Evaluation of the in-situ performance of neutron detectors based on EJ-426 scintillator screens for spent fuel characterization

In the framework of the MICADO (Measurement and Instrumentation for Cleaning And Decommissioning Operations) European Union (EU) project, aimed at the full digitization of low- and intermediate-level radioactive waste management, a set of 32 solid state thermal neutron detectors named SiLiF has been built and characterized. MICADO encompasses a complete active and passive characterization of the radwaste drums with neutrons and gamma rays, followed by a longer-term monitoring phase. The SiLiF detectors are suitable for the monitoring of nuclear materials and can be used around radioactive waste drums possibly containing small quantities of actinides, as well as around spent fuel casks in interim storage or during transportation. Suitable polyethylene moderators can be exploited to better shape the detector response to the expected neutron spectrum, according to Monte Carlo simulations that were performed. These detectors were extensively tested with an AmBe neutron source, and the results show a quite uniform and reproducible behavior.

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In-situ calibration of TFTR (Tokamak Fusion Test Reactor) neutron detectors

10.2172/7164819 ◽

1990 ◽

Cited By ~ 1

Author(s):

H.W. Hendel ◽

R.W. Palladino ◽

C.W. Barnes ◽

M. Diesso ◽

J.S. Felt ◽

...

Keyword(s):

Reactor Neutron ◽

Neutron Detectors ◽

In Situ Calibration ◽

Test Reactor

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Strengthening the SSTC NRS Scientific and Technical Potential through Participation in the IAEA Coordinated Research Projects

Nuclear and Radiation Safety ◽

10.32918/nrs.2021.1(89).01 ◽

2021 ◽

pp. 5-13

Author(s):

Yu. Balashevska ◽

D. Gumenyuk ◽

Iu. Ovdiienko ◽

O. Pecherytsia ◽

I. Shevchenko ◽

...

Keyword(s):

Best Practices ◽

Safety Assessment ◽

Radiation Safety ◽

Research Projects ◽

Spent Fuel ◽

Research Activities ◽

Technical Potential ◽

Technical Issues ◽

Fuel Characterization ◽

Definition Of

The State Scientific and Technical Center for Nuclear and Radiation Safety (SSTC NRS), a Ukrainian enterprise with a 29-year experience in the area of scientific and technical support to the national nuclear regulator (SNRIU), has been actively involved in international research activities. Participation in the IAEA coordinated research activities is among the SSTC NRS priorities. In the period of 2018–2020, the IAEA accepted four SSTC NRS proposals for participation in respective Coordinated Research Projects (CRPs). These CRPs address scientific and technical issues in different areas such as: 1) performance of probabilistic safety assessment for multi-unit/multi-reactor sites; 2) use of dose projection tools to ensure preparedness and response to nuclear and radiological emergencies; 3) phenomena related to in-vessel melt retention; 4) spent fuel characterization. This article presents a brief overview of the abovementioned projects with definition of scientific contributions by the SSTC NRS (participation in benchmarks, development of methodological documents on implementing research stages and of IAEA technical documents (TECDOC) for demonstration of best practices and results of research carried out by international teams).

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Applications of Nuclear Data Covariances to Criticality Safety and Spent Fuel Characterization

Nuclear Data Sheets ◽

10.1016/j.nds.2014.04.075 ◽

2014 ◽

Vol 118 ◽

pp. 341-345 ◽

Cited By ~ 5

Author(s):

M.L. Williams ◽

G. Ilas ◽

W.J. Marshall ◽

B.T. Rearden

Keyword(s):

Nuclear Data ◽

Spent Fuel ◽

Fuel Characterization ◽

Criticality Safety

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In situ calibration of TFTR neutron detectors

Review of Scientific Instruments ◽

10.1063/1.1141115 ◽

1990 ◽

Vol 61 (7) ◽

pp. 1900-1914 ◽

Cited By ~ 50

Author(s):

H. W. Hendel ◽

R. W. Palladino ◽

Cris W. Barnes ◽

M. Diesso ◽

J. S. Felt ◽

...

Keyword(s):

Neutron Detectors ◽

In Situ Calibration

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REPRO: Through Diffusion Experiment (TDE) - Diffusion and Porosity Properties of Rock Matrix in Stress Field of Repository Level

10.5194/egusphere-egu2020-74 ◽

2020 ◽

Author(s):

Kalle Rahkola ◽

Antti Poteri ◽

Lasse Koskinen ◽

Peter Andersson ◽

Kersti Nilsson ◽

...

Keyword(s):

Stress Field ◽

Performance Assessment ◽

Spent Nuclear Fuel ◽

Diffusion Experiment ◽

Spent Fuel ◽

Tracer Transport ◽

Rock Matrix ◽

Safety Case ◽

Through Diffusion

Radionuclides usually migrate slower than the flowing water due to sorption and matrix diffusion. The performance assessment assumes that retention takes place mostly in the vicinity of the deposition holes. REPRO (REtention Properties of ROck matrix) experiments analyzed the matrix retention properties of the rock matrix under realistic conditions deep in the bedrock in ONKALO underground characterization facility at Olkiluoto, Finland. The objective was to investigate tracer transport in the rock matrix, which was representative to the near-field of the final disposal repository of the spent nuclear fuel, and to demonstrate that the assumptions made in the safety case of the deep geological spent fuel repository were in line with site evidence.REPRO is composed of several supporting laboratory and in-situ experiments which investigate the retention properties under different experimental configurations. The first in-situ experiments were water phase diffusion experiments performed 2012-2013. Through Diffusion Experiment (TDE) studies diffusion and porosity properties of rock matrix in stress field of repository level and sorption properties of nuclides in intact rock circumstances.The TDE experiment has been performed in three parallel drillholes drilled near to each other. Breakthrough of the radioactive tracer is monitored with on-line measurements and samplings along and perpendicular to the foliation. The non-sorbing radioactive isotope traces of HTO and 36Cl, as well as slightly sorbing 22Na and strongly sorbing 133Ba and 134Cs were used. TDE was designed to control advective flow, as it had caused problems in previous in-situ tests.Supporting laboratory studies were performed for drillcore samples sampled from the experimental drillholes. In these laboratory experiments, i.e. porosity, permeability and diffusion coefficients of the drillcores were determined using different methods.The TDE experiment was carried out from 2016 to 2019. A breakthrough was seen in the timeframe predicted by scoping calculations carried out. REPRO has produced data and knowledge to the safety case and the performance assessment. According to the preliminary results, values measured in the laboratory are applicable also in larger scale and in-situ conditions.

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Modelling coupled processes in bentonite: recent results from the UK's contribution to the Äspö EBS Task Force

Mineralogical Magazine ◽

10.1180/minmag.2012.076.8.19 ◽

2012 ◽

Vol 76 (8) ◽

pp. 3033-3043 ◽

Cited By ~ 8

Author(s):

D. Holton ◽

S. Baxter ◽

A. R. Hoch

Keyword(s):

Large Scale ◽

Task Force ◽

Fractured Rock ◽

Coupled Processes ◽

Spent Fuel ◽

Rock Interaction ◽

Waste Container ◽

Buffer Material ◽

The Uk

AbstractA range of potential concepts for the geological disposal of high level wastes and spent fuel are being studied and considered in the UK. These include concepts that use bentonite as a buffer material around the waste containers. The bentonite will be required to fulfil certain safety functions, the most important being (1) to protect the waste containers from detrimental thermal, hydraulic, mechanical and chemical processes; and (2) to retard the release of radionuclides from any waste container that fails. The bentonite should have a low permeability and a high sorption capacity.These safety functions could be challenged by certain features, events and processes (FEPs) that may occur during the evolution of the disposal system. A consideration of how these FEPs may affect the safety functions can be used to identify and to prioritize the important areas for research on bentonite. We identify these important areas (which include hydration of compacted bentonite, illitization and erosion of bentonite), and describe how they are being investigated in current international research on bentonite.The Äspö EBS Task Force is a collaborative international project designed to carry out research on bentonite. In 2011, the Nuclear Decommissioning Authority Radioactive Waste Management Directorate joined the EBS Task Force partly to benefit from its collective experience. The work of the EBS Task Force is split into two research subareas: (1) the THM subarea, which includes tasks to understand homogenization of bentonite as it resaturates, to investigate the hydraulic interaction between bentonite and fractured rock, and to model in situ experiments; and (2) the THC subarea, which includes tasks to investigate the issue of understanding transport through bentonite, and to model in situ experiments. In particular, the bentonite rock interaction experiment is a large-scale in situ experiment concerned with understanding groundwater exchange across bentonite rock interfaces, with the objective of establishing better understanding of bentonite wetting. In this paper, we describe our work to model the spatial and temporal resaturation of bentonite buffer in a fractured host rock.

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Modelling Oxidative Dissolution of Spent Fuel

MRS Proceedings ◽

10.1557/proc-465-573 ◽

1996 ◽

Vol 465 ◽

Cited By ~ 6

Author(s):

Ivars Neretnieks

Keyword(s):

Redox Reaction ◽

Uranium Oxide ◽

Reaction Rate ◽

Spent Nuclear Fuel ◽

Large Fraction ◽

Reaction Rates ◽

Spent Fuel ◽

First Case ◽

Rate Of Movement

ABSTRACTSpent nuclear fuel will, by the radiation, split nearby water into oxidizing and reducing compounds. The reducing compounds are mostly hydrogen that will diffuse away. The remaining oxidizing compounds can oxidize the uranium oxide of the fuel and make it more soluble. The oxidised uranium will dissolve and diffuse away. The nuclides previously incorporated in the spent fuel matrix can then be released and also migrate away from the fuel.A model is proposed where the produced oxidizing species compete for reaction with the fuel and for escaping out of the system. The chemical reaction rate of oxygen and fuel is taken from literature values based on experiments. The escape rate of oxidants to a receding redox front in the backfill is modelled assuming a redox reaction of oxidizing component and reducing component in the surrounding. The rate of movement of the redox front is determined from the rate of production of oxidants. This is estimated using a previously devised model that has been calibrated to in situ observed radiolysis.Three cases are modelled. In the first case it is assumed that the reducing compound is insoluble and that the reaction between oxygen and reducing mineral is very fast. In the second case it is assumed that the reducing component has a known solubility and that it can migrate to meet the oxygen and quickly react. In a third case a finite reaction rate is modelled between the oxygen and the reducing species.The sample calculations show that if the reducing mineral has to be supplied from the backfill a large fraction of the spent fuel could be oxidised. If the corrosion products of a degraded steel canister can supply the reducing species and the redox reaction is fast, very small amounts of the fuel could be oxidised. Literature data indicate that the redox reaction rate may not be so fast that it can be considered instantaneous and then a considerable fraction of the fuel could be oxidised. The model gives a means of exploring which mechanisms and data may be of most importance for radiolytic fuel dissolution, but the realism of the data and the model must be tested further. There is a lack of understanding and data on reaction rates, heterogeneous as well as homogeneous. This is crucial to the results.

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HEDL contribution to ONWI, September monthly report: spent fuel characterization

10.2172/6155318 ◽

1980 ◽

Author(s):

R.J. Cash

Keyword(s):

Spent Fuel ◽

Monthly Report ◽

Fuel Characterization

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BOUNDARY CONDITION MODELING EFFECT ON THE SPENT FUEL CHARACTERIZATION AND FINAL DECAY HEAT PREDICTION FROM A PWR ASSEMBLY

EPJ Web of Conferences ◽

10.1051/epjconf/202124712008 ◽

2021 ◽

Vol 247 ◽

pp. 12008

Author(s):

Augusto Hernandez-Solis ◽

Klemen Ambrožič ◽

Dušan Čalič ◽

Luca Fiorito ◽

Bor Kos ◽

...

Keyword(s):

Boundary Condition ◽

Fuel Assembly ◽

Emission Rate ◽

Neutron Emission ◽

Spent Fuel ◽

Decay Heat ◽

Fuel Pin ◽

Condition Modeling ◽

Fuel Characterization ◽

Relative Differences

In this paper, two main exercises have been carried out to describe the effect that varying an albedo boundary condition has in the computation of observables such as decay heat, neutron emission rate and nuclide inventory from a PWR fuel assembly (or a configuration of assemblies) during a depletion scenario. The SERPENT2 code was then employed to emphasize the importance of modeling a proper boundary condition for such purposes. Moreover, the effect of taking into account more than a single fuel-pin region for depletion studies while varying the type of boundary condition, was also accounted for. The first exercise has the main objective of comparing in a single fuel assembly the albedo variations ranging from 1.1 up to full vacuum conditions. By comparing to the reference assembly (considered to be the case of full reflective conditions), relative differences up to +17% were observed in decay heat and up to almost -30% in neutron emissions. Also, a clear dependence on the albedo was detected if more than one depletable zone was considered while computing the integral value of observables of interest. Regarding the second exercise, where a 3 × 3 configuration of fuel assemblies is being now considered with a reflector section in the middle, a negligible effect on the observables was observed for the single fuel pin zone case; instead, an effect in the 244Cm computation when analyzing two fuel pin-zones produced a change in the neutron emission rate during cooling time up to 2.5% (while comparing it to the reference single assembly case).

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