• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1464-1478
• /
• 2021

Selective Production of Exotic Species is an innovative plant for advanced nuclear physic studies. A radioactive beam, generated by using an UCx target-ion source system, is ionized, selected and accelerated for experimental objects. Very high vacuum conditions and appropriate safety systems to storage exhaust gases are required to avoid radiological risk for operators and people. In this paper, Failure Mode, Effects, and Criticality Analysis of a preliminary design of high activity gas recovery system is performed by using a modified Fuzzy Risk Priority Number to rank the most critical components in terms of failures and human errors. Comparisons between fuzzy approach and classic application allow to show that Fuzzy Risk Priority Number is able to enhance the focus of risk assessments and to improve the safety of complex and innovative systems such as those under consideration.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1226-1233
• /
• 2017

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. The database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.18-24
• /
• 2018

This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor ($k_{eff}$) versus burnup (B) are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B)-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE)14 $10{\times}10$ boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC)-68 storage cask. The results revealed that the curves of $k_{eff}$ versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of $k_{eff}$, ${\Delta}k$) in some compound effects was not a summation of the all ${\Delta}k$ resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of $k_{eff}$ versus B for both single and compound effects.

• Nuclear Engineering and Technology
• /
• v.20 no.1
• /
• pp.27-34
• /
• 1988

Since the lack of the spent fuel storage capcity has been expected for all Korean nuclear power plants in the mid-1990s, the maximum density rack (MDR) with consolidated fuels can be proposed to overcome the shortage of the storage capacity in KNU 9 & 10 which have most limited capacities. To ensure the safety when the alternatives are applied in the KNU 9 & 10, the multiplication factor are calculated with varying the rack pitch and the thickness of consolidated storage box by the AMPX-KENO IV codes. The computing system is verified by the benchmark calculation with criticality experiments for arrays of consolidated fuel modules, which was reported by B & W in 1981. Also an abnormal condition, i.e. malposition accident, is simulated. The results indicate that the KNU 9 & 10 storage pools with consolidated fuel are safe in the view of the criticality. Thus the storage capacity can be expanded from 9/3 cores into 27/3 cores even with considering equipments and cooling spaces.

• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.187-201
• /
• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4759-4775
• /
• 2022

The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo' based similarity law was employed for the in-vessel, while Ishii's similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood's metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.1
• /
• pp.41-50
• /
• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11a
• /
• pp.153-162
• /
• 2005

In this Paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ${\~}$20 tons.

• Nuclear Engineering and Technology
• /
• v.46 no.3
• /
• pp.313-342
• /
• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• Journal of the Ergonomics Society of Korea
• /
• v.32 no.1
• /
• pp.117-124
• /
• 2013

Objective: This paper presents additional considerations related to organization and safety culture extracted from recent human error incidents in Korea, such as station blackout(i.e., SBO) in Kori#1. Background: Safety culture has been already highlighted as a major cause of human errors after 1986 Chernobyl accident. After Fukushima accident in Japan, the public acceptance for nuclear energy has taken its toll. Organizational characteristics and culture became elucidated as a major contributor again. Therefore many nuclear countries are re-evaluating their safety culture, and discussing any preparedness and its improvement. On top of that, there was an SBO in 2012 in the Kori#1. Korean public feels frustrated due to the similar human errors causing to a catastrophe like Fukushima accident. Method: This paper reassesses Japan's incidents, and revisits Korea's recent incidents. It focuses on the analysis of the hazards rather than the causes of human errors, the derivation of countermeasures, and their implementation. The preceding incidents and conclusions from Japanese experience are also re-analyzed. The Fukushima accident was an SBO due to the natural disaster such as earthquakes and a successive tsunami. Unlike the Fukushima accident, the Kori#1 incident itself was simple and restored without any loss and radioactive release. However, the fact that the incident was deliberately concealed led to massive distrust. Moreover, the continued violation of rules and organized concealment of the accident are serious signs of a new distorted type of human errors, blatantly revealing the cultural and fundamental weakness of the current organization. Result: We should learn from Japanese experiences who had taken pride in its safety technology and fairly high confidence in safety culture. Japan's first criticality accident in JCO facility splashed cold water on that confidence. It has turned out to be a typical case revealing the problems in the organization and safety culture. Since Japan has failed to gain lessons and countermeasure, the issue persists to the Fukushima incident. Conclusion: Safety culture is not a specific independent element, which makes it difficult to either evaluate it properly or establish countermeasures from the lessons. It may continue to expose similar human errors such as concealment of incident and manipulation of bad data. Application: Not only will this work establish the course of research for organization and safety culture, but this work will also contribute to the revitalization of Korea's nuclear industry from the disappointment after the export contract to UAE.