Post Irradiation Examination Facility Research Articles

Considerations of the Optimized Protective Action Distance to Meet the Korean Protective Action Guides Following Maximum Hypothesis Accidents of Major KAERI Nuclear Facilities

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea.Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered.Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility.Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.

Evaluation of Low Dose Silicon Carbide Temperature Monitors

The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy’s only designated nuclear energy user facility. Its mission is to provide nuclear energy researchers access to world-class capabilities and to facilitate the advancement of nuclear science and technology. This mission is supported by providing access to state-of-the-art experimental irradiation testing, postirradiation examination facilities, and high-performance computing capabilities, as well as technical and scientific assistance for the design and execution of projects. As part of an NSUF project, low dose silicon carbide monitors were irradiated in the Belgian Reactor 2 and were then evaluated both at the SCK•CEN and at the Idaho National Laboratory (INL) High Temperature Test Laboratory to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This article will discuss the evaluation process, the irradiation test, and the performance of the low dose silicon carbide temperature monitors.

Analysis of Irradiated Pebble Bed Fuel Transfer System in Hot Cell 101 Radiometallurgy Installation

As the operating body that has authority to construct and operate a non-commercial nuclear reactor in Indonesia, BATAN has decided to develop RDE, an HTGR based nuclear reactor with pebble bed fuels type. As the post-irradiation examination facility, IRM has to prepared and developed PIE for pebble bed fuel in hot cell. Mechanisms and supporting apparatus has to be considered during transfer processes from RDE sites to hot cell. This paper aims to analyze technical design of pebble bed fuel compatibility to existing transfer system in hot cell 101 such as wall-plug and transfer cask dimension, storage rack and fuel remote handling system. Several associated R&Ds on hot cell facility in HTGR countries are reviewed to get safer and simpler transfer system. Based on determined RDE fuels and wall plug dimension, RSG-GAS existing transfer cask can be adopted. Storage rack in hot cell 101 need to be modified and for easier handling by using manipulator, aids need to be added. The technical compatibility based on the results of the study indicates that the hot cell 101 is ready to executePIE of RDE pebble bed fuels.

Application of Compton-suppressed self-induced XRF to spent nuclear fuel measurement

Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)’s spent nuclear fuel at the Korea atomic energy research institute (KAERI)’s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum’s background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed selfinduced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel—Design concept and experimental demonstration

This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument׳s response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

조사후시험시설에서의 핫셀 배기포집시스템의 성능평가

Radioactivity of high concentrations have existed in the handling nuclear materials in hot cell of PIEF(Post Irradiation Examination Facility). The exhaust filtration system was enabled to process cylindrical filters by using a manipulator in the hot cell. By establishing a double filtration system with two filters, backup protection against leakage or failure of the first is provided by the second filter. Additionally, this a arrangement is arrange intended to increase the total filtration efficiency. The result of the pressure drop changing in the air flow of the cylindrical and HEPA filters is observed by a curved line. A filtering efficiency of more than 99.99% to particle appears in the upstream and downstream during the efficiency test on the HEPA filters. The V-pleats type had a lower pressure drop than the separator type. There was no damage during usage and was found to be suitable with high capacity of air volume. Therefore, by carrying out performance tests of the exhaust filtration system, efficiency and safety can be achieved.

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.

New progresses in the IFMIF target and test cell design and a proposal for the specimen flow

The development of the conceptual design of the IFMIF target and test cell (TTC) is briefly summarized by outlining the previous reference TTC design and the current modular test cell (MTC) concept. Based on the MTC concept, the latest progresses of the preliminary engineering design of the key TTC components, including the TTC vessel, the top shielding plugs, the removable intermediate ring, and the test module interface heads, are described. A specimen flow, based on handling requirements of the high flux test module, between the TTC, access cell (AC), test module handling cell (TMHC) and the post irradiation examination facilities is proposed as well as the function of the AC and TMHC is preliminary defined. The TMHC is proposed to be divided into a component handling cell and a rig handling cell regarding the dimension differences of the components to be handled inside of the cells. Re-use of irradiated specimens for another campaign of irradiation is also considered in this specimen flow.

Utilisation of a hydrogen uranyl phosphate-based ion exchanger supported on a biofilm for the removal of cobalt, strontium and caesium from aqueous solutions

The problems of nuclear waste treatment and decontamination have been paramount for a number of years and continue to adversely affect the acceptability and expanded uses of nuclear power. By and large, methodologies for waste treatment exist at varying efficiency and correspondingly high expense. With a view to nuclear waste remediation, a novel bio-inorganic ion exchanger was developed which utilises a microbial system to synthesise hydrogen uranyl phosphate which serves as an intercalative ion exchanger for the fission products 137Cs, 90Sr and the activation product 60Co. Hydrogen uranyl phosphate-coated Serratia sp. biofilm, immobilised onto polyurethane reticulated foam, was used in a flow through reactor, with complete substitution of H + in the matrix by the intercalating target metal species. Experiments carried out in South Korea have confirmed these results using test nuclides ( 137Cs, 85Sr and 60Co), as well as water from a pool, where spent fuels are stored and cooled, sited in a post irradiation examination facility at the Korea Atomic Energy Research Institute. Over 97% and ∼ 85% removal of 137Cs and 60Co, respectively, from the pool water were reported, at a flow residence time of 7 days.

Criticality safety of fuel pins in a post irradiation examination cell

This paper discusses the criticality safety of mixed carbide (70 PuC: 30 UC) fast reactor fuel pins inside the hot cells of a post irradiation examination facility. The criticality calculations are done as a part of the licensing requirements. The present calculations pertain to a situation where 183 fuel pins from three fuel sub-assemblies are inadvertently present at a time in one of the hot cells. In this study, apart from the normal operating condition of the hot cell, two scenarios are considered. One is an accidental situation in which the cell is flooded with water. The other is a hypothetical roof collapse resulting in the fuel pins being embedded in concrete. The calculations are done using Monte Carlo code KENO Va.

Postirradiation examination of Kori-1 nuclear power plant fuels

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institute. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity.

Design and operation of post-irradiation examination facilities: Part I. Analysis of JHL-1 four year operation experience

The Hot Laboratory of the Japan Atomic Energy Institute consists of two parts: JHL-1 and JHL-2. The former is a row of multi-purpose caves, and the latter consists of single-purpose caves. The latter's design was based on the JHL-1's operation data. In this paper the results of analysis of the JHL-1's 4 years of operation are reported. This evaluation of accumulated operational experience may serve as a guide for the rational planning and design development of similar hot laboratory installations.

Design and equipment of the Jaeri post irradiation examination facility

This paper gives some details of design and construction of mechanical devices for the post irradiation examination facility in JAERI (Japan Atomic Energy Research Institute). Shielding door drive mechanism, cell conveyor system, sample transfer ports were devised with respect to complete remote control. Also are presented general descriptions of layout, cell compositions and other mechanical equipments.