Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 22

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

CIELO collaboration summary results; International evaluations of neutron reactions on uranium, plutonium, iron, oxygen and hydrogen

Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.

Nuclear Data Sheets, 148, p.189 - 213, 2018/02

 Times Cited Count:20 Percentile:3.46(Physics, Nuclear)

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - $$^{235}$$U, $$^{238}$$U, $$^{239}$$Pu, $$^{56}$$Fe, $$^{16}$$O and $$^{1}$$H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.

Journal Articles

EBR-II passive safety demonstration tests benchmark analyses; Phase 2

Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08

The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.

Journal Articles

IAEA NAPRO Coordinated Research Project; Physical properties of sodium

Passerini, S.*; Carardi, C.*; Grandy, C.*; Azpitarte, O. E.*; Chocron, M.*; Japas, M. L.*; Bubelis, E.*; Perez-Martin, S.*; Jayaraj, S.*; Roelofs, F.*; et al.

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.780 - 790, 2015/05

Journal Articles

IAEA benchmark calculations on control rod withdrawal test performed during Phenix End-of-Life experiments; JAEA's calculation results

Takano, Kazuya; Mori, Tetsuya; Kishimoto, Yasufumi; Hazama, Taira

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09

This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region.

Journal Articles

IAEA benchmark calculations on control rod withdrawal test performed during Phenix End-of-Life experiments; Benchmark results and comparisons

Pascal, V.*; Prulhi$`e$re, G.*; Vanier, M.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Chenu, A.*; et al.

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09

no abstracts in English

Journal Articles

Journal Articles

A New IAEA coordinated research project on sodium properties and safe operation of experimental facilities in support of the development and deployment of sodium-cooled fast reactors

Monti, S.*; Latge, C.*; Long, B.*; Azpitarte, O. E.*; Chellapandi, P.*; Stieglitz, R.*; Eckert, S.*; Ohira, Hiroaki; Lee, J.*; Roelofs, F.*; et al.

Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.474 - 481, 2014/04

Journal Articles

Benchmark calculations on control rod withdrawal tests performed during Phenix End-of-Life experiments

Pascal, V.*; Prulhi$`e$re, G.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Semenov, M.*; Taiwo, T.*; et al.

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 11 Pages, 2013/04

The control rod withdrawal test was one of the various Phenix End-of-Life tests performed in 2009. The main goal was to determine the impact of a rod insertion and/or extraction on the radial power distribution in the fissile core at nominal power. The framework of the Technical Working Group on Fast Reactors (TWG-FR) activities in IAEA, decided to launch a Coordinated Research Project (CRP), devoted to benchmarking analyses on the test. The CRP was performed by experts coming from CEA, ANL, IGCAR, IPPE, IRSN, JAEA, KIT and PSI. After a short description of the test conducted in the Phenix reactor, this paper presents some results obtained in the course of the CRP with special emphasis on control rod efficiencies and power deformation by subassemblies. The paper also discusses the discrepancies found when comparing calculated results with experimental data as well as some preliminary conclusions on the source of these discrepancies.

Journal Articles

Implementation of dynamic cross-talk correction (DCTC) for MOX holdup assay measurements among multiple gloveboxes

Nakamura, Hironobu; Beddingfield, D.*; Montoya, J.*; Nakamichi, Hideo; Mukai, Yasunobu; Kurita, Tsutomu

Proceedings of INMM 53rd Annual Meeting (CD-ROM), 9 Pages, 2012/07

Plutonium holdup inventory in gloveboxes are measured by HBAS for the nuclear material accountancy (NMA) at PCDF. Because the gloveboxes are installed close to one another, we must make a correction for neutron cross-talk between the gloveboxes. In order to address the issue of variable cross-talk contributions to holdup assay values, we developed a dynamic cross-talk correction (DCTC) method to obtain the actual doubles signal cross-talk between multiple gloveboxes. With the HBAS improvement, the DCTC improves PCDF NMA by eliminating the double-counting of material that stems from cross-talk in the holdup assay data and eliminates this source of bias in the assay results. Since the DCTC methodology can be used to determine the cross-correlation among multiple inventories in small areas and substantially reduce cross-talk-induced biases in assay results, it is expected that DCTC technology can reflect as a safeguards-by-design.

Journal Articles

Deposition of boron on fuel rod surface under sub-cooled boiling conditions; An Approach toward understanding AOA occurrence

Uchida, Shunsuke; Asakura, Yamato*; Suzuki, Hiroaki*

Nuclear Engineering and Design, 241(7), p.2398 - 2410, 2011/07

 Times Cited Count:22 Percentile:13.55(Nuclear Science & Technology)

In PWR, it has been assumed that Li and B ions deposited on fuel under sub-cooled boiling conditions. Accumulated B on the fuel led to AOA. The amount of B deposited on the fuel was evaluated from two directions. The first calculated the amount with the extended MED model and the other estimated it from the viewpoint of reactor reactivity. It was concluded that: (1) the calculated B deposition amount on the fuel, which was one or two orders larger than measured amounts and Ni oxides compounds, was seldom measured in the fuel deposits due to its high release rate; (2) its hideout return during the reactor shutdown period was seldom observed due to its high concentration in the primary coolant; (3) one of the most promising approaches to evaluate its accumulation on the fuel during plant operation was the MED model calculation; and (4) control of Ni concentration in the primary coolant resulted in decreased Ni oxide deposition and then mitigation of AOA occurrence.

Journal Articles

Validity of the master curve temperature dependence assumption for highly embrittled RPV materials; Results from the IAEA Coordinated Research Project (CRP-8)

Planman, T.*; Onizawa, Kunio; Server, W.*

Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 8 Pages, 2009/07

The fracture toughness transition curve shape in the Master Curve (MC) has been discussed since the original empirical definition of the curve in 1991. In most cases the standard MC approach, assuming a constant transition curve shape, has proven to give a realistic description for also highly irradiated ferritic steels. The fracture toughness data collected and analysed in the IAEA CRP-8 Topic Area 3 supports the validity of the curve shape assumption of ASTM E1921 also in case of irradiated steels and gives no rise to change the present definition. The Master Curve C-parameter (the shape parameter) estimation is proposed as an appropriate analysis method when there is need to estimate also the temperature dependence, whereas the SINTAP procedure is recommended for ensuring conservative lower bound estimates when material in homogeneity is suspected. The results show that irradiation may slightly lower the fracture toughness in the upper transition region in relation to that predicted by E1921, but the effect after moderate T0 shift values seems to be negligible.

Journal Articles

IAEA coordinated research project on master curve approach to monitor fracture toughness of RPV steels; Final results of the experimental exercise to support constraint effects

Nanstad, R.*; Brumovsky, M.*; Callejas, R.*; Gillemot, F.*; Korshunov, M.*; Lee, B.*; Lucon, E.*; Scibetta, M.*; Minnebo, P.*; Nilsson, K.-F.*; et al.

Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 13 Pages, 2009/07

IAEA has developed a coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water RPVs. Topic Area 1 of the CRP is focused on the issue of test specimen geometry effects, with emphasis on determination of reference temperature T$$_{0}$$ with the pre-cracked Charpy (PCC) specimen and the bias effect on T$$_{0}$$. Participating organizations for the experimental part of the CRP performed fracture toughness testing of various steels with various types of specimens under various conditions. Results from fracture toughness tests are compared with regard to effects of specimen size and type on the T$$_{0}$$. It is apparent from the results that the bias observed between the PCC specimen and larger specimens for Plate JRQ is not nearly as large as that obtained for other steels (-11$$^{circ}$$C to -45$$^{circ}$$C). This observation is consistent with observations in the literature that show significant variations in the bias that are dependent on the specific materials being tested.

Journal Articles

Final results of an analytical round robin exercise to support constraint effects

Scibetta, M.*; Altstadt, E.*; Callejas, R.*; Lee, B.*; Miura, Naoki*; Onizawa, Kunio; Paffumi, E.*; Serrano, M.*; Tatar, L.*; Yin, S.*

Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 11 Pages, 2009/07

IAEA has developed a coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water RPVs. Topic Area 1 of the CRP is focused on the issue of test specimen geometry effects, with emphasis on determination of reference temperature T0 with the pre-cracked Charpy specimen and the bias effect. Within the analytical part, elastic plastic finite element methods are used in order to access local stress and strain information. This analytical round robin exercise has been performed by ten laboratories from nine different countries focusing on the modeling of realistic three dimensional geometries containing shallow and deep crack. Independently of the used code and of relatively small user effect differences, it is found that shallow crack specimens are more sensitive to loss of constraint than deep crack specimens for a given specimen size. The difference in terms of reference temperature between the two geometries is evaluated to be about 40$$^{circ}$$C.

JAEA Reports

Data description for coordinated research project on benchmark analyses of sodium natural convection in the upper plenum of the Monju reactor vessel under supervisory of Technical Working Group on Fast Reactors, International Atomic Energy Agency

Yoshikawa, Shinji; Minami, Masaki*

JAEA-Data/Code 2008-024, 28 Pages, 2009/01

JAEA-Data-Code-2008-024.pdf:5.83MB

A series of information required for numerical simulation of sodium thermal stratification observed at the plant trip test of "Monju" conducted in 1995 is provided, which consists of the test outline, geometry data of the reactor vessel upper plenum between the reactor core top and reactor outlet nozzles, and flow inlet boundary conditions at the reactor core top surface.

Journal Articles

Microscopic studies of spherical particles for nuclear safeguards

Donohue, D.*; Ciurapinski, A.*; Cliff, J.*; R$"u$denauer, F.*; Kuno, Takehiko; Poths, J.*

Applied Surface Science, 255(5, Part2), p.2561 - 2568, 2008/12

 Times Cited Count:13 Percentile:45.63(Chemistry, Physical)

A combination of micro-analytical techniques was used for the characterization of spherical particles in the size range 9-12$$mu$$m as a part of nuclear safeguards verification activities pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). The particles were removed from cotton swipe samples taken in a nuclear facility under safeguards and examined first by scanning electron microscopy combined with energy-dispersive X-ray spectrometry (SEM-EDX). Particles of interest were then relocated under an optical microscope and manipulated. One such particle was subjected to destructive analysis by secondary ion mass spectrometry (SIMS) in order to determine if uranium was present in the core of the particle. A second particle was examined using focused ion beam (FIB) etching to allow an examination of the interior by SEM-EDX. The particle manipulation and relocation techniques presented here allow the sequential examination of a single particle of interest by a combination of analytical techniques, thus yielding surface morphological, elemental, isotopic and depth-profiling information. The objective of these investigations is to provide assurance of the absence of clandestine or undeclared nuclear activities in States coming under comprehensive safeguards obligations.

Journal Articles

Improvement of analytical activities in the Tokai reprocessing plant, Japan, by measuring destructive and non-destructive assays

Surugaya, Naoki; Taguchi, Shigeo; Kurosawa, Akira; Watahiki, Masaru

STI/PUB/1298 (CD-ROM), p.673 - 679, 2007/08

We have been analyzing nuclear materials at the Tokai pilot reprocessing plant, Japan, since 1977. To obtain reliable measurements for nuclear material such as uranium and plutonium, we have developed various kinds of measurement techniques and implemented effective ones for accountability and verification analyses in a nuclear material accountancy system. One of our role as a pilot plant has been successfully accomplished with the effort put into various analytical activities. Now, it is time to transfer the experience gained with our technology to the next large-scale commercial plant in Rokkasho. This paper presents our analytical methods and their results obtained using analytical techniques we have applied over recent years.

Journal Articles

IAEA coordinated research project on master curve approach to monitor fracture toughness of RPV steels; Applicability for highly embrittled materials

Planman, T.*; Onizawa, Kunio; Server, W.*; Rosinski, S.*

Proceedings of 2007 ASME Pressure Vessels and Piping Division Conference/8th International Conference on Creep and Fatigue at Elevated Temperatures (PVP 2007/CREEP-8) (CD-ROM), 9 Pages, 2007/07

In the Master Curve (MC) fracture model, a universal temperature dependence is assumed for reactor pressure vessel (RPV) steels. The assumed curve shape has been observed to be generally valid for highly irradiated materials. Lower than predicted fracture toughness behavior has been occasionally observed, however, in the upper transition range. One objective of the present IAEA CRP is to clarify the MC shape issue by collecting and analyzing relevant fracture toughness data on irradiated or thermally aged RPV steels. The data reviewed in this CRP show, in general, a very consistent fracture behavior with the basic MC model that further confirms the applicability of the assumed curve shape. In cases where the basic assumptions of the MC model were not satisfied due to high proportions of intergranular fracture, correspondence with the measured and predicted behavior could be markedly improved by applying available models developed to address inhomogeneous materials.

JAEA Reports

Confirmation of the availability of an analytical technique, Pu(VI) spectrophotometry for HALW; Technical Support for the Joint IAEA/Japan On-site Analytical Laboratory at the Rokkasho Reprocessing Plant (JASPAS JU-01-01)

Kitao, Takahiko; Sato, Soichi; Kuno, Takehiko; Keiji, Yamada,; Watahiki, Masaru; Kamata, Masayuki

JNC-TN8410 2003-014, 29 Pages, 2003/11

JNC-TN8410-2003-014.pdf:1.82MB

The Agency requested the Tokai Reprocessing Plant(TRP) to confirm the applicability of three kinds of analytical procedure for Pu(VI) spectrophotometry in On-Site Analytical Laboratory (OSL) at Rokkasho, in order to obtain accurate plutonium(Pu) concentration in High Active Liquid Waste (HALW). Three analytical procedures, (1) Calibration method, (2) Nd internal standard method and (3) Reduction method, were tested. The measurement sample was prepared by adding the known amount of plutonium in the actual HALW after removing original Pu by solid extraction. We measured the Pu concentration in the sample by three methods and calculated the accuracy and precision. The results of each method are summarized as follows:(1) Calibration method Plutonium concentration calculated by the calibration method agreed with that by adjusted concentration. (2) Nd internal standard method Accurate results were obtained by this method. The error of pretreatment, especially dilution, has not influenced on the Pu measurements. (3) Reduction method The measured Pu concentrations were higher than those by adjusted. From the comparison with these results, the calibration method is the most simple and rapid in the three methods. Analysis time was within 1 hour including sample preparation. The detection limit, with the calibration method, was 1.3 mgPu/L in the actual HALW measurements.

Oral presentation

Activities in proliferation resistance area of INPRO

Aso, Ryoji

no journal, , 

no abstracts in English

Oral presentation

Data description for the second Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on "Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel"

Yoshikawa, Shinji; Minami, Masaki*

no journal, , 

This document provides a set of technical data for the thermal hydraulic analysis of liquid sodium in the upper plenum of Monju reactor vessel, consisting of the additional information which JAEA stated to be offered at first RCM held in September 2008 in Vienna, and the table of temperature versus time measured by the vertical array of thermocouples inserted in the reactor upper plenum during the turbine trip test of the system start-up tests conducted in December 1995.

22 (Records 1-20 displayed on this page)