• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1935-1945
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• 2023

The radiological safety of workers in a newly developed microwave-based spent resin treatment facility was assessed based on work location and operational scenarios. The results show that the remote-operation room worker was exposed to maximum annual dose of 3.19E+00 mSv, which is 15.9% of the dose limit, thereby confirming radiological safety. Inside the pathway, annual doses in the range of 7.87E-02-2.07E-01 mSv were measured initially at the mock-up tank and later at the point between the spent resin separation and treatment parts. The dose of emergency maintenance workers was below the dose limit (4.08E-03-4.99E+00 mSv); however, before treatment (separation and microwave), the dose of maintenance and repair workers exceeded the dose limit. The doses of the effluent removal workers at the zeolite and activated carbon storage tank and spent resin storage tank were the lowest at 2.79E-01-2.87E-01 mSv and 9.27E-01 mSv in "1 h" and "4-5 h of operation", respectively. The immediately lower and upper layers of the facility room exhibited the highest annual doses of 1.84E+00 and 3.22E+00 mSv, respectively. Through this study, a scenario that can minimize the dose considering the movement of spent resin through the facility can be developed.

• Journal of Energy Engineering
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• v.7 no.1
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• pp.96-102
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• 1998

The use of computers as part of nuclear safety systems elicits additional requirements-software verification and validation (v/v), hardware qualification-not specifically addressed in general industry fields. The computer used in nuclear power plants is a system that includes computer hardware, software, firmware, and interfaces. To develop the computer systems graded with nuclear safety class, the developing environments have to be required in advance and the developed software have to be verified and validated in accordance with nuclear code and standards. With this requirements, the test facility for Inadequate Core Cooling Monitoring System (ICCMS) as one of safety systems in the nuclear power plants was developed. The test facility consists of three(3) parts such as Input/Output (I/O) simulator, Plant Data Acqusition System (PDAS) cabinets and supervisory computer. The performance of the system was validated by manual test procedure.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.347-357
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• 2023

United States Nuclear Regulatory Commission (U.S. NRC) specifies regulations on obtaining licenses and describes the technical position on the average waste concentration, also known as Concentration Averaging and Encapsulation Branch Technical Position (CA BTP); CA BTP helps classify blendable waste and discrete items and address concentration averaging. The technical position details are reviewed and compared in a real environment in Korea. A few cases of concentration averaging based on the application of CA BTP to domestic radioactive waste are presented, and the feasibility of the application is assessed. The radioactive waste considered herein does not satisfy the Disposal Concentration Limit (DCL) of the second-phase disposal facility while applying the preliminary classification. However, if CA BTP is applied when the radioactive waste is mixed with other radioactive waste items in a large and heavy container, it can be disposed of at the second-phase disposal facility in Gyeongju Repository. To apply the CA BTP of the U.S. NRC, it is necessary to investigate the safety assessment conditions of the US and Korea.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.05a
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• pp.37-38
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• 2012

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1190-1191
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• 2004

New $4^{th}$ generation synchrotron facility, XFEL, is almost similar to previous $3^{rd}$ generation synchrotron facility in the view of radiological aspects and most important positions are a dump and synchrotron radiation beam line. In this paper, tile radiation protection solutions for them and undulator are suggested and discussed.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.415-422
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• 1996

Tritium removal from tritiated heavy water in a PHWR is the most effective way in reducing workers' internal dose and radioactivity emissions from Wolsong NPP. The optimum design of the Wolsong TRF (Tritium Removal Facility) was carried out using an approximate short-cut method with an assumption that the TRF, designed to extract 8 MCi per year of elemental tritium from a heavy oater feedstream, uses Liquid Phase Catalytic Exchange (LPCE) front-end process and Cryogenic Distillation (CD) process.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.505-510
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• 1996

A simple model for RAM (Reliability, Availability and Maintainability) analysis and its computer code are developed for application to DUPIC fuel fabrication system. The approach is obtained by linking the allocation model (top-down method) to bottom-up method for RAM analysis. As a result, the availability requirement of subsystem, as well as the buffer storage requirement between processes, are evaluated for the DUPIC facility..