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  • 1
    Online Resource
    Online Resource
    Hindawi Limited ; 2017
    In:  Science and Technology of Nuclear Installations Vol. 2017 ( 2017), p. 1-10
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2017 ( 2017), p. 1-10
    Abstract: Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.
    Type of Medium: Online Resource
    ISSN: 1687-6075 , 1687-6083
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2017
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
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  • 2
    Online Resource
    Online Resource
    Hindawi Limited ; 2020
    In:  Science and Technology of Nuclear Installations Vol. 2020 ( 2020-08-25), p. 1-11
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2020 ( 2020-08-25), p. 1-11
    Abstract: The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm 3 and the 235 U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235 U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 10 14 and 8.9 ×10 13  n·cm −2 ·s −1 are found at the neutron trap and the beryllium reflector, respectively.
    Type of Medium: Online Resource
    ISSN: 1687-6075 , 1687-6083
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2020
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 3
    Online Resource
    Online Resource
    Hindawi Limited ; 2020
    In:  Science and Technology of Nuclear Installations Vol. 2020 ( 2020-12-7), p. 1-10
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2020 ( 2020-12-7), p. 1-10
    Abstract: Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4  μ Sv / h is small enough for people to work normally.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2020
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 4
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2022 ( 2022-5-18), p. 1-10
    Abstract: This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2022
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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