Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

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Study on Ultra-Long Life, Small U-ZR Metallic Fuelled Core With Burnable Poison

10th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone10-22355 ◽

2002 ◽

Author(s):

Kenji Tsuji ◽

Hiromitsu Inagaki ◽

Akira Nishikawa ◽

Hisato Matsumiya ◽

Yoshiaki Sakashita ◽

...

Keyword(s):

Conceptual Design ◽

Fast Reactor ◽

Major Part ◽

Reactor Core ◽

Core Concept ◽

Safety Requirements ◽

Burnable Poison ◽

The Core ◽

Long Lifetime ◽

Long Life

A conceptual design for a 50MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up.

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Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽

10.1515/kern-2019-0054 ◽

2021 ◽

Vol 86 (1) ◽

pp. 45-49

Author(s):

N. V. Maslov ◽

E. I. Grishanin ◽

P. N. Alekseev

Keyword(s):

Fast Reactor ◽

Reactor Core ◽

Heat Removal ◽

Design Solution ◽

Fast Neutrons ◽

Steel Shell ◽

Primary Circuit ◽

Coated Particles ◽

The Core ◽

Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

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CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

Volume 9: Student Paper Competition ◽

10.1115/icone26-81626 ◽

2018 ◽

Author(s):

Jing Chen ◽

Dalin Zhang ◽

Suizheng Qiu ◽

Kui Zhang ◽

Mingjun Wang ◽

...

Keyword(s):

Fast Reactor ◽

Power Distribution ◽

Three Dimensional ◽

Reactor Core ◽

Heat Removal ◽

The Core ◽

Computational Fluid Dynamics Cfd ◽

Heat Removal System ◽

China Experimental Fast Reactor ◽

Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

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Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

EPJ Web of Conferences ◽

10.1051/epjconf/202022503007 ◽

2020 ◽

Vol 225 ◽

pp. 03007

Author(s):

Tanja Goričanec ◽

Domen Kotnik ◽

Žiga Štancar ◽

Luka Snoj ◽

Marjan Kromar

Keyword(s):

Monte Carlo ◽

Neutron Transport ◽

Reactor Core ◽

Response Prediction ◽

Detector Response ◽

Design Calculation ◽

Monte Carlo Code ◽

Control Rod ◽

The Core ◽

Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

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Structural Confinement and Safety Design of a Fast Reactor Core

Journal of Mechanical Design ◽

10.1115/1.3254906 ◽

1981 ◽

Vol 103 (2) ◽

pp. 289-294

Author(s):

F. D. Ju ◽

J. G. Bennett

Keyword(s):

Fast Reactor ◽

Lateral Pressure ◽

Numerical Solutions ◽

Homogeneous Medium ◽

Safety Margin ◽

Reactor Core ◽

Gas Pressure ◽

Equivalent Stiffness ◽

Friction Forces ◽

The Core

In certain fast-reactor designs, the core is an assemblage of a large number of containers of long, hexagonal, hollow cylinders mounted vertically. These so-called “hex-cans” sit individually on coolant nozzles held down by their own weight, and are held as a group laterally at two levels by two constraint rings. At operating temperature, the rings bear on the hex-can assembly because of differences in thermal expansion. The compression of the rings on the hex-can assembly serves to prevent lifting of the can individually or in groups because of any accidental buildup of gas pressure. In the analysis, it is observed that the large number of hexcans and the distribution of the temperature field are such that the cross section of the reactor core can be treated as in a locally uniform dilatational field. An approximate equation was developed relating the plane deformation of a hollow hex cylinder to the global lateral pressure. The parameters are the material constitution and the thickness index (the ratio of the interior and the exterior cross-flat dimensions). The effective range of the equation covers the thickness ratio from zero to the stability limit when the wall becomes too thin resulting in buckling under the lateral pressure. The design equation is exact for zero thickness index. For hollow hex cylinders, numerical solutions were also obtained by the finite element method as a comparison. For a thickness index of 0.9 to 0.95, the difference is less than 0.1 percent. The cylinder constitutive equation is then used to determine an equivalent stiffness for a solid hex cylinder that is to have the same deformation as the given hex-can. The entire planar core region is then analyzed as a homogeneous medium of the equivalent stiffness. The method was applied to the core confinement design for a fast reactor. The thermoelastic solution was then applied to a relatively simpler configuration than the actual case to give a measure of the lateral pressure. The available friction forces for various lift configurations were then obtained. The gas pressure necessary to overcome the minimum friction force thus resulted. In addition, using the lateral pressure, the safety margin of the wall thickness of the hex-can for stability failures was determined.

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Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

ASME 2010 13th International Conference on Environmental Remediation and Radioactive Waste Management, Volume 1 ◽

10.1115/icem2010-40201 ◽

2010 ◽

Author(s):

Rizwan Ahmed ◽

Gyunyoung Heo ◽

Dong-Keun Cho ◽

Jongwon Choi

Keyword(s):

Radioactive Waste ◽

Cross Sections ◽

Reactor Core ◽

International Standards ◽

Monte Carlo Code ◽

Capture Cross ◽

Structural Components ◽

The Core ◽

Waste Classification ◽

Capture Cross Sections

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

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Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications ◽

10.1115/icone17-75946 ◽

2009 ◽

Author(s):

Atso Suopaja¨rvi ◽

Teemu Ka¨rkela¨ ◽

Ari Auvinen ◽

Ilona Lindholm

Keyword(s):

Pressure Vessel ◽

Source Term ◽

Reactor Core ◽

Severe Accident ◽

Ruthenium Oxides ◽

The Core ◽

The Standard Model ◽

Reactor Building ◽

Release Model ◽

Scoping Studies

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.

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