• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.386-391
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• 1997

It is important to understand correctly a CHF under low flow condition for the purpose of enhancing the reactor safety and performance in the LWRs. The CHF experiments have been carried out for an internally heated vertical annulus in RCS loop facility. The experimental conditions cover ranges of pressure from 1.82 to 12.08 MPa, mass flux from 300 to 550kg/$m^2$. s and inlet subcooling of 210kJ/kg. The CHF data decrease with increasing pressure at high value of mass flux. For mass flux of about 300kg/$m^2$. s, the CHF rue little influenced by pressure. The CHF data are correlated well by using the dimensionless heat flux and dimensionless mass flux for a fixed inlet subcooling except the data group of 12.08 MPa. It seems that the Doerffer correlation and Katto correlation overestimate the CHF for low pressure and lower value of mass flux within this experimental ranges. The Bowling correlation gives a better prediction than the other two correlations.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.113-121
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• 2014

In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.907-914
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• 2015

The energetic steam explosion caused by contact between the high temperature molten core and water is one of the phenomena that may threaten the integrity of the containment vessel during severe accidents of light water reactors (LWRs). We examined the dependence of steam explosion loads in a typical reactor cavity geometry on selected model parameters and initial/boundary conditions by using a steam explosion simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters, we put an emphasis on the water pool depth that has significance in terms of accident mitigation strategies including cavity flooding. The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load. The water pool depth also showed a significant impact. The energy conversion ratio based on the enthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, we proposed a simplified method for evaluation of the steam explosion load. The results showed fair agreement with JASMINE.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.9-20
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• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.431-439
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• 2018

Less mature nuclear reactor technologies are characterized by a greater uncertainty due to insufficient detailed design information, operational data, cost information, etc., but the expected performance characteristics of less mature options are usually more attractive in comparison with more mature ones. The greater uncertainty is, the higher economic risks associated with the project realization will be. Within a comparative evaluation of less and more mature nuclear reactor technologies, it is necessary to apply economic risk measures to balance judgments regarding the economic performance of less and more mature options. Assessments of any risk metrics involve calculating different characteristics of probability distributions of associated economic performance indicators and applying the Monte-Carlo method. This paper considers the applicability of statistical risk measures for different economic performance indicators within a trial case study on a comparative evaluation of less and more mature unspecified LWRs. The presented case study demonstrates the main trends associated with the incorporation of economic risk metrics into a comparative evaluation of less and more mature nuclear reactor technologies.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4338-4347
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• 2022

The CUPID code was developed and is continuously updated in KAERI. Verification and validation (V&V) is mainly done for light water reactors (LWRs). This paper describes a benchmarking of the detailed mesh level compared with sub-channel level for application to pressurized heavy water reactors (PHWRs), even though component scale comparison for the PHWR moderator system was done once before. We completed a sub-channel level comparison between the CUPID code and the ASSERT code and a CUPID code analysis. Because the ASSERT code has already been validated with numerous experiments, benchmarking with the ASSERT code will offer us more trust on the CUPID code. The target channel has high power and thus high pressure deformation. The high power channel tends to have a high possibility of critical heat flux (CHF), because a high void fraction and quality in channel exit region appear. In this research, after determining the reference grid and T/H model, we compared the sub-channel level results of the CUPID code with those of the ASSERT code.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.819-823
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• 2024

The main criteria used in NEI 18-04 to define SSCs as risk-significant include (1) the SSC is required to keep all LBEs within the F-C target, and (2) the total frequency with the SSC failed exceeds 1% of the limit for at least one of the three cumulative risk metrics used for evaluating the integrated plant risk. The first one is a reasonable criterion in determining the risk significant SSCs. However, the second criterion may not be adequate to serve the purpose of determining the risk significance of SSCs. In the second criterion, the cumulative risk metric values representing the integrated plant risk (less the preventive and mitigative effects of the SSC being evaluated) are compared to a risk limit that represents a very small contribution to the overall integrated plant risk, which corresponds appropriately to the contributions from individual SSCs. The easiest approach to redefine the NEI 18-04 definition of risk-significant SSCs in relation to the integrated plant risk metrics is to compare the difference, between the risk metric value calculated with the SSC failed and the risk metric value calculated with the SSC credited, with 1% of the risk limit established for the integrated plant risk metrics.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.2893-2905
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• 2024

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.