• Title, Summary, Keyword: PHWR

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Cost Comparison of PWR and PHWR Nuclear Power Plants in Korea

  • Kim, Chang-Hyo;Chung, Chang-Hyun;So, Dong-Sub
    • Nuclear Engineering and Technology
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    • v.11 no.4
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    • pp.263-274
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    • 1979
  • A statistical approach is used to investigate the relative economic advantages of pressurized water reactor (PWR) and pressurized heavy water reactor (PHWR-CANDU) nuclear power plants for hypothetical 900Mwe systems with the throwaway fuel cycle to be built in the Republic of Korea. Power cost is decomposed into the cost components related to the plant capital, operation and maintenance, working capital requirements and fuel cycle operation. The calculation of construction cost is performed with the modified version of computer code ORCOST, and the modified POWERCO-50 is used to evaluate the cost components. Most of economic parameters are treated as statistical variables, each being given with a certain range. Through a random sampling procedures. the probability histograms on unit plant construction costs and power generating costs are obtained. The power cost probability histograms of the PWR and the PHWR plants overlap considerably, and the power costs of two systems appear to be almost same with the PHWR power cost being 0.4mil1/kwh lower compared with 39.4 mills/kwh for the PWR plant (July 1986 US-dollars). When a construction period of PHWR plant is longer by one year than that of PWR plant, there is no difference in the unit power cost of two plants. This comparison leads to no definite conclusion on the cost advantage of the PWR plant versus the PHWR plant. We conclude that the selection issue of nuclear power plants in Korea still remains an open question and that future effort to solve this question should be made toward economic quantification of those factors such as technology transfer and localization.

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A Numerical Model for Predicting the Radial Power Profile in CANDU-PHWR Fuel Pellet (CANDU-PHWR 핵연료 소결체의 반경방향 출력분포 수치모형)

  • Woan Hwang;Suk, Ho-Chun;Jae, Won-Mok
    • Nuclear Engineering and Technology
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    • v.23 no.4
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    • pp.444-455
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    • 1991
  • An accurate and fast running NEDAR model for calculating radial power profile throughout fuel life in both solid and annular pellets for existing and advanced CANDU-PHWR-fuel was developed in this work. This model contains resultant flux depression equations and neutron depression data tables which have been developed for CANDU-PHWR fuel of pellet with the diameter 8.0 to 19.5 mm and enrichment 0.71-6.0 wt % U-235, over a bumup range of 0 to 840 MWh /kgU (35000 MWD/T). In order to obtain the neutron flux distribution in the fuel pellet, the CE-HAMMER physics code was run for a neutron flux spectrum appropriate to a CANDU-PHWR to give predictions of radial power profile for several ranges of fuel design parameters. The results, which were calculated by the CE-HAMMER physics code, were fitted to an equation which is solved in terms of Bessel and exponential functions in order to obtain the parameters, $textsc{k}$, $\beta$ and λ in the resultant equation. The present NEDAR model produce a radial profile which, when normalized to unity at the pellet surface, is slightly higher than the profile of the original ELESIM data table. The predictions of the fission gas release by KAFEPA-NEDAR are in slightly better agreement with the experiments than those of ELESIM. The NEDAR model described in this study has been shown to provide an effective, reliable, and accurate method for determining radial power profiles in CANDU-PHWR fuel rods without incurring a significant increase in computing time.

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Burst criterion for Indian PHWR fuel cladding under simulated loss-of-coolant accident

  • Suman, Siddharth
    • Nuclear Engineering and Technology
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    • v.51 no.6
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    • pp.1525-1531
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    • 2019
  • The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

Seismic Analysis of Spent Fuel Storage Structures for PHWR Plant (중수로형 핵연료 저장대의 내진해석 방법)

  • 신태명
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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    • pp.338-344
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    • 2003
  • The seismic analysis method of spent fuel storage structures for PHWR plant is introduced in comparison with the method for PWR plant. Investigating the structural characteristics of the storage structures, the former is vertically stacked fuel storage trays, while the latter is welded honeycomb type structure. However, as both structures are submerged and free standing, the analysis methods to anticipate the seismic response of both structures are complicated. For the better estimation of actual seismic response, how to model the dynamic properties and the structural behaviour is the key issue. In this paper, the overall procedures of the seismic modelling and stability check for seismic sliding and overturning of the two different storage structures are discussed in the viewpoint of analysis reliability

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Hydride Embrittlement Behavior at the LBB Evaluation of PHWR Pressure Tube (중수로 압력관 LBB 평가에서의 수소화물에 의한 취화거동)

  • Oh, Dong-Joon;Kim, Young-Suk
    • Proceedings of the KSME Conference
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    • pp.1192-1197
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    • 2003
  • The aim of this study is to investigate the hydride embrittlement when the LBB evaluation is carried out for the integrity of PHWR Pressure Tubes. The transverse tensile and CCT tests were performed at three hydrogen concentrations while the test temperatures were changed (RT to $300^{\circ}C$). The specimens were directly machined from the pressure tube retaining original curvature. 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 $250^{\circ}C$. Using the DHC test results, the CCL and LBB time were calculated and compared. The hydride embrittlement behavior at the LBB evaluation was definitely showed.

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Incremental Cross Sections for CANDU-PHWR Core Analysis (CANDU-PHWR의 증분단면적 계산방법에 대한 연구)

  • Hang Bok Choi;Seong Yun Kim;Chang Hyun Chung
    • Nuclear Engineering and Technology
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    • v.17 no.2
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    • pp.98-104
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    • 1985
  • A number of reactivity devices are distributed in a CANDU-PHWR core to control the power distribution and excess reactivity. The effects of these devices are represented by incremental cross sections in core analysis. The incremental cross sections are generated by the SUPERCELL code using the two-group constant set calculated by the lattice code, WIMS. The incremental cross sections are then assessed for adjusters and zone controller by core simulation. Reactivity worth and channel powers are compared to the reference values. The deviation of reactivity worth and the maximum channel power are less than 0.97% and 0.6%, respectively, for the initial and equilibrium core.

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Effect of Hydride of the PHWR Pressure Tube on the LBB Evaluation (중수로 압력관의 수화물이 LBB평가에 미치는 영향)

  • Oh, Dong-Joon;Kim, Young-Suk
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.28 no.5
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    • pp.610-616
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    • 2004
  • The aim of this study was to investigate the hydride embrittlement when the LBB evaluation was carried out for the integrity of PHWR Pressure Tubes. The transverse tensile and CCT toughness tests were performed at three hydrogen concentrations while the test temperatures were changed (RT to 30$0^{\circ}C$). Both the transverse tensile and the fracture toughness tests showed the hydrogen embitterment clearly at RT but this phenomenon was disappeared while the test temperature arrived at 25$0^{\circ}C$. Using the DHC test results, the CCL and LBB time were calculated and compared. The hydride embrittlement at the LBB evaluation made the LBB time short definedly. If the operating temperature, DHCV and LBB deterministic parameters such as A and m were known, LBB time could be estimated without the calculation of CCL.

Development of Fault Diagnosis System for Ram in PHWR Plant (램집합체 이상진단 시스템의 개발)

  • 변승현;조병학;신창훈;양장범
    • Proceedings of the Korean Society of Precision Engineering Conference
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    • pp.1319-1322
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    • 2004
  • In this paper, a fault diagnosis system for ram in PHWR plant is developed. The developed diagnosis system can detect the ram stuck phenomena due to increased ball wear and damage in ball nut using discrete wavelet transform before the ram is stuck. The validity of developed diagnosis system is shown via experiments using ball nut characteristic test equipment.

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PROGRESS IN NUCLEAR FUEL TECHNOLOGY IN KOREA

  • Song, Kun-Woo;Jeon, Kyeong-Lak;Jang, Young-Ki;Park, Joo-Hwan;Koo, Yang-Hyun
    • Nuclear Engineering and Technology
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    • v.41 no.4
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    • pp.493-520
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    • 2009
  • During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

Fuel-Coolant Interaction Visualization Test for In-Vessel Corium Retention External Reactor Vessel Cooling (IVR-ERVC) Condition

  • Na, Young Su;Hong, Seong-Ho;Song, Jin Ho;Hong, Seong-Wan
    • Nuclear Engineering and Technology
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    • v.48 no.6
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    • pp.1330-1337
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    • 2016
  • A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.