A program of decommissioning of MR research reactor in the Kurchatov institute started in 2008. The decommissioning work presumed a preliminary stage, which included: removal of spent fuel from near reactor storage; removal of spent fuel assemble of metal liquid loop channel from a core; identification, sorting and disposal of radioactive objects from gateway of the reactor; identification, sorting and disposal of radioactive objects from cells of HLRW storage of the Kurchatov institute for radwaste creating form the decommissioning of MR. All these works were performed by a remote controlled means with use of a remote identification methods of high radioactive objects. A distribution of activity along high radiated objects was measured by a collimated radiometer installed on the robot Brokk-90, a gamma image of the object was registered by gammavisor. Spectrum of gamma radiation was measured by a gamma locator and semiconductor detector system. For identification of a presence of uranium isotopes in the HLRW a technique, based on the registration of characteristic radiation of U, was developed. For fragmentation of high radiated objects was used a cold cutting technique and dust suppression system was applied for reduction of volume activity of aerosols in air. The management of HLRW was performed by remote controlled robots Brokk-180 and Brokk-330. They executed sorting, cutting and parking of high radiated part of contaminated equipment. The use of these techniques allowed to reduce individual and collective doses of personal performed the decommissioning. The average individual dose of the personnel was 1,9 mSv/year in 2011, and the collective dose is estimated by 0,0605 man×Sv/year. Use of the remote control machines enables reducing the number of working personal (20 men) and doses. X-ray spectrometric methods enable determination of a presence of the U in high radiated objects and special cans and separation of them for further spent fuel inspection. The sorting of radwaste enabled shipping of the LLRW and ILRW to special repositories and keeping of the HLRW for decay in the Kurchatov institute repository.

The focus of this research was to analyze the dose rate profile around a waste repository using Monte Carlo techniques. Dose rates at various heights and distances were analyzed outside of the waste repository using MCNPX [1]. The heights measured extended the height of the building and the distances varied between 0 and 22 m away from the waste repository. The simulation data were fitted by smooth analytical functions for different height levels and distances, such that vertical and horizontal dose rates as functions of source-detector distance were achieved.

In the frame of its radwaste disposal research programme, SCK•CEN started the construction of the HADES underground research facility in 1980. Including several extensions and a comprehensive experimental programme, it has provided a lot of experience on monitoring. Monitoring is performed for many reasons: construction follow-up, field characterisation, investigation of phenomena, and model validations — in which the underground lab offers the opportunity for upscaling conventional laboratory set-ups. Construction monitoring has allowed to develop and optimise the underground construction techniques in a previously poorly known environment, resulting in a well-mastered application of mechanised methods for gallery construction with minimal damage to the host formation. Access to this formation also allows its characterisation, both geotechnical, geological and geochemical, and the detailed investigation of phenomena such as fracturing and oxidation. Finally, instrumented set-ups allow to test various numerical models by comparing the observations with the predicted behaviour. The specific conditions of the underground laboratory put particular requirements to the sensors. These conditions include the long-term nature of many set-ups — typically several years to decades, the inaccessibility of many sensors after installation, high mechanical and water pressures, and corrosion. Combined with the fact that many sensors are custom made, obtaining and maintaining a fully functional instrumented set-up can be challenging. A lot of experience has therefore been gained which is very valuable when designing the monitoring of radwaste repositories — and it has allowed us to determine the critical success factors for monitoring. Engineers tend to look at this first from a technical viewpoint — and there are many technical aspects indeed that determine the reliability of monitoring. A first one is the combination of different observations (“redundancy”) which can be implemented by the use of several sensors, different sensor principles, different (coupled) parameters, and the combination of point measurements with geophysical techniques. Cabling is also a critical issue as it is often considered as the primary enemy of barrier integrity. Minimal cabling techniques, such as distributed fibre optic monitoring and wireless signal transmission, therefore get increasing attention. Also the interpretation of the monitoring data — in particular those that are perceived as “wrong” or “unexpected”, needs sufficient attention. The long-term experience has however also shown that the design of a monitoring programme must look beyond the technical part. In particular for long-term applications, issues such as data management and record keeping are vital to guarantee success in this.

Radioactive waste systems and structures (RWSS) are safety-critical facilities in need of monitoring over prolonged periods of time. Structural health monitoring (SHM) is an emerging technology that aims at monitoring the state of a structure through the use of networks of permanently mounted sensors. SHM technologies have been developed primarily within the aerospace and civil engineering communities. This paper addresses the issue of transitioning the SHM concept to the monitoring of RWSS and evaluates the opportunities and challenges associated with this process. Guided wave SHM technologies utilizing structurally-mounted piezoelectric wafer active sensors (PWAS) have a wide range of applications based on both propagating-wave and standing-wave methodologies. Hence, opportunities exist for transitioning these SHM technologies into RWSS monitoring. However, there exist certain special operational conditions specific to RWSS such as: radiation field, caustic environments, marine environments, and chemical, mechanical and thermal stressors. In order to address the high discharge of used nuclear fuel (UNF) and the limited space in the storage pools the U.S. the Department of Energy (DOE) has adopted a “Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste” (January 2013). This strategy endorses the key principles that underpin the Blue Ribbon Commission’s on America’s Nuclear Future recommendations to develop a sustainable program for deploying an integrated system capable of transporting, storing, and disposing of UNF and high-level radioactive waste from civilian nuclear power generation, defense, national security, and other activities. This will require research to develop monitoring, diagnosis, and prognosis tools that can aid to establish a strong technical basis for extended storage and transportation of UNF. Monitoring of such structures is critical for assuring the safety and security of the nation’s spent nuclear fuel until a national policy for closure of the nuclear fuel cycle is defined and implemented. In addition, such tools can provide invaluable and timely information for verification of the predicted mechanical performance of RWSS (e.g. concrete or steel barriers) during off-normal occurrence and accident events such as the tsunami and earthquake event that affected Fukushima Daiichi nuclear power plant. The ability to verify the conditions, health, and degradation behavior of RWSS over time by applying nondestructive testing (NDT) as well as development of nondestructive evaluation (NDE) tools for new degradation processes will become challenging. The paper discusses some of the challenges associated to verification and diagnosis for RWSS and identifies SHM technologies which are more readily available for transitioning into RWSS applications. Fundamental research objectives that should be considered for the transition of SHM technologies (e.g., radiation hardened piezoelectric materials) for RWSS applications are discussed. The paper ends with summary, conclusions, and suggestions for further work.

Radiological characterization of nuclear waste is essential for storage sites management. However, most of Long-Lived RadioNuclides (LLRN), important for long-term management, are difficult to measure since concentration levels are very low and waste matrices generally complex. In an industrial approach, LLRN concentrations are not directly measured in waste samples but assessed from scaling factors with respect to easily measured gamma emitters. Ideally, the key nuclide chosen ( 60 Co, 137 Cs) should be produced by a similar mechanism (fission or activation) as the LLRN of interest and should have similar physicochemical properties. However, the uncertainty on the scaling factors, determined from experimental and/or calculation data, can be quite important. Consequently, studies are performed to develop analytical procedures which would lead to determine precisely the concentration of LLRN in nuclear waste. In this context, the aim of this study was to determine the concentrations of three LLRN: 129 I (T 1/2 = 15.7×10 6 a), 41 Ca (T 1/2 = 9.94×10 4 a) and 10 Be (T 1/2 = 1.387×10 6 a) in spent resins used for primary fluid purification in Pressurized Water Reactors using Accelerator Mass Spectrometry (AMS) for measurement. The AMS technique combined mass spectrometry and nuclear physics to achieve highly efficient molecular and elemental isobars separation. Energies of several Million Electron-Volt transferred to the ions in the first accelerating part of specifically developed tandem accelerators lead to molecular isobars destruction through interaction with the argon gas used to strip the injected negative ions to positive ones. At the exit of the tandem accelerator, the energy acquired in both accelerating parts allows an elemental isobars separation based on their significantly different energy loss (dE) while passing through a thickness of matter dx that is proportional to their atomic number (Z) and inversely proportional to ions velocity (ν) according to the Bethe-Block law (1). (1) d E d x = k * Z 2 ν 2 The use of a particle accelerator in conjunction with a selective ion source, mass and energy filters and a high-performance detector thus allow unambiguously identifying and measuring analyte concentration against much more abundant interfering isobars. The development of AMS and of related applications have recently been extensively reviewed [1–3]. Up to now, the potentialities of the accelerator mass spectrometry technique were explored for the measurement of cosmogenic radionuclides produced in the Earth’s environment either in the atmosphere or in the Earth’s crust (in situ-production). Many applications aiming to date and/or quantify Earth surface processes have been developed in the fields of geology, geomorphology and planetary sciences as well as archeology paleoanthropology and biomedicine. The present study extends the scope of AMS to nuclear industry. Because AMS facilities are not widely accessible and difficult to handle, LLRN concentrations in nuclear waste are usually determined using Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and radiometric techniques. However for the measurement of very low LLRN concentrations, AMS becomes the most effective measurement method with detection limits of 10 5 –10 6 atoms per sample. In this study, AMS measurements were performed using the French AMS national facility ASTER located at the Centre Européen de Recherche et d’Enseignement des Géosciences de l’Environnement (CEREGE). The challenge was to define a chemical treatment procedure allowing the measurement of the three nuclides, 10 Be, 41 Ca and 129 I, by AMS. Each method selection was based on three main requirements: 1) a quantitative recovery in solution of Be, Ca, I and key radionuclides after resin mineralization, 2) a selective extraction from the sample matrix and the separation from β-γ emitters ( 3 H, 14 C, 55 Fe, 59 Ni, 60 Co, 63 Ni, 90 Sr, 125 Sb, 134 Cs, 137 Cs) and isobars, 3) the precipitation of each element under the best suited forms (i.e. AgI, CaF 2 , BeO) for AMS measurements. The chosen methods were optimized on synthetic solutions and finally applied for the determination of the three LLRN concentrations in spent resins from a 900 MWe Nuclear Power Reactor.

Knowledge of the radiological state of processes and equipment of a nuclear facility is essential to supervise a wide variety of sensitive tasks: building of intervention scenarios in order to optimize maintenance or dismantling operations, optimization of waste categorization, monitoring the effectiveness of decontamination processes, monitoring of nuclear facility decommissioning, etc. In order to meet the diversity of the issues involved, the CEA has developed in situ radiological characterization methods and techniques to acquire reliable radiological data. The data gathered is necessary to build robust radiological models which can be used as input data for dismantling studies. Over the last 30 years, the main nuclear measurement techniques, such as gamma imaging and gamma spectrometry, have been widely deployed by the CEA on many facilities under dismantling and more recently, on the Phénix nuclear power plant. Phénix was a small-scale prototype of a sodium-cooled fast breeder reactor, located at the Marcoule nuclear site. These techniques have been implemented on this reactor in order to meet the increased need for radiological knowledge to prepare for future dismantling operations following its final shutdown in 2009. This paper will focus on the description of three radiological characterization methods which take advantage of advanced nuclear measurement techniques. For each method, an example of a specific application on the Phénix reactor will be presented. Firstly, the so-called “gamma scanning” method will be explained. The objective of this method is to determine the activity profile of equipment based on collimated gamma spectrometry measurements with compact probes like CdZnTe. This method was applied to a neutron shielding of the reactor core to estimate the 60 Co activity profile. Then the measured activities helped to validate the theoretical activities resulting from neutron activation calculations. Secondly, this paper will focus on the interest of combining different measurement techniques such as gamma imaging, gamma spectrometry and collimated/uncollimated dose rate mapping to characterize complex equipment or processes. In this case, a specific methodology was developed to define the radiological state of a shielded cell used for the processing of irradiated nuclear fuels. Finally, an isotopic characterization technique using a high purity germanium detector will be discussed. This technique was applied to a non-irradiated fertile fuel sub-assembly in order to determine the level of uranium enrichment. The processing was carried out by three types of analysis: two automated, with the MGA-U and IGA software, and one manual.

The underwater spectrometric system for survey the bottom of material science multi-loop reactor MR ponds was elaborated. This system uses CdZnTe (CZT) detectors that allow for spectrometric measurements in high radiation fields. The underwater system was used in the spectrometric survey of the bottom of the MR reactor pool, as well as in the survey located in the MR storage pool of highly radioactive containers and parts of the reactor construction. As a result of these works irradiated nuclear fuel was detected on the bottom of pools, and obtained estimates of the effective surface activity detected radionuclides and created by them the dose rate.

In a previous paper (1) it was proposed that a simple matrix inversion method could be used to extract source distributions from gamma-count maps, using simple models to calculate the response matrix. The method was tested using numerically generated count maps. In the present work a 100 kBq Co 60 source has been placed on a gridded surface and the count rate measured using a NaI scintillation detector. The resulting map of gamma counts was used as input to the matrix inversion procedure and the source position recovered. A multisource array was simulated by superposition of several single-source count maps and the source distribution was again recovered using matrix inversion. The measurements were performed for several detector heights. The effects of uncertainties in source-detector distances on the matrix-inversion method are also examined. The results from this work give confidence in the application of the method to practical applications, such as the segregation of highly active objects amongst fuel-element debris.

We propose a radiation detection system which generates its own discrete sampling distribution based on past measurements of background. The advantage to this approach is that it can take into account variations in background with respect to time, location, energy spectra, detector-specific characteristics (i.e. different efficiencies at different count rates and energies), etc. This would therefore be a “machine learning” approach, in which the algorithm updates and improves its characterization of background over time. The system would have a “learning mode,” in which it measures and analyzes background count rates, and a “detection mode,” in which it compares measurements from an unknown source against its unique background distribution. By characterizing and accounting for variations in the background, general purpose radiation detectors can be improved with little or no increase in cost. The statistical and computational techniques to perform this kind of analysis have already been developed. The necessary signal analysis can be accomplished using existing Bayesian algorithms which account for multiple channels, multiple detectors, and multiple time intervals. Furthermore, Bayesian machine-learning techniques have already been developed which, with trivial modifications, can generate appropriate decision thresholds based on the comparison of new measurements against a non-parametric sampling distribution.

For making the spectrometric determination of the exposure rate from the environment as well as the radioactive material more practical, an accurate calculation method of the individual exposure rate for the detected gamma nuclides from that spectrum should be suggested without the sophisticated calibration procedure. In this study, the calculation method for the individual exposure rate for detected gamma nuclides from a 3″×3″ NaI(Tl) detector was suggested by introducing the concept of the dose rate spectroscopy and the peak-to-total ratio in the energy spectrum for the exposure rate, which means just a form of multiplied counts and the value of a G-factor in the spectrum.

This paper presents aspects of site decommissioning and clearance of a former fuel fabrication facility (development and production of fuel assemblies for research reactors and HTR) at Hanau (Germany). The main pathways for environmental contamination were deposition on soil surface and topsoil and pollution of deep soil and the aquifer by waste water channel leakage. Soil excavation could be done by classical excavator techniques. An effective removal of material from the saturated zone was possible by using advanced drilling techniques. A large amount of demolished building structure and excavated soil had to be classified. Therefore the use of conveyor detector was necessary. Nearly 100000 Mg of material (excavated soil and demolished building material) were disposed of at an underground mine. A remaining volume of 700 m 3 was classified as radioactive waste. Site clearance started in 2006. Groundwater remediation and monitoring is still ongoing, but has already provided excellent results by reducing the remaining Uranium considerably.

The ability to monitor critical environment parameters of nuclear plants at all times, particularly during and after a disruptive accident, is vital for the safety of plant personnel, rescue and recovery crews, and the surrounding communities. Conventional hard-wired assets that depend on supplied power may be decimated as a result of such events, as witnessed in the Japanese Fukushima nuclear power plant in March 2011. Self-powered monitoring devices operating on a wireless platform, on the other hand, may survive such calamity and remain functional. The devices would be prepositioned at strategic locations, particularly where the dangerous build-up of contamination and radiation may preclude subsequent manned entrance and surveillance. Equipped with sensors for β-γ radiation, neutrons, hydrogen gas, temperature, humidity, pressure, and water level, as well as with criticality alarms and imaging equipment for heat, video, and other capabilities, these devices can provide vital surveillance information for assessing the extent of plant damage, mandating responses (e.g., evacuation before impending hydrogen explosion), and enabling overall safe and efficient recovery in a disaster. A radio frequency identification (RFID)-based system — called ARG-US — may be modified and adapted for this task. Developed by Argonne for DOE, ARG-US (meaning “watchful guardian”) has been used successfully to monitor and track sensitive nuclear materials packages at DOE sites. It utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users when any of the preset sensor thresholds is violated. By adding plant-specific monitoring sensors to the already strong sensor suite and adopting modular hardware, firmware, and software subsystems that are tailored for specific subsystems of a plant, a Remote Area Modular Monitoring (RAMM) system, built on a wireless sensor network (WSN) platform, is being developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations.

In the frame of the decommissioning of nuclear power plants or laboratories, the penetration depth of the contamination in concrete surfaces: walls, floor, is unknown. Its knowledge requires sample analysis, that is time consuming and expensive. The main goal of the work is to define and evaluate a non-destructive measurement technique for the evaluation of the contamination depth in concrete. Estimation of accuracy of measurements for different contamination levels, time of measurement, value of natural radionuclides (NRN) concentration in concrete and background radiation dose were carried out. The type of relevant detector depends of selected limiting sensitivity, the weight and sizes of the device.

The paper relates to the determination of the amount of nuclear material contained in a closed, concrete lined vault at the Aktau fast breeder reactor in Kazakhstan. This material had been disposed into the vault after examination in an experimental hot cell directly above the vault. In order to comply with IAEA Safeguards requirements it was necessary to determine the total quantities of nuclear materials — enriched uranium and plutonium — that were held with Kazakhstan. Although it was possible to determine the inventory of all of the accessible nuclear material — the quantity remaining in the vault was unknown. As part of the Global Threat Reduction Programme the UK Government funded a project to determine the inventory of these nuclear materials in this vault. This involved drilling three penetrations through the concrete lined roof of the vault; this enabled the placement of lights and a camera into the vault through two penetrations; while the third penetration enabled a lightweight manipulator arm to be introduced into the vault. This was used to provide a detailed 3D mapping of the dose rate within the vault and it also enabled the collection of samples for radionuclide analysis. The deconvolution of the 3D dose rate profile within the vault enabled the determination of the gamma emitting source distribution on the floor and walls of the vault. The samples were analysed to determine the fingerprint of those radionuclides producing the gamma dose — namely 137 Cs and 60 Co — to the nuclear materials. The combination of the dose rate source terms on the surfaces of the vault and the fingerprint then enabled the quantities of nuclear materials to be determined. The project was a major success and enabled the Kazakhstan Government to comply with IAEA Safeguards requirements. It also enabled the UK DECC Ministry to develop a technology of national (and international) use. Finally the technology was well received by IAEA Safeguards as an acceptable methodology for future studies.

Cambridge University had worked in the 1960s and 1970s with Pa-231, a decay product of U-235. The fume cupboards discharged into ventilation ducting made from asbestos cement. The university wished to refurbish the laboratory and the RPA had negotiated over many years with the Environment Agency to set up a project to remove the ducting both to reduce the radiological hazards and as part of a programme to remove unwanted circuits and upgrade the ventilation system to modern standards. Contamination levels were significant and low dose rates were measurable on the external surface. The aim was to be able to remove the ducting and treat it as asbestos waste, rather than to have to treat the debris as asbestos contaminated radioactive waste. The age of the contaminant was such that a large fraction of the decay chain had grown in, giving a mixture of alpha, beta and gamma emissions. The most useful nuclides for surface monitoring were Pb-211 and Tl-207, both of which are energetic beta emitters. A wide energy range beta detector was used, but it was fitted with a filter to absorb any alpha radiation which otherwise would have contributed to the signal for good surfaces but not for dusty, damp or rough surfaces and would have contributed to the uncertainty in the activity assessment. Samples were checked using gamma spectrometry to confirm that only Pa-231 and its progeny were present in significant quantities. The gamma spectrum is complicated and this paper describes the difficulties in confirming that the spectrum only contained the Pa-231 decay chain. The vast majority of the contaminated ducting was successfully consigned as asbestos, rather than radioactive, waste. The other problem was dealing with the soft waste produced during the dismantling process. This was monitored using simple equipment and it was possible to demonstrate that it could be disposed of with the rest of the waste under the relevant UK legislation.

Normally, beta and alpha surface contamination monitors are used with a simple counting threshold, i.e. any pulse over a predetermined amplitude is counted. This is very different from gamma monitoring, where the use of counting windows is very popular and the use of full multi-channel analysis is common. Many current surface contamination ratemeters have the capacity to drive dual phosphor detectors and can be set up to provide beta and alpha channels. Effectively, the beta channel is a counting window, i.e. all pulses which are bigger than the threshold and smaller than the alpha threshold are counted. Larger pulses go into the alpha channel. This paper addresses how this can be used with beta only and alpha only detectors to provide information on the source. The detector is set up conventionally to a defined point for the lowest beta energy anticipated. The instrument is then switched to alpha + beta mode and the alpha threshold set to 3 times the beta threshold. With this set up, the alpha to beta channel count rate ratio varies smoothly by a factor of 14 between Y-90 (Emax 2.27 MeV) and C-14 (Emax 0.16 MeV). Hence the instrument can be used to estimate the energy of an unknown beta contaminant or to confirm that a mixed beta fingerprint has essentially the same mix. The same approach can be used with alpha probes to confirm the source quality. The main worry with alpha monitoring is the surface condition. A poor surface condition will lead to a low count rate. Using the channel ratio method will identify grubby sources. The resulting ratio can be used either as a go/no trigger, i.e. any surface with a low ratio will be treated as untrustworthy, or alternatively the ratio can be used to correct the reading to give a better estimate of surface activity.

The Springfields facility manufactures nuclear fuel products for the UK’s nuclear power stations and for international customers. Fuel manufacture is scheduled to continue into the future. In addition to fuel manufacture, Springfields is also undertaking decommissioning activities. Today it is run and operated by Springfields Fuels Limited, under the management of Westinghouse Electric UK Limited. The site has been operating since 1946 manufacturing nuclear fuel. As part of the decommissioning activities, there was a need was to quantify contamination in a large redundant building. This building had been used to process uranium derived from uranium ore concentrate but had also processed a limited quantity of recycled uranium. The major non-uranic contaminant was Tc-99. The aim was to be able to identify any areas where the bulk activity exceeded 0.4 Bq/g Tc-99 as this would preclude the demolition rubble being sent to the local disposal facility. The problems associated with this project were the presence of significant uranium contamination, the realisation that both the Tc-99 and the uranium had diffused into the brickwork to a significant depth and the relatively low beta energy of Tc-99. The uranium was accompanied by Pa-234m, an energetic beta emitter. The concentration/depth profile was determined for several areas on the plant for Tc-99 and for uranium. The radiochemical analysis was performed locally but the performance of the local laboratory was checked during the initial investigation by splitting samples three ways and having confirmation analyses performed by 2 other laboratories. The results showed surprisingly consistent concentration gradients for Tc-99 and for uranium across the samples. Using that information, the instrument response was calculated for Tc-99 using the observed diffusion gradient and averaged through the full 225 mm of brick wall, as agreed by the regulator. The Tc-99 and uranium contributions to the detector signal were separated using a simple absorber, which essentially eliminated the Tc-99 count rate and reduced the uranium contribution only marginally. The outcome of the project was that it was possible to demonstrate that the complete building met the criterion for acceptance at the local waste facility.

Alpha contamination detection usually relies on good surface conditions — unobstructed, smooth, clean and flat. However, in many circumstances, the conditions may be difficult, or it may be that there is a significant chance that contamination has been deliberately or accidentally painted over. Hence an alternative method of monitoring would be useful. In some areas of fuel manufacture and reprocessing, and in weapons manufacture, potential contamination is dominated by plutonium isotopes and by ingrown Am-241, derived from the beta decay of Pu-241. Plutonium is often thought of as a mainly an alpha emitter, together with a few weak betas detectable by direct means only with extreme difficulty. However, the alpha emitting isotopes of plutonium also emit significant L x-rays in the 11 to 20 keV energy range, as does Am-241, which also emits a 60 keV gamma with a 36% probability. These X-rays are unattenuated to any extent in air over a range of 1 metre. They also penetrate paint significantly, which makes them detectable by a suitable probe. The Fidler probe was designed as an efficient detector of these X and gamma radiations. The window is thin beryllium and the scintillator is thin sodium iodide. This leads to a very efficient detection of both the X-rays in question and the 60 keV gamma radiation while keeping the background as low as possible. The signal from such a detector can be processed in several ways — gross counting above a threshold, counting in regions of interest or full spectrometry. The advantages of the latter include the minimisation of background, easy background correction, the ability to use the recorded X-ray spectrum to correct the measured counts and the identification of the presence of other gamma emitters. Nuvia Ltd have been developing techniques to look at the spectrum from Fidler probes and derive a calibration factor, either from using the 60 keV gamma measurement only where the fingerprint is well known and stable, or by using samples of paint and small area sources to actually measure the transmission factor where the Am-241 fraction is variable. Using these techniques, it is possible to obtain a good estimate of the plutonium contamination level under paint over a range of conditions.

Material control and accountability (MC&A) operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. MC&A activities, from monitoring to inventory measurements, provide critical information about target materials and define security elements that are useful against insider threats. However, these activities have been difficult to characterize in ways that are compatible with the path analysis methods that are used to systematically evaluate the effectiveness of a site’s protection system. The path analysis methodology focuses on a systematic, quantitative evaluation of the physical protection component of the system for potential external threats, and often calculates the probability that the physical protection system (PPS) is effective ( P E ) in defeating an adversary who uses that attack pathway. In previous work, Dawson and Hester observed that many MC&A activities can be considered a type of sensor system with alarm and assessment capabilities that provide reccurring opportunities for “detecting” the status of critical items. This work has extended that characterization of MC&A activities as probabilistic sensors that are interwoven within each protection layer of the PPS. In addition, MC&A activities have similar characteristics to operator tasks performed in a nuclear power plant (NPP) in that the reliability of these activities depends significantly on human performance. Many of the procedures involve human performance in checking for anomalous conditions. Further characterization of MC&A activities as operational procedures that check the status of critical assets provides a basis for applying human reliability analysis (HRA) models and methods to determine probabilities of detection for MC&A protection elements. This paper will discuss the application of HRA methods used in nuclear power plant probabilistic risk assessments to define detection probabilities and to formulate “timely detection” for MC&A operations. This work has enabled the development of an integrated path analysis methodology in which MC&A operations can be combined with traditional sensor data in the calculation of PPS effectiveness. Explicitly incorporating MC&A operations into the existing evaluation methodology provides the basis for an effectiveness measure for insider threats, and the resulting PE calculations will provide an integrated effectiveness measure that addresses both external and insider threats. The extended path analysis methodology is being further investigated as the basis for including the PPS and MC&A activities in an integrated safeguards and security system for advanced fuel cycle facilities.

During operation and maintenance, or decommissioning of nuclear power plant, various kind of waste should be treated, and exposure control is also required. These wastes have a wide range of contamination, different composition of nuclides, and a different shape, so each measurement instrument would be optimized for its use especially for very low level radioactivity measurement. TOSHIBA provides appropriate equipment for any needs to discriminate the very low and non radioactive waste to save cost of waste disposal, based on our original and innovative technology. For alpha emitting nuclides, we are ready to supply instruments based on ionized particle measurement technology. For beta, gamma-emitting nuclides, we are ready to customize a shape of detector based on our original plastic scintillation material. Some examples are introduced.