Modelling and simulation of the source term for a sodium cooled fast reactor under hypothetical severe accident conditions; Final report of a coordinated research project

Arokiaswamy, J. A.*; Batra, C.*; Chang, J. E.*; Garcia, M.*; Herranz, L. E.*; Klimonov, I. A.*; Kriventsev, V.*; Li, S.*; Liegeard, C.*; Mahanes, J.*; et al.

IAEA-TECDOC-2006, 380 Pages, 2022/00

https://www.iaea.org/publications/15086/modelling-and-simulation-of-the-source-term-for-a-sodium-cooled-fast-reactor-under-hypothetical-severe-accident-conditions

The IAEA coordinated research project on "Radioactive Release from the Prototype Sodium Cooled Fast Reactor under Severe Accident Conditions" was devoted to realistic numerical simulation of fission products and fuel particles inventory inside the reference sodium cooled fast reactor volumes under severe accident conditions at different time scales. The scope of analysis was divided into three parts, defined as three work packages (WPs): (1) in-vessel source term estimation; (2) primary system/containment system interface source term estimation; and, (3) in-containment phenomenology analysis. Comparison of the results obtained in WP-1 indicates that the release fractions of noble gases and cesium radionuclides, and fractions of radionuclides released to the cover gas are in a good agreement. In the analysis using a common pressure history in WP-2, the results were in good agreement indicating that the accuracy of the analysis method of each institution is almost the same. The standalone case, which uses a set of pre-defined release fractions, was defined for WP-3 which enables to decouple this part of analysis from previous WPs. There is broad consensus among the predicted results by all the participants in WP-3.

Development of a reference database for beta-delayed neutron emission

Dimitriou, P.*; Dillmann, I.*; Singh, B.*; Piksaikin, V.*; Rykaczewski, K. P.*; Tain, J. L.*; Algora, A.*; Banerjee, K.*; Borzov, I. N.*; Cano-Ott, D.*; et al.

Nuclear Data Sheets, 173, p.144 - 238, 2021/03

https://doi.org/10.1016/j.nds.2021.04.006

-delayed neutron emission has been of interest since the discovery of nuclear fission. In nuclear power reactors, delayed-neutron data play a crucial role in reactor kinetics calculations and safe operation. -delayed neutron data also have a significant impact in the field of nuclear structure and astrophysics especially as nuclei farther away from stability are explored at the new generation of radioactive beam facilities. Several compilations of -decay half-lives and delayed-neutron emission probabilities are available, however, complete documentation of measurements and evaluation procedures is often missing for these properties. Efforts to address this gap in nuclear data and create an updated compilation and evaluation of -delayed neutron properties were undertaken under the auspices of the International Atomic Energy Agency (IAEA) which formed a Coordinated Research Project (CRP) on "Development of a Reference Database of Beta-delayed Neutron Emission Data". In this paper we summarize the work that was performed and present the results of the CRP.

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.

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.

Benchmark analyses on the control rod withdrawal tests performed during the PHNIX end-of-life experiments

Monti, S.*; Toti, A.*; Stanculescu, A.*; Pascal, V.*; Fontaine, B.*; Herrenschmidt, A.*; Prulhiere, G.*; Vanier, M.*; Varaine, F.*; Vasile, A.*; et al.

IAEA-TECDOC-1742, 247 Pages, 2014/06

http://www-pub.iaea.org/books/IAEABooks/10745/Benchmark-Analyses-on-the-Control-Rod-Withdrawal-Tests-Performed-during-the-PHNIX-End-of-Life-Experiments

http://www-pub.iaea.org/MTCD/Publications/PDF/TE-1742_web.pdf

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

Pascal, V.*; Prulhire, 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.

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

https://doi.org/10.1016/j.nucengdes.2011.04.009

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.

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 with the pre-cracked Charpy (PCC) specimen and the bias effect on T. 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. 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 (-11C to -45C). 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.

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.

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.