Mitsutoshi Suzuki

Study on Loss Detection Algorithms Using Tank Monitoring Data

An application of solution monitoring to material balance evaluation has been investigated with actual data from the Solution Monitoring Management System (SMMS) in the Tokai Reprocessing Plant (TRP). Loss detection capabilities of the proposed multivariate statistical methods are examined numerically for a simulated loss corresponding to parameters of a significant quantity in the wait mode. Multiscale statistical analysis as well as multivariate cumulative sum analysis are successfully used to demonstrate the protracted-loss detection. Because the actual tank data is composed of records in both the wait and transfer modes, the tank-to-tank transferring data is extracted from the sequence of monitoring data, and the error model effectiveness is evaluated in comparison with calculations. Although the data taken from the SMMS does not describe the entire solution process, an extended application of solution monitoring to the nuclear material accounting is advanced using the plant data, and future research subjects are described.

Loss detection results on simulated tank data modified by realistic effects

Solution monitoring (SM) is a type of process monitoring (PM) intended to improve nuclear safeguards in large commercial facilities that contain solutions. Typically, masses (M) and volumes (V) are estimated from frequent in-process measurements. Transfers between tanks can be identified in these data, segments of which can then be compared to generate transfer differences (TDs). A safeguards concern might then be raised if either these TDs or deviations in M or V data during “wait” modes become significant. Average M and V TDs should be 0 (perhaps following a bias adjustment) to within a historical limit that is a multiple of the standard deviation of the M or V TD, as should deviations during “wait” modes. Statistical test options can be compared on the basis of their estimated probabilities to detect various material loss scenarios. Multivariate statistical PM options have previously been applied to residuals produced from simulated SM data that had no process variation, only random and systematic measurement errors. This article examines how detection probabilities might be estimated with real data. To do this, realistic effects such as pump carryover, evaporation, condensation, and mixing/sparging, are included in simulated data. In real facilities false alarms are a concern, particularly when data need to be evaluated regularly, on a day-to-day basis. The need to widen control limits to avoid alarming on innocent process variation effects is discussed, and the consequential reduction in DPs is illustrated with numerical examples.

Basic Analysis on Isotopic Barrier of Material Attractiveness Based on Plutonium Composition of FBR

Fuel behaviors of the large fast breeder reactor have been investigated, as well as material attractiveness based on isotopic plutonium composition for evaluating proliferation resistance with regards to a combined evaluation of decay heat and spontaneous fission neutron barrier as key parameters of isotopic material barrier. Trans-uranium fuel (TRU) (MA + U-Pu) in the core regions and MA doping (MA + natural U) in the blanket regions as options of MA loading produce a higher Pu-238 composition for denaturing plutonium, which mainly comes from converted Np-237. The isotopic plutonium composition of TRU fuel is relatively less than the Pu composition of MOX fuel except for the Pu-238 composition that is higher than that of MOX fuel. MA in the core or blanket regions, which produces a higher Pu-238 composition, plays a key role in obtaining a high-level material barrier of decay heat and spontaneous fission neutron compositions. The material attractiveness level of plutonium composition in the core regions can be categorized as practically unusable and its level becomes less by adopting TRU fuel. In addition, the material attractiveness level in the blanket regions as being practically unusable can be reached from weapon grade by loading MA at a 2% doping rate.

Numerical Consideration for Multiscale Statistical Process Control Method Applied to Nuclear Material Accountancy

The multiscale statistical process control (MSSPC) method is applied to clarify the elements of material unaccounted for (MUF) in large scale reprocessing plants using numerical calculations. Continuous wavelet functions are used to decompose the process data, which simulate batch operation superimposed by various types of disturbance, and the disturbance components included in the data are divided into time and frequency spaces. The diagnosis of MSSPC is applied to distinguish abnormal events from the process data and shows how to detect abrupt and protracted diversions using principle component analysis. Quantitative performance of MSSPC for the time series data is shown with average run lengths given by Monte-Carlo simulation to compare to the non-detection probability β. Recent discussion about bias corrections in material balances is introduced and another approach is presented to evaluate MUF without assuming the measurement error model.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to ...

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by convertion rasio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and its produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present...

Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of thereactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to runthe analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor typeas a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day)...

Solution Monitoring Evaluated by Proliferation Risk Assessment and Fuzzy Optimization Analysis for Safeguards in a Reprocessing Process

Solution monitoring (SM) has been used in a nuclear reprocessing plant as an additional measure to provide assurance that the plant is operated as declared. The inline volume and density monitoring equipment with dip tubes is important for safety and safeguards purposes and is a typical example of safeguards by design (SBD). Recently safety, safeguards, and security by design (3SBD) are proposed to promote an efficient and effective generation of nuclear energy. In 3SBD, proliferation risk assessment has the potential to consider likelihood of the incidence and proliferation risk in safeguards. In this study, risk assessment methodologies for safeguards and security are discussed and several mathematical methods are presented to investigate risk notion applied to intentional acts of facility misuse in an uncertainty environment. Proliferation risk analysis with the Markov model, deterrence effect with the game model, and SBD with fuzzy optimization are shown in feasibility studies to investigate the potential application of the risk and uncertainty analyses in safeguards. It is demonstrated that the SM is an effective measurement system using risk-informed and cost-effective SBD, even though there are inherent difficulties related to the possibility of operator’s falsification.

Analysis on Even Mass Plutonium Production of Different Loading Materials in FBR Blanket

Spent nuclear fuel (SNF) from nuclear facilities such as from accumulated SNF commercial reactors becomes one of the important issues in term of reducing environmental impact and fuel sustainability as well as nuclear nonproliferation point of view when those SNF materials can be recycled and utilized as new fuel loaded into the reactors. Minor actinides (MA) as one of the important material of spent nuclear fuels can be recycled and transmuted into some useful materials which can be utilized to increase the fuel breeding capability as well as for increasing protected plutonium production from the view point of nuclear nonproliferation issue. Increasing some even mass isotopic plutonium compositions are estimated to increase the level of proliferation resistance level in term of material barrier point of view. The objective of this study is to analyze the proliferation resistance aspect of nuclear fuel based on plutonium production of different loading materials in the FBR blanket. Evaluation is based on some basic parameters of reactor operation analysis, such as reactor operation time which is adjusted to 800 days operation per cycle for 4 fuel batches systems which is refered to the large FBR type of Japan Sodium Fast Reactor (JSFR) design. The results show some nuclear fuels behavior during reactor operation for different loading materials and cycles. Minor actinide (MA) material loading as doping material gives some significant plutonium productions during reactor operation. Some obtained actinide productions have different profiles such as some reducing compositions in americium and neptunium actinide compositions with the time which depends on initial loading material. Some plutonium vector compositions are evaluated from Pu-238 to Pu-242 to estimate the proliferation resistance level as isotopic material barrier of plutonium. Some significant contributions for increasing even mass plutonium as plutonium protected material are shown by Pu-238 from all doping material as well as additional production of Pu-240 and Pu-242 in certain conditions.

Irradiation and Cooling Process Effects on Material Barrier Analysis Based on Plutonium Composition of LWR

Nuclear fuel utilization program from front-end up to back-end processes especially spent fuel management have been monitored and safeguarded by the IAEA in order to ensure the utilization of nuclear fuels from all nuclear facilities including nuclear fuel reprocessing facilities are dedicated only for civil and peaceful purposes. Nuclear fuel production processes including reactor criticality condition is one of the major topics in term of nuclear fuel sustainability which related to energy security issues. Meanwhile, reduction level or preventing processes of nuclear fuel utilization from its potential risk from nuclear explosive purposes should be also strengthened and prioritized. To increase the intrinsic proliferation resistance of nuclear fuel, one of the potential ways is by increasing the material barrier level such as isotopic barrier. In case of plutonium, increasing the intrinsic properties of plutonium isotopes can be used by increasing material barrier of even mass number (Pu-238, Pu-240 and Pu-242). In this study, the effect of different irradiation process during reactor operation which related to discharged fuel burnup have been used and decay time to analyzed its dependeny to plutonium production as well as plutonium production dependency to decay or cooling time processes. Fuel production analysis of the reactor are based on the spent fuel of light water reactor (LWR) with different discharged fuel burnup (33 GWd/t, 50 GWd/t and 60 GWd/t) and different decay or cooling time process (1 to 30 years cooling time). Fuel behavior optimization of LWR design are obtained by using ORIGEN code by employing some modules for analyzing fuel production dependencies to burnup and decay time processes. In this study, two parameters for investigating the material barriers are adopted such as decay heat (DH) and spontaneous fission neutron (SFN) compositions. The compositions of DH and SFN are sensitive to the composition of isotopic plutonium especially more sensitive to even mass plutonium composition. Higher discharged fuel burnup level produces more even mass plutonium compositions and effectively reduce Pu-239 production because of more fissile Pu-239 are consumed for higher burnup. Isotopic Pu-238 gives the highest DH contributor, while Isotope Pu-240 obtains the highest contribution of SFN followed by other plutonium isotopes. DH and SFN compositions of plutonium can be increased effectively by increasing burnup process. Longer decay time is also effective to increase SFN compositions because of its dependency to all even mass plutonium while it gives less DH compositions because of its dependency to the contribution of Pu-238.