Methods for Real- nous Active Neutron Detection System Time Analysis of Spent-Fuel measure- for Spent Fuel Assay," presented at the ment Data", In: Proceedings of the 29th I&EC Special Symposium, American Annual INMM Meeting, Vol. XVII, June Chemical Society, (Sept. 1994). 26-29, 1988.

[32] Eccleston GW, Menlove HO, Echo MW. [40] Nelson AJ, Bosler GE, Augustson RH, "A Measurement Technique for High Cowder LR. "Underwater Measurement Enrichment Spent fuel Assemblies and of a 15x15 MOX PWR-Type Fuel As- Waste Solids," Nuclear Materials Man- sembly", LANL Report LA-11850-MS, agement 8, (1979). December 1990.

[33] Cole JD, Aryaeinejad R, Drigert MW. [41] Hsue S-T, Menlove HO, Bacca J, Bosler "Gamma Neutron Assay Technique", pre- GE, Dye HR, Walton G, Halbig J, sentation at the National Spent Fuel Siebelist R. Research reactor Fork Users Program, Nondestructive Assay Meeting, Manual, LANL Report LA-12666-M. February 7-8, 1996 (unpublished). [42] Chesterman AS, Clark PA. Spent fuel and [34] Mihalczo JT, Hutchinson DP, Williams residue measurement instrumentation at JA, Thacker LH. "Criticality Evacuation the Sellafield nuclear fuel reprocessing Detectors That Locate the Accident," facility, In: INMM 36th Annual Meeting Trans. Am. Nucl. Soc. (75) 1996: 184- Proceedings, Palm Desert, California, 186. July 9-12, 1995: 185-192.

[35] Klosterbuer SF et. al. "Continuous Re- [43] Nicolau G, Koch L, "Characterization of mote Unattended Monitoring for Safe- Spent Nuclear Fuel by Non-destructive guards Data Collection Systems," in Assay", In: Proceedings of the 5th Int. Proceedings IAEA Symposium on Inter- Conf. on Radioactive Waste Measure- national Safeguards 1994: Vision for the ment and Env. Rem., Berlin 3-7 Sep- Future, IAEA, Vienna, (1994). tember, 1995.

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[44] Tarvainen M, Paakkunainen M, Tiitta A, [48] Eccleston GW, Johnson SS, Menlove Sarparanta K. "BWR SEAT, gross-defect HO, Van Lyssel TV, Black D, Carlson B, verification of spent BWR fuel", Report Decker L, Echo MW. "FAST Facility STUK-YTO-TR 72, STUK, Finland, Spent-Fuel and Waste Assay Instrument", April 1994. In: Proc. of the Conference on Safeguards Technology. The Process-safeguards In- [45] Tarvainen M, Backlin A, Hakansson A. terface, Hilton Head Island, South Caro- "Calibration of the TVO spent BWR lina, November 28-December 2, 1983, reference fuel assembly", Report STUK- USDOE, New Brunswick Laboratory, YTO-TR 37, STUK, Finland, February Argonne, Illinois, CONF-831106, August 1992. 1984: 253-264.

[46] af Ekenstam G, Tarvainen M. "Indepen- [49] Piper TC, Kirkham RJ, Eccleston GW, dent burnup verification of BWR-type Menlove HO. "Analysis of Spent, Highly nuclear fuel by means of the 137-Cs Enriched Reactor Fuel by delayed Neu- activity", Report STUK-A52, STUK, tron Interrogation", In: Proc. of Int. Finland, June 1987. Topical Meeting on Safety Margins in Criticality Safety, San Francisco, Cali- [47] Hildingsson L, af Ekenstam G, Tarvainen fornia, November 26-30, 1989, ANS, M, Tiitta A. "Feasibility of Gamma Ray 1989:168-175. Verification of Non-standard Fuel Items at CLAB", Report SKI 96:23, SKI, Sweden, March 1996.

23 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

ANNEX 1 DESCRIPTION OF NDA METHODS

Some of the techniques included in this review cooling time with the Cerenkov radiation can be do not have any potential for application to checked. The effectiveness of this method is spent fuel measurement, but nevertheless have reduced for verification of long cooled and/or been included for completeness. Measurement low burnup assemblies. This technique is only methods have been identified that are under applicable for wet storage of individual rods or development that could further aid in the assemblies, and does not have potential for control and safeguards of spent nuclear fuel, measurement of canisters. and where additional development is necessary for practical application, it will be so noted. The A 1.2 Passive gamma-ray NDA methods have been categorized into several distinct areas: passive and active gam- techniques ma, passive and active neutron, and other NDA Total counting techniques. In addition, fissile material tracking methods are discussed. Passive and active NDA The total gamma-ray activity of spent fuel techniques are those methods which can verify assemblies can be measured with scintillators, the presence of spent fuel or provide a quanti- thermoluminescent dosimeters, or ionization tative measure of the uranium and/or plutonium chambers as is done by the FORK [17, 24, 25, content in spent fuel. Passive and active NDA 38, 45] detector which the IAEA commonly methods are both useful depending on the type uses for spent fuel verification. For cooling of information needed. Fissile material tracking times greater than 1 year, the total gamma-ray methods employ measurement systems that can activity is roughly proportional to the burnup. aid in tracking spent fuel in a conditioning Consistency of operator declared values for facility or repository site. This article summa- burnup and cooling time can be determined by rizes many NDA methods and specifies their repeat measurements over several months time. limitations and the specific spent fuel meas- These measurements provide an indirect signa- urement areas in which these methods may be ture of the fuel burnup and age, which can be applied. related to Pu content by calculation.

Al.l Cerenkov Total gamma counting has limited potential for quantitative assay of fuel canisters. The gamma Measurements of Cerenkov light emanating attenuation of the canister and contents would from a spent fuel assembly in water has been have to be accurately calculated, and mixtures successfully used to verify spent fuel as- of low and high burnup fuel inside the same semblies in storage pools [15]. High energy canister would result in a non-unique solution electrons, gamma-rays, and neutrons from spent for the average calculated Pu content. However, nuclear fuel are capable of producing Cerenkov total gamma counting would provide a valuable light directly or indirectly in storage pools. component of a radiation signature, and could Cerenkov viewing devices (CVD) are used to be used to help verify the integrity of a canister identify physical characteristics of spent fuel with prior knowledge of its content. assemblies and to provide an indication of the burnup of an assembly in storage pools. These Spectroscopy devices are not capable of identifying missing fuel rods or the substitution of dummy fuel rods Gamma-ray spectroscopy is a common tech- in a spent fuel assembly. The absolute Cerenkov nique used to characterize radioactive materials light level and its decay with time is related to [18-20, 44—47]. The decay of radioactive iso- spent fuel burnup and it may provide a possi- topes results in the emission of gamma-rays and bility for quantitative measurements of the x-rays that are characteristic of the radioactive glow from spent fuel [16]. In practice, however, isotope. In spent fuel applications the spectral CVD verification is a single measurement, so lines are used to verify the presence of fission only consistency of the declared burnup and product actinides. For example, absolute ac-

24 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

DESCRIPTION OF NDA METHODS ANNEX 1 tivity measurements of 137Cs [45] can be The method has additional capabilities of performed on spent fuel rods and assemblies to measuring assemblies of new design, e.g. axial get a qualitative measure of the spent fuel enrichment variations and MOX. Gamma-ray burnup; however, this requires the assembly measurement can be made on the side of an attenuation and measurement geometry to be irradiated fuel assembly that is partially raised known or precisely calculated. Subsequent from the storage rack or moved to a meas- calculations can then be performed to obtain urement position. fissile mass content. The burnup of spent fuel can also be obtained from the ratio of activities A1.3 Active gamma-ray techniques for some fission product isotopes. The ratio of 134Cs to 137Cs is nearly linear with burnup; Densitometry however, corrections must be made for the decay of the isotopes for long burned fuels. The Densitometry entails the use of a photon source gamma activity ratio method must also be and a detector to measure the attenuation of the corrected for gamma-ray attenuation in as- photons from the source [18,19]. The photon semblies. Fission product activity ratios are source may emit particles at a single energy or easier to determine than absolute activities for multiple energies. These measurements are field measurements because only ratios of the useful for scanning rods and assemblies to detector efficiencies, which are a function of identify burnup variations and holes in as- gamma-ray energy, need to be known. Due to semblies. This method may not be useful for attenuation affects these methods are most quantitative measures because it measures the likely not applicable for canisters or casks of bulk attenuation of all materials in the rod and spent fuels. is best used as a means to map out holes in fuel assemblies. The applicability of this method is High energy gamma emission tomography limited mainly to individual fuel rods and (passive gamma) single fuel assemblies because canisters or casks will shield the assembly from the source. A tomographic measurement system has been developed and built for cross sectional viewing X-ray fluorescence of spent fuel assemblies and multi-element bottles (MEB) [4-7]. The method can reveal the X-ray fluorescence can be used to get a rod structure of spent fuel assemblies and the qualitative and quantitative assay of fissile presence of fuel assemblies in closed MEBs. material [18, 22]. X-rays emitted from an The hardware developed allows measurement ionized atom have energies that are cha- of the emission of fission product gamma-rays racteristic of the element. The X-ray intensity is (Eu-154 and Pr-144), in the air or under water. dependent on the concentration of the element The cross sectional image of a spent fuel and the intensity of the ionizing source. This assembly gives a rod-to-rod distribution of method cannot discriminate between isotopes gamma emitter concentration in the section of of the same elements because X-rays originate spent fuel mapped. The high sensitivity to the from electron transitions and not as a result of a removal of irradiated rods is explained by the nuclear process. Because of the low energy of following facts: (i) No need for a reference data X-rays this method may only be practically set, because the activity map provides an applied to assaying a single fuel rod. The high inherent rod-to-rod comparison of fission pro- background radiation from a spent fuel assem- duct activities, (ii) the inherent feature of the bly or canister and the complicated attenuation image reconstruction method provides noise for X-rays would make measurements in such reduction by averaging measured data obtained geometry impossible. at different views.

25 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR133

ANNEX 1 DESCRIPTION OF NDA METHODS

Photon induced fission neutrons [18-20]. A number of passive coincidence measurement have been developed This method requires the use of high energy that rely on the time correlation of radiation photons from an accelerator to induce fission in from spontaneous fission. Most spontaneous uranium and plutonium materials and meas- fissile isotopes yield on average more than two uring the subsequent emission of prompt and prompt neutrons and more than five prompt delayed neutrons from the fissile material [19]. gamma-rays emitted essentially simultaneously. However, the method is limited in that it is These measurements can be made in the difficult to discriminate between different fissile presence of background radiation because the materials and there is interference from (y,n) background radiation is random in nature. reactions in the materials. This method has not Passive coincidence measurements are suscep- been applied to a spent fuel assay. tible to self multiplication effects that obscure the actual amount of material This self- A 1.4 Passive neutron techniques multiplication effect can be significant for spent fuel systems with high (a,n) rates because the Total counting (cc,n) neutrons may induce fission in the fissile materials. This method has been applied to The total neutron emission rate from spent fuel fresh fuel [26], but has not been applied to spent can serve as an indicator of the burnup of the fuel. spent fuel [18, 19, 23, 24]. The total neutron counting method has some advantages and Neutron albedo disadvantages when compared to gamma-ray spectroscopy. Neutrons emitted from the spent Spent fuel assemblies emit spontaneous fission fuel are not attenuated as much as gamma-rays, and (a,n) neutrons at a high rate. Some of these and neutron counting measurements can be neutrons moderate in the water surrounding the made soon after the fuel is removed from the assembly and return to further induce fission in core whereas gamma-ray measurements cannot the assembly. If cadmium is placed around the because short lived decay products dominate assembly, these returning neutrons, or albedo the gamma-ray emissions. neutrons, are absorbed in the cadmium. Meas- urements made with and without a cadmium Two primary neutron-emitting isotopes are liner can be used to determine the multiplication 244Cm and 242Cm. To use total neutron counting in the assembly resulting from the albedo measurements the cooling time of the fuel must neutrons, therefore the fissile mass can be be precisely known because of the short half- estimated. This method does not have potential lives of these isotopes. The initial 235U enrich- to assay canisters or casks. ment and the irradiation history are factors that can affect the interpretation of the results of Multiplicity total neutron counting measurements. Veri- fication of reactor operating history can be Multiplicity analysis is an extension of co- accomplished by total neutron counting in incidence counting, but in addition to doubles, conjunction with total gamma counting as in two or more neutrons detected coincident in the modified FORK detector [25]. time, three or more coincident neutrons are observed. This higher order correlation provides Coincidence counting additional information that can be useful in separating the fission neutron source, (a,n) Passive neutron coincidence counting relies on neutron sources, and the effect from the assem- the presence of spontaneous fissile isotopes in bly's intrinsic multiplication. Since spent fuel fissile materials to produce time correlated assemblies in water have relatively high re-

26 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

DESCRIPTION OF NDA METHODS ANNEX 1 activity, there is a substantial probability of a An assembly monitor based on combined active neutron causing a fission chain, or super- and passive neutron counting has been used to fission, which can release large numbers of determine the burnup and initial enrichment of nearly simultaneous neutrons. These super- spent LWR fuel [28]. This method can provide fissions can be seen in the neutron multiplicity an indirect indication of Pu content through distribution, and yield valuable fissile content isotopic correlation. information. This method does not have potential to assay dry canisters, which have relatively Active coincidence counting low reactivity and complicated geometry. Active coincidence counting systems employ a Neutron resonance absorption neutron source to induce fission in fissile material and measure the subsequent emission All of the passive neutron techniques discussed of prompt radiation from the material [18, 19, so far only have the capability to indirectly 22]. The emission of prompt radiation occurs indicate the Pu or total fissile content in spent essentially simultaneously and provides a me- fuel by correlation with calculations. Neutron chanism to time correlate the detector response. resonance absorption techniques have the po- Coincidence counting measurements are typi- tential to passively separate the Pu content in cally not affected by the neutron source and are LEU fuel from other fissile isotopes. This independent of the background radiation from method utilizes the resonance structure in the (a, n) sources but are affected by the inherent neutron fission cross sections and relies on spontaneous fission rate in a spent fuel assem- detectors that are sensitive to the resonance bly. By removing the neutron source, a passive absorption lines [27]. This method has the coincidence count can be made to estimate the potential to verify Pu content in the outer pins contribution of the inherent spontaneous fission of an assembly underwater, with further devel- sources. A profile of the fissile mass content of opment, however does not have potential for a spent fuel assembly could be obtained by assay of canisters. performing a vertical scan of the assembly. This method has been applied to the measurement of A1.5 Active neutron techniques HEU spent fuel assemblies [29].

Total counting Noise analysis

Active total neutron counting uses a moderated The noise analysis method [30] uses a 252Cf neutron source to induce fission in a fissile source in an ionization chamber or a neutron material and a detection system to measure the generator to induce fissions in a fissile material emission of prompt neutrons from the material and two or more detectors to measure the [18, 19]. Prompt neutrons can be distinguished emission of prompt neutrons and gamma-rays from source neutrons using energy discrimi- from the fissile material. The source ionization nation. Moderated neutrons sources have been chamber or neutron generator provides a timed used to scan individual fuel rods and measure signal for each source event (99.9% of the the prompt neutrons emitting from the rod with spontaneous fissions are typically counted for fast neutron detectors. This methodology may 1 u.g 252Cf sources). The source and detector be applied to spent fuel assemblies but care signals are correlated with each other to obtain must be taken to discriminate the prompt signatures that depend only on the induced neutrons from background radiation. An ad- fission rate in the fissile material. The detector ditional measurement would be required to signals are also correlated with each other to estimate the contribution of the background obtain signatures that depend on both the radiation to the detector response. inherent and induced fission rates in the fissile

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ANNEX 1 DESCRIPTION OF NDA METHODS

material. The method has an advantage in that Pulse neutron—prompt/delayed the correlation between the source and the detector is independent of the inherent sources Methods have been employed to assay HEU and background radiation and depends only on spent fuel based on Cf shuffler techniques [32]. detector efficiency to the first power. Other The technique employs fast neutron irradiation correlation and coincidence techniques depend of the fuel by a Cf source followed by delayed on detection efficiency to the second power neutron counting after the source is transferred resulting in longer measurement times and to storage. Since the delayed neutron yield is require a passive measurement to estimate the significantly different for U and Pu fission, this contribution of inherent spontaneous fission method has the potential to separate the fissile sources to the detector response. Feasibility isotopes. There are presently no pulsed systems studies have been performed to assess the in operation measuring LWR fuel assemblies. application of this technique to spent fuel assay. Delayed neutron yields are only a small fraction These feasibility studies have shown that the of the prompt yield, therefore, measurement of correlation between the source and detector is LWR fuel would require extremely large neu- nearly linear with fissile mass content for a tron sources, on the order of 10" neutrons per single fuel assembly [8, 9, 34] and for an second, or neutron sources capable of producing unshielded canister of spent fuel [10] and that extremely large bursts of neutrons. Develop- this method could be used to scan a fuel ment of such sources is underway, with proto- assembly to obtain the profile of the fissile type sources expected within two years. This mass. method has the potential to determine the Pu content in assemblies and canisters. Pulse neutron—prompt Neutron radiography Pulsed neutron methods have been used to induce fission in fissile materials to measure the Neutron radiography uses a source of thermal, subsequent emission of delayed neutrons after epithermal or fast neutrons, and the penetrating the pulse [19, 22]. These methods require the qualities of neutrons are used in the study of use of a pulsed accelerator, a movable source, distribution of material variations. The image of or a shutter device to pulse the fissile material. transmitted neutrons can provide information of The detection system is triggered on after the the neutron attenuation property of materials. pulse to reduce the contribution of source Special features of the techniques are that (i) neutrons to the detector response. Repetitive uranium and each plutonium isotopes have burst of neutrons would be required to obtain different absorption properties for thermal neu- estimates of the total fissile mass in the spent trons, (ii) epithermal and fast neutrons are fuel assembly. Neutron source strengths must highly penetrating radiations capable of imag- be on the order of 108 or more neutrons per ing thick objects and (iii) pure neutron image second to induce a fission signal that is may be obtained even in case of gamma active comparable in size to the passive neutron spent fuels. The method is mainly used for background from a spent assembly. Novel imaging single spent fuel rods, especially for methods have been investigated to improve the post irradiation examination purposes. signal to background ratio using pulsed prompt sources. A factor of 4 improvement has been Neutron resonance absorption shown by a synchronous active technique using lock-in amplifiers [31]. These pulsed neutron Neutron resonance absorption is a neutron methods can potentially determine the fissile transmission measurement that uses a neutron content in an assembly and, with a large enough time of flight spectrometer to identify reso- source, can be used to assay canisters of fuel. nances in fissile materials [20]. There is little

28 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

DESCRIPTION OF NDA METHODS ANNEX1 interference from fission product resonances in tions [33]. The gamma emission from the the time of flight spectra This method may be fission fragments are coincidence counted to applied to a single fuel rod but is not practical reduce the gamma background from the other for a fuel assembly because of the assembly (non-correlated) gamma sources. A time gate is thickness and geometry. opened by triggering on the arrival of a specific gamma-ray from one of the two fission frag- Another active method for assay of spent fuel ments. Gamma spectra are then obtained in a that relies on resonance absorption is the narrow time window, which enables many slowing down time spectrometer [11]. This gamma lines to be discriminated from the method currently relies on a Linac to provide a background radiation. Among the gamma lines burst of neutrons that slow down by scatter with that can be seen in this window are the lines lead. The burst neutrons as they slow down from the partner fragment. The analysis of the passes through distinctive resonances in the spectra yields the isotope of the fissile material. fuel, producing fast fission neutrons in the Count rate limitations in the high resolution process. The fast neutrons are detected by gamma spectroscopy system may effectively threshold detectors and the time history of the limit the efficiency of this technique, however detected neutrons is related to the isotopic quantitative assay of the Pu content in spent content This method has the potential to fuel may be possible with further development. separate Pu from U, and may be able to assay LWR fuel assemblies. The development of a Calorimetry practical burst neutron source is essential to the application of the concept. Efforts are underway Calorimetry is the quantitative measurement of to develop such sources in the next several the heat produced by a sample. Calorimetry can years. This method does not have potential to be used to measure the heat production rate of assay canisters of fuel. radionuclides, and is in common use measuring Pu and tritium. The total heat produced by a A1.6 Other techniques spent fuel assembly or canister is a function of the initial fissile mass and the burnup. Multiple Gamma neutron assay technique (GNAT) calorimetric measurements of the spent fuel over time can be used to confirm characteristics This technique is a combination of active of spent fuel, and can be used on canisters and neutron interrogation with the measurement of casks. the prompt gamma emission from fission reac-

29 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR133

ANNEX 1 DESCRIPTION OF NDA METHODS A1.7 Table Al.I, NDA methods for spent HEU fuel Table ALL Nondestructive measurement methods applicability to spent fuel in a conditioning facility, high-enriched uranium fuel.

Technique Fuel Rod Fuel Low Capacity High Capacity Cask Limiting Assembly Canister Canister (Shielded) Conditions Cerenkov Indirect Indirect (D) Qualitative Qualitative N/A Wet only Passive gamma Total Counting Indirect Indirect Qualitative Qualitative N/A Spectroscopy Indirect Indirect Qualitative Qualitative N/A Tomography Indirect Indirect Indirect Qualitative N/A Active gamma Densitometry Qualitative N/A N/A N/A N/A X-ray fluorescence Qualitative N/A N/A N/A N/A Photon induced N/A N/A N/A N/A N/A fission Passive neutron Total Counting N/A N/A N/A N/A N/A Coincidence Counting N/A N/A N/A N/A N/A Neutron Albedo N/A N/A N/A N/A N/A Multiplicity N/A N/A N/A N/A N/A Resonance Absorption N/A N/A N/A N/A N/A Active neutron Total Counting Direct Total Direct Total Direct Total Qualitative N/A Fissile Fissile Fissile (D) Active Coincidence Direct Total Direct Total Direct Total Indirect N/A Fissile Fissile Fissile (D) Noise Analysis Direct Total Direct Total Direct Total Direct Total Qualitative Fissile Fissile Fissile (D) Fissile (D) (D) Pulse Neutron Direct Total Direct Total Direct Total Direct Total Qualitative Prompt Fissile Fissile (D) Fissile (D) Fissile (D) (D) Pulse Neutron N/A N/A N/A N/A N/A Prompt/Delayed Neutron Radiography Qualitative Qualitative Qualitative (D) Qualitative (D) N/A *) Neutron Resonance N/A N/A N/A N/A N/A Absorption Gamma Neutron Direct Indirect (D) N/A N/A N/A Active (GNAT) Calorimetry Indirect Indirect Qualitative Qualitative Qualitative *) = State of art uses accelerator facilities Qualitative = 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 on assumptions. 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 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

DESCRIPTION OF NDA METHODS ANNEX 1 A1.8 Table Al.II, NDA methods for spent LEU fuel Table A1.II. Nondestructive measurement methods applicability to spent fuel in a conditioning facility, low-enriched uranium fuel.

Technique Fuel Rod Fuel Assembly Low Capacity High Capacity Cask Limiting Canister Canister (Shielded) Condition Cerenkov Indirect Indirect (D) Qualitative Qualitative N/A Wet only Passive gamma Total Counting Indirect Indirect Qualitative Qualitative N/A Spectroscopy Indirect Indirect Qualitative Qualitative N/A Tomography Indirect Indirect Indirect Qualitative N/A Active gamma Densitometry Qualitative N/A N/A N/A N/A X-ray fluorescence Qualitative N/A N/A N/A N/A Photon induced N/A N/A N/A N/A N/A fission Passive neutron Total Counting Indirect Indirect Indirect Indirect Qualitative Coincidence Counting Indirect Indirect Indirect Qualitative NA/ Neutron Albedo Direct Total Direct Total N/A N/A N/A Fissile (D) Fissile (D) Multiplicity Indirect Direct Total Indirect Qualitative N/A Fissile (Wet) Indirect (Dry) Neutron Resonance Direct (D) N/A (Dry) N/A N/A N/A Absorption Qualitative (Wet)

Active neutron Total Counting Direct Total Direct Total Direct Total Qualitative N/A Fissile Fissile Fissile (D) Active Coincidence Direct Total Direct Total ? N/A N/A Fissile Fissile (D) Noise Analysis Direct Total Direct Total Direct Total Direct Total Qualitative Fissile (D) Fissile (D) Fissile (D) Fissile (D) (D) Pulse Neutron Direct Total Direct Total Direct Total Direct Total Qualitative Prompt Fissile Fissile (D) Fissile (D) Fissile (D) (D) Pulse Neutron Direct (D) Direct (D) Direct (D) Direct Total Qualitative Prompt/Delayed Fissile (D) (D) Neutron Radiography Qualitative Qualitative Qualitative Qualitative N/A Neutron Resonance Direct (D) Direct (D) N/A N/A N/A Absorption Gamma Neutron Direct Indirect (D) N/A N/A N/A *) Active (GNAT) Calorimetry Indirect Indirect Qualitative Qualitative Qualitative *) = State of art uses accelerator facilities Other markings as in Table A1.1.

31 UI No. Meas.Meas. principle Description Possibilities and limitations JO Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) 1 1 Passive gamma Shielded gamma spectrometer attached to air-filled Gross detect verification of spent LWR fuel. collimator pipe suspended over fuel element under Distinguishes irradiated fuel from irradiated X LWR SFAT Spent Fuel water. metal and dummies by detecting spent fuel Attribute Tester Manipulation using fuel handling machine (or specific radiation (finger prints). manually). No fuel movements needed, verification based Gamma Taucher Collects gamma-ray spectrum of fission products using on the top part of the assembly. n (Euratom) Na(I) or CdZnTe detectors and standard measurement No practical limitations of cooling time or r electronics. Used normally in point-wise mode, burnup. > jo scanning mode also possible. Cheap to construct, maintain and use and 00 Measurement time in point-wise mode 30-200 sec, modify for different facility conditions. > 20-40 assemblies per hour depending on measurement Assemblies with very short cooling time may conditions. be difficult to verify if located next to Attribute verified: Fission product gamma-rays e.g. assemblies with long cooling time and/or low c Cs-137,Pr-144. burnup. H Application: Verification of spent fuel specific O JO gamma-ray emissions, raw estimation of burnup and H cooling time. Passive neutron, U-shaped polyethylene fork with two fission chambers Gross defect verification of spent fuel. passive gamma (neutrons) and one ionization chamber (gamma-rays) in Consistency measurement of burnup and each tine. Cd-wrapped fission chamber for fast and bare cooling time based on gross neutron and gross GRAND/Fork fission chamber for thermal neutron detection. Fork is gamma detection. Detector placed around assembly to obtain neutron and gamma Neutron emission rate depends on initial o emission data. enrichment, irradiation history and amount of Fork is used from mobile bridge or fixed to a wall for material. oo underwater measurements. Form a bridge, assemblies Potential for partial defect verification by G do not need to be totally removed from the storage rack. combining fission product gamma rate z Measurement time 30-60 sec, 6-12 assemblies per (HRGS) with neutron rate. on hour. Distinguishes between irradiated MOX and H Attribute verified: Gross neutron emissions, mainly LEU fuel. JO from even Cm isotopes, gross gamma detection of I fission and activation products. on No. IVIeas. Meas. principle Description Possibilities and limitations Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) Application: Verification of declared burnup and Fuel lifting or movement needed, any vertical cooling time based on gamma and neutron emission position verifiable. 3 levels versus expected values for quantity and initial Determine irradiation history (number of enrichment. cycles); already been applied as well as unattended mode with camera surveillance. Visual Detection of Cerenkov glow using a CVD device Easy, fast and non-intrusive to use. with increased sensitivity to characteristic Cannot distinguish irradiated fuel from CVD Cerenkov wavelengths (UV). irradiated metal structures. Detected glow Viewing Device Positioned directly above the assembly, visual caused by any highly radioactive material detection of Cerenkov glow used to verify the under water. presence of irradiated fuel. Requires transparent, non-rippling pool water. Visual pattern of the top structure of the assembly Long cooling time (> 10 a), low burnup (< 10 used to verify the type of the object (assembly, MWd/kgU) and storage pattern limit usability. dummy, etc.). No recording of the observed pattern. Verifies only the top part of the assembly. Attribute verified: Gamma-ray emissions of the assembly. Light emission caused mainly by O secondary energetic electrons originating from Compton scattering of gamma-rays in water. Application: Verification that fuel assembly has been irradiated. o Passive neutron, Passive gamma and neutron detectors provide nuclear BUD measurement provides an estimate of the passive gamma signatures to permit verification of burnup of spent burnup. fuel assemblies. Neutron sensitivity depends on water thick- BUD Burnup Device/ Gamma and neutron detectors coupled to unattended ness between assembly and fission detector. CONSULHA video surveillance of events. Typically lower detection limit ~ 8 GWd/tU Containment and Two units together form the verification device BUD/ for neutrons > Surveillance for High CONSULHA BUD is more process oriented device while Activities Attribute verified: Neutron and gamma emissions. CONSULHA is more surveillance oriented. Application: Unattended monitoring of spent fuel signatures to verify nuclear material movements and inventories. No. Meas. Meas. principle "description Possibilities and limitations Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) Passive gamma, In passive mode, a 15 % HPGe detector and 5 fission Cooling time verification is based on the 134 154 IO6 137 X! o passive neutron, chambers are used in a monitor station for gamma-ray activity ratios Cs/ Eu and Ru/ Cs. z active neutron and neutron detection. Three separate values of irradiation are In active mode, a 252Cf source is transferred next to the determined using absolute I37Cs rate, iso-topic 10 l37 134 FPFM Feed Pond assembly and neutrons are measured from fissions ratios of including *Ru, Cs and Cs rates 3 Fuel Monitor induced by the source. as well as passive neutron rate originating 244 n Each assembly is measured up to four measurement mainly from Cm. rm heights during rotation in a monitor station. Local In operation at THORP, Sellafield. > jo values of cooling time, irradiation, initial and final Used for verification of burnup and cooling enrichments are calculated in addition to an average for time without using operator declared data each assembly. other than fuel type. Attribute verified: Burnup, cooling time, initial and residual enrichment 235U. Application: Unattended monitoring of spent fuel signatures to verify nuclear material movements and o inventories 50 Passive/active In passive mode, two detector heads under water with Gross defect verification of spent fuel. neutron/gamma collimated gamma and neutron detectors to measure Burnup verification, burnup evaluation emitted neutron and gamma radiation from the assembly code used for interpretation. PYTHON Spent brought inside the device. Consistency measurement of burnup and Fuel Detector, Device is positioned on the top of the storage rack, cooling time based on gross neutron and C/3 Cadarache assemblies lifted inside the device using fuel handling gross gamma detection. machine. Neutron emission rate depends on initial 2S2 In active mode, an external Cf source can be moved enrichment, irradiation history and amount C/3 adjacent to an assembly using a cable and neutrons of material. G counted from fissions induced in the residual fissile In active mode potential for partial defect material. verification by measuring amount of on Applicable both for criticality safety and safeguards residual fissile content. Calibration needed.

applications. Measurement time 600 sec per assembly in Fuel lifting or movement needed, any ffl jo passive mode. vertical position verifiable. C/3 No. Meas. Meas. principle Description Possibilities and limitations Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) 1 Attribute verified: Gross neutron emissions, mainly Commercially available. from even Cm isotopes, gross gamma detection of fission and activation products (passive mode). Neutrons originating from induced fissions in the residual fissile material (active mode). Application: Verification of declared burnup and cooling time based on gamma and neutron emission levels versus expected values for quantity and initial enrichment (passive mode). Verification of residual fissile contents (active mode). Passive gamma Detection of fission product gamma-rays (mainly 137Cs) Quantitative gross defect verification of using standard high resolution gamma spectrometry burnup. GBUV Gamma (HRGS) with HPGe detector and facility specific fuel Limited potential to reveal missing of fuel Burnup Verification handling. rods. Potential for partial defect verification if Assembly moved and/or rotated in front of a horizontal combined with passive neutron counting using gamma collimator. e.g. Fork detector. O Point-wise measurement allows any vertical position to Fuel movement is needed, verification of any •z be measured. Scanning mode allows verification of vertical position possible. most of the fuel volume. Verification based on 137Cs has no dependence Calibration source (l37Cs) or reference assembly used on initial enrichment or irradiation history. for normalization. No limitations of burnup or cooling time. Attribute verified: Fission product gamma-rays. Cooling time verification possible based on Application: Verification of declared burnup based on fission product isotopic ratios. on linear relationship between 137Cs activity and burnup. 8 2,4 Passive gamma Reconstruction of 2-D activity cross section map of Real partial defect verification on rod level. assembly from measured activity profiles. Measurement No operator declared information needed of TOMOGRAPHY time for a BWR assembly to reveal missing of a single burnup, cooling time or irradiation history. a Passive High Energy rod around 1 h using an array of 10 CdTe detectors. Possible to detect missing of individual LWR o Gamma Emission Activity profiles of emitted gamma-rays of l44Pr (2186 fuel rods, not a direct measurement of Pu 2 Tomography keV and 1489 keV) and 154Eu (1275 keV) are measured content. 3 ON No. Meas. Meas. principle Description Possibilities and limitations Point # Xnstn/method name (Techniques, handling, attribute verified, application) (level of verification) using a well collimated array of detectors (CdTe, Fuel movement and rotation needed. CdZnTe). Systematic scanning across the detector X o 2: array at specified angles. Mathematical program used to calculate the cross section activity map from measured profiles. Attribute verified: Fission product gamma-rays with o locations of emitting fuel rods. Application: Verification of integrity of spent fuel JO content in spent fuel assembly. 00 > Passive neutron Use of passive neutron multiplicity counting to Impractical for LWR spent fuel because determine content directly and to measure the 240Pu/ neutron rate from Cm » rate from Pu; poor Passive Neutron 244Cm ratio for the indirect verification of plutonium. Pu/Cm accuracy. Multiplicity Counting Neutron multiplicity of singles, doubles and triples neutrons for measuring 240Pu, 244Cm. X Attribute verified: Coincidence neutrons from 240Pu. o Application: Verification of plutonium content in 70 t-H spent fuel.

10 4 Active neutron A method measuring sub-critical neutron multiplication Large neutron source is required (5 x 109 n/ factor k of LWR fuel arrays. 252Cf source provides sec) to overcome background from 244Cm. C/E3T Cf-252-Source-Driven neutrons to initiate the fission chain multiplication Total fissile content is measured. O Noise measurements process in the spent fuel array. Cannot detect natural uranium substitution in Fission counters measure the frequency-dependent irradiated fuel. cross-power spectral densities (CPSDs) between a pair Very long measurement time needed (order of O of detectors located in or near the fissile material as days is possible). C well as measurements of CPSDs between these same z H detectors and a source of neutrons emanating from the ?3 252Cf source ionization chamber positioned in or near the fissile material. OJ 00 1 .1 No. Meas. IVIeas. principle Description Possibilities and limitations Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) H O Attribute verified: Quantity of fissile material versus reference condition. Application: Verification of fissile content of assemblies and casks. 11 2 Active neutron, Measurement performed in one axial position in of fuel and initial enrichment can be passive neutron approximately 15 min/assembly. Neutron source (252Cf determined in cases where 244Cm is the main and neutron detector are positioned on opposite sides of neutron emitter (cooling time > 2 a, burnup Spent fuel fuel assembly. Passive neutron emissions determined by >15 GWd/tU) by correlating the measured Identification System four neutron detectors positioned on four sides of neutron emission and multiplication. assembly. Active and passive neutron methods are used. Attribute verified: Passive and induced neutron Both uranium and MOX fuel assemblies can emissions from higher actinides and fissile materials. be measured. Application: Verification of burnup, total fissile, initial enrichment and type of fuel thus permitting calculation of amount of plutonium contained in uranium spent fuel assemblies. 12 4, 4b Active neutron 124Sb-Be neutron source activation of spent fuel assembly. Penetration of assembly with epithermal Active l24Sb-Be neutrons from the source and the subsequent detection Epithermal Neutron of the prompt and delayed neutrons produced by the O Method resulting fissions. Four 124Sb-Be sources penetrate the > assembly and provide a nearly uniform interrogation of the fuel. Assembly's high neutron and gamma back- 00 grounds are overridden by intense interrogation sources, thus ensuring a fission neutron production rate higher than the background neutron emission rate. Attribute verified: Total fissile content. Application: Verification of fissile actinide content of i spent fuel assemblies. O X 3 00 No. Meas. Meas. principle Description Possibilities and limitations I Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) 2 13 2 Active neutron Delayed neutrons, counted immediately after the Large background rate from Cm in LWR, so neutron source is removed, come from neutron rich you need a large 252Cf source (> 3 mg). o 2: Californium Shuffler fission products resulting from fissions induced by the Practical (and already used in Idaho Chemical neutron source presence. Number delayed neutrons is Processing Plant, US) on HEU fuel. proportional to number of induced fissions which is proportional to the fissile material present. o Attribute verified: Total fissionable content. m Application: Verification of uranium and plutonium > content of spent fuel assemblies. > 14 4 Active neutron Lead spectrometer consists of lead pile driven by Direct assay of total mass of fissile material, neutron pulse from accelerator. As neutron pulse slows spatial fissile material distribution using Lead Slowing Down down in lead, neutrons are focused in energy about a tomography under development. Time Spectrometer mean energy which is inversely proportional to the Only method that can measure Pu and U c H slowing down time. Amount of fissile material of the separately by detection of induced fast fission X fuel assembly inside the lead pile is obtained by neutrons as a function of time (between o measuring prompt fissile neutrons emitted from fuel interrogating pulses). 70 which reflects the effective spectrum averaged cross H Large, massive (heavy lead construction is < section of fissile isotopes present. needed) not practical for wider use. Attribute verified: Quantity of fissile isotopes. Intense pulse neutron source (Linac) is Application: Determination of quantity of uranium and needed. plutonium (and other actinides) in spent fuel C/5 H assemblies. o "11 15 4 Active neutron 14 MeV neutron generator and a novel detection system Laboratory testing only so far 2! 00 (digital lock-in amplifier) applied to direct measurement (LANL, LA-UR-94-3135). O > G Synchronous Active of spent fuel by detecting small signals in presence of Neutron Detection large background. Interrogating neutrons are non-thermal z and penetrating. 73 Attribute verified: Fission neutrons from fissile content. H Application: Verification of fissile material content in 70 spent fuel. No. Meas.Meas. principle Description Possibilities and limitations Point # Instn/method name (Techniques, handling, attribute verified, application) (level of verification) 16 2 Passive gamma Multiple detectors fixed on reactor face, triggered by Installed in Canada. movement of spent fuel out of reactor. Only shows movement and position of CANDU Fuel Monitor Identify reactor's region where fuel is expelled. expelled fuel; no burnup analysis. Secure network for information flow and analysis. Attribute verified: Gross gamma emissions from spent fuel fission products. Application: Unattended monitoring of spent fuel signatures to verify nuclear material movements and inventories. 17 4 Passive neutron, lonization chambers used as gamma radiation Installed in Joyo, Japan. passive gamma detectors for monitoring fuel movement and direction Installed in THORP (Transfer Channel of movement through channels. Monitor). Underwater Fuel Secure network for information flow and analysis. Used by Euratom and IAEA at Vandellos, Movement Attribute verified: Gross gamma emissions from Spain (SPEFAC, 2 gamma detectors) spent fuel fission products used to trigger cameras to No burnup analysis. photograph IDs on fuel or "skip" with fuel. o Application: Unattended monitoring of spent fuel z signatures to verify nuclear material movements and inventories. 18 4 Passive neutron "Can", GRAND-I and JSR-11 (#i) for low-power For low burnups, measure Pu directly with n research reactor fuel. passive count. 3 NDA of Research He neutron detectors and ionization chambers for rods • Used at Savannah River, US. GO Reactor Fuel #1 & #2 measured in groups of 4-5 at a time inside a canister (LANL) under water. Neutron rates correlated to exposure, ion chambers record 3 exposure profile. "Fork", GRAND-I and HRGS (#2) for low-power X research reactor fuel. o RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

ANNEX 2 LIST OF NDA INSTRUMENTS

'a.

DX VS cas 11 Hn i

g =

40 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

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 the irreversible nature of the final deposition, higher level verification of the correctness and completeness of the operator declared spent fuel data is needed. Even though some special NDA methods are known or have been shown feasible, additional research and development efforts are needed to further improve the techniques and optimize the instrumentation especially for partial defect 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

He detector

poly moderator

poly reflector

Figure A3.1 a) Arrangement (Source detector cask configuration). 41 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

ANNEX 3 RELEVANT R&D PROJECTS

0,12 fresh

_ 0,11 CM o

0,10

0,09

0,08 32 GWdMTU

0,07 0,10 0,15 0,20 0,25 0,30 0,35

Fissile material content, gcc

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 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

RELEVANT R&D PROJECTS ANNEX 3

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

Spent fuel assembly

Optimized LSDTS (36 ton lead slowing down medium)

Electron linear accelerator i (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 and 30 cm long)

I 15 ton lead shield CM

25cm ^14,6cm

160 cm

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

43 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR133

ANNEX 3 RELEVANT R&D PROJECTS

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::::::

\ Scanning Collimator Detector direction

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

AA STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

RELEVANT R&D PROJECTS ANNEX 3

ISSCaoo"

JTO.SBO

.SQG

Be«e* | fw«4

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

,fffffff|<

£>» «• • »

"lliliti Figure A3.5 A simulated tomograph revealing missing of one inner BWR assembly in a MEBfor 7 BWR assemblies.

45 RADIATION AND NUCLEAR SAFETY AUTHORITY STUK-YTO-TR 133

ANNEX 3 RELEVANT R&D PROJECTS

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 can be 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 STUK-YTO-TR 133 RADIATION AND NUCLEAR SAFETY AUTHORITY

SUMMARY OF WORK PLAN FOR SAGOR (TASK C 799) ANNEX 4

Revised: 15 February 15 1997

& Conditioning B. Operating i Phut Repository P Draft M$SP? Draft MSSPs && imp* Activity Report mi No. Activity Description 9m Date Da 0 Introduction %i9& US*' 5/96 US* M

1 Describe model facility 5/96 CAN* 5fl

2 Identify diversion paths and detection points m SEL 6/96 BEL n us* CAN* zm 3 Identify events and conditions requiring DIV 8/96 UK* W and examination procedures 4 Evaluate IAEA use of operator safeguards, wm .swE* 5/96 SWE* m safety, and process system outputs us us 5 Identify potentially applicable geophysical 6/96 CAN* m techniques FIN m FRA UK UK US us 6 Evaluate NDA techniques for spent fuel $!M FIN*' , 5/96 FIN Ht^ HA i verification and radiation monitoring HUH HUN* > US US : 7 Evaluate C/S techniques and integrated ®m us* 6/96 us* m verification systems for spent fuel monitoring I 8 Determine guidelines for acceptable tm us* 8/96 us* m safeguards approaches All All ^s 9 • Design safeguards approach and evaluate mr w 5/97 CAN* mw BB: (formerly candidate approaches Ait All m activities) • Develop redundancy and reliability 9-13 requirements for verification systems • Select safeguards approach and identify R&D needs • Develop QA program for detection systems • Specify system design requirements for model safeguards approach 4 10 (for- Integrated final report m as 9/97 us* w merly 14) All AS AGM-SAGOR mm AI 12/97 All 12*

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