• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.180-185
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• 1987

A computer code, KAFEPA, for analysing in-reactor behavior of a PHWR-fuel rod under reactor normal operating condition was developed. This code, KAFEPA, corresponds to the ELESIM code that was developed for the same purpose by AECL. Even though the KAFEPA originated from the ELESIM, it contains more accurate and theoretical models in comparison with the ELESIM, such as fission gas release model, in-reactor densification model and a new database for neutron flux depression across the radial direction in a fuel pellet. The KAFEPA code was verified by comparing the predictions with 22 measurements of fission product gas release. The predictions of the KAFEPA was well agreed with the experimental data.

• 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.

• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.33-45
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• 2013

Systems Engineering Approach was applied to the development of operator-support core management system based on the on-site operation experience and document of core management procedures, which is for enhancing operability and safety in PHWR (Pressurized Heavy Water Reactor) operation. The dissertation and definition of the system were given on th basis of investigating and analyzing the core management procedures. Fuel management, detector calibration, safety management, core power distribution monitoring, and integrated data management were defined as main user's requirements. From the requirements, 11 upper functional requirements were extracted by considering the on-site operation experience and investigating documents of core management procedures. Detailed requirements of the system which were produced by analyzing the upper functional requirements were identified by interviewing members who have responsibility of the core management procedures, which were written in SRS (Software Requirement Specification) document by using IEEE 830 template. The system was designed on the basis of the SRS and analysis in terms of nuclear engineering, and then tested by simulation using on-site data as a example. A model of core power monitoring related to the core management was suggested and a standard process for the core management was also suggested. And extraction, analysis, and documentation of the requirements were suggested as a case in terms of systems engineering.

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.187-196
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• 2004

The health effects resulting from severe accidents of typical 1,000MWe KSNP(Korea Standard Nuclear Plant) PWR and typical 600MWe CANDU(CANada Deuterium Uranium) plants were estimated and compared. The population distribution of the site extending to 80km for both site were considered. The releaese fraction for various source term categories(STC) and core inventories were used in the estimation of the health effects risks by using the MACCS2(MELCOR Accident Consequence Code System2) code. Individuals are assumed to evacuate beyond 16km from the site. The health effects considered in this comparative study are early and cancer fatality risk, and the results are presented as CCDF(Complementary Cumulative Distribution Function) curves considering the occurrence probability of each STC's. According to the results, the early and cancer fatality risks of PHWR plants we lower than those of PWR plants. This is attributed the fact that the amount of radioactive mateials that released to the atmosphere resulting from the postulated severe accidents of PHWR plants are smaller than that of PWR plants. And, the dominating initiating event of STC that shows maximum early and cancer fatality risk is SGTR(Steam Generator Tube Rupture) for both plants. Therefore, the appropriated actions must be taken to reduce the occurrence probability and the amounts of radioactive materials released to the environment in order to protect the public for both PWR and PHWR plants.

• Corrosion Science and Technology
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• v.7 no.2
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• pp.85-91
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• 2008

By operation in aqueous environment at high temperature and pressure, the structural materials from Primary Heat Transport System (PHTS) cover with protective oxide films, which maintain the corrosion rate in admissible limits. A lot of potential factors exist, which conduct to degradation of the protective films and consequently to intensification of the corrosion processes. The existing experience of different nuclear reactors shows that the water chemistry has an important role in integrity maintaining of the protective oxide films. To investigate the influence of water chemistry (pH, O2 dissolved, $Cl^-$, $F^-$) on corrosion of some structural materials (carbon and martensitic steel, Zr and Ni alloys) and to establish the maximum permissible values, corrosion experiments by static autoclaving and electrochemical methods were performed. The experimental results allowed us to establish the contribution of the water chemistry in initiation and evolution of some accelerated corrosion processes.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.78-83
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• 2003

The aim of this study is to investigate the effect of hydrogen embrittlement od Zr-2.5Nb CANDU pressure tube. The test were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300$^{\circ}C$). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250$^{\circ}C$. From the fracture toughness test, it was found that fracture toughness dJ/da was increased up to 200$^{\circ}C$ and then decreased.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.40-44
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• 2015

Sensitivity on the reactor safety was evaluated for the safety margin and time delay applied to the opening pressure of liquid relief valve(LRV) of the primary heat transport system(PHTS) in the pressurized heavy water reactor(PHWR) type nuclear power plant. Since the LRV is the pressure boundary for the PHTS in the safety analysis, the operating of LRV has a significant effect on the safety analysis results. Therefore it is required during the regulatory review of Wolsong Unit 1 safety analysis to find the safety effect of the application of safety margin and time delay to the LRV opening pressure for the safety analysis of PHTS pressurizing events.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.369-377
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• 2003

Since Zr-2.5Nb pressure tubes have a high risk for the formation of blisters during their operation in pressurized heavy water reactors, there has been a strong incentive to develop a method for the non-destructive detection of blisters grown on the tube surfaces. However, because there is little mismatch in acoustic impedance between the hydride blisters and zirconium matrix, it is not easy to distinguish the boundary between the blister and zirconium matrix with conventional ultrasonic methods. This study has focused on the development of a special ultrasonic method, so called ultrasonic velocity ratio method for a reliable detection of blisters formed on Zr-2.5Nb pressure tubes. Hydride blisters were grown on the outer surface of the Zr-2.5Nb pressure tube using a cold finger attached to a steady state thermal diffusion equipment. To maximize a difference in the ultrasonic velocity in hydride blisters and the zirconium matrix, the ultrasonic velocity ratio of longitudinal wave to shear wave, $V_L/V_S$, has been determined based on the flight time of the longitudinal echo and reflected shear echo from the outer surface of the tubes. The feasibility of the ultrasonic velocity ratio method is confirmed by comparing the contour plots reproduced by this method with those of the blisters grown on the Zr-2.5Nb pressure tubes.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.147-157
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• 2004

The production of $^{14}C$ occurs in the Moderator(MOD), Primary Heat Transport System (PHTS), Annulus Gas System(AGS) and Fuel in the CANDU reactor. Among the four systems, The MOD system is the largest contributor to $^{14}C$ production(approximately 94.8%). $^{14}C$ is distributed of $^{14}CO_2$, $H_2^{14}CO_3$, $H^{14}{CO_3}^-$ and $^{14}{CO_3}^{2-}$ species as a function of the pH of water. Of these species, $H_2^{14}CO_3$ and $H^{14}{CO_3}^-$ form are predominant because the pH of MOD system is ＞ 5. In this paper, adsorption-desorption characteristics of bicarbonate ion (${HCO_3}^-$) by IRN 150 resin was investigated. ${HCO_3}^-$ ion existed in neutral condition(app. pH 7)was reacted with ion exchange resin (IRN-150) and saturated with it. Then $NaNO_3$ and $Na_3PO_4$ solutions selected as extraction materials were used to make an investigation into feasibility of ${HCO_3}^-$ extraction from resin saturated with ${HCO_3}^-$. Desorption of $CO^{2+}$ and $Cs^+$ ion by $Na^+$ ion was not occurred, and desorption of ${HCO_3}^-$ ion by ${NO_3}^-$ and ${PO_4}^{3-}$ was occurred slowly. Also, the status of ion exchange which is used in Wolsong NPPs and generation of spent resin yearly were surveyed.