• Title/Summary/Keyword: Pressurized water reactor

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Effects of High Damping Rubber Bearing on Horizontal and Vertical Seismic Responses of a Pressurized Water Reactor

  • Bong Yoo;Lee, Jae-Han;Koo, Gyeong-Hoi
    • Proceedings of the Korean Nuclear Society Conference
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    • 1995.05a
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    • pp.1021-1026
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    • 1995
  • The seismic responses of a base isolated Pressurized Water Reactor (PWR) are investigated using a mathematical model which expresses the superstructure as lumped mass-spring model and the seismic isolator as an equivalent spring-damper. Time history analyses are performed for the 1940 E1 Centre earthquakes in both horizontal and vertical directions. In the analysis, structural damping of 5% is used for the superstructure. The isolator damping ratios of 12% for horizontal and 5% for vertical directions are used. The acceleration responses in base isolated PWR superstructure with high damping rubber bearings are much smaller than those in fixed base structure in horizontal direction. However, the vertical acceleration responses at the superstructure in the base isolation system are amplified to some extent. It is suggested that the vertical seismic responses at the superstructure should be reduced by introducing a soft vertical isolation device.

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Reliability Assessment by the Scoring Model for the Advanced Pressurized water Reactor 1400MWe Project Selection under Uncertainty (신형경수로 1400을 위해 점수산정 모형에 의한 신뢰성 평가)

  • 강영식
    • Journal of Korean Society of Industrial and Systems Engineering
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    • v.25 no.6
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    • pp.23-35
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    • 2002
  • The problem of system reliability is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, environment destruction, and fatal damage of human. Therefore the purpose of this study has developed the reliability evaluation model through the scoring model by the quantitative and qualitative factors in order to justify the evaluation considering the advanced safety factors in the Advanced Pressurized water Reactor 1400MWe(APR 1400MWe) under uncertainty. Especially, the qualitative factors considering the human, information control, and quality factors for the systematic and rational justification have been closely analyzed. The proposed model can be simply applied in real fields in order to minimize the industrial accidents in the digitalized nuclear power plant.

ROSA/LSTF test and RELAP5 code analyses on PWR steam generator tube rupture accident with recovery actions

  • Takeda, Takeshi
    • Nuclear Engineering and Technology
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    • v.50 no.6
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    • pp.981-988
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    • 2018
  • An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

Power Density Distribution Calculation of a Pressurized Water Reactor with Fullscope Explicit Modeling by MCNP Code

  • Kim, Jong-Oh;Kim, Jong-Kyung
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.05a
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    • pp.179-184
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    • 1996
  • Power density distribution and criticality of a pressurized water reactor are calculated with a Monte Carlo calculation using the MCNP code. The MCNP model is based on one-eighth core symmetry. Individual fuel assemblies are modeled with fullscope three dimensional description except grid spacer. The fuel rod is divided into eight axial segments. Core internals above and below the active fuel region is represented as coolant. After 400 cycle calculations, the system converges to a k value of 1.09151$\pm$0.00066. Fission reaction rate in each rod is also calculated to use as the source term in pressure vessel fluence calculation.

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A Study on the Application of CRUDTRAN Code in Primary Systems of Domestic Pressurized Heavy-Water Reactors for Prediction of Radiation Source Term

  • Song, Jong Soon;Cho, Hoon Jo;Jung, Min Young;Lee, Sang Heon
    • Nuclear Engineering and Technology
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    • v.49 no.3
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    • pp.638-644
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    • 2017
  • The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature

  • Takeda, Takeshi
    • Nuclear Engineering and Technology
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    • v.50 no.8
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    • pp.1412-1420
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    • 2018
  • An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

Review on the Management for Radioactive Effluent and Methodology for Setting of Derived Release Limits at Pressurized Heavy Water Reactors in Korea (중수로원전 방사성유출물 관리와 유도배출한계 설정방법에 대한 고찰)

  • Kim, Hee-Geun;Kong, Tae-Young;Jeong, Woo-Tae;Kim, Seok-Tae
    • Journal of Radiation Protection and Research
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    • v.35 no.4
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    • pp.172-177
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    • 2010
  • The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed.

Calculation of Low-Energy Reactor Neutrino Spectra for Reactor Neutrino Experiments

  • Riyana, Eka Sapta;Suda, Shoya;Ishibashi, Kenji;Matsuura, Hideaki;Katakura, Jun-ichi
    • Journal of Radiation Protection and Research
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    • v.41 no.2
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    • pp.155-159
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    • 2016
  • Background: Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. Materials and Methods: To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% $^{235}U$ contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. Results and Discussion: We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. $^{241}Pu$) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate Conclusion: Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.