• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.470-482
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• 1995

The Ulchin 3&4, which are 2825 MWt PWRs, adopted Safety Depressurization System (SDS) to mitigate the beyond design basis event of Total Less of Feedwater(TLOFW). In this study the results and methodology of the analyses for the determination of SDS bleed capacity are discussed. The SDS design bleed capacity has been determined from the CEFLASH-4AS/REM simulation according to the following design criteria : 1) Each SDS flow path, in conjunction with one of two High Pressure Safety Injection (HPSI) pumps, is designed to have a sufficient capacity to prevent core uncovery if one SDS path is opened simultaneously with the opening of the Pressurizer Safety Valves (PSVs). 2) Both SDS bleed paths are designed to have sufficient total capacity with both HPSI pumps operating to prevent core uncovery if the Feed and Bleed (F&B) initiation is delayed up to thirty minutes from the time of the PSVs lift. To verify the results of CEFLASH-4AS/REM simulation a comparative analysis kas also been per-formed by more sophisticated computer code, RELAP5/MOD3. The TLOFW event without operator recovery and TLOFW event with F&B are analyzed. The predictions by the CEFLASH-4AS/REM of the transient too phase system behavior are in good qualitative and quantitative agreement with those by the RELAP5/MOD3 simulation. Both of the results of analyses by CEFLASH-4AS/REM and RELAP5/MOD3 have demonstrated that decay heat removal and core inventory make-up can be successfully accomplished by F&B operation during now event for the Ulchin 3&4.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.179-188
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• 2007

In this work $^{14}C\;and\;^3H$ distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated. It was found that the recovery percent of $^{14}C$ by the wet oxidation-acid stripping was $81%{\sim}100%$ for the added activity range of $^{14}C,\;0.72\;Bq{\sim}460\;Bq$, and it was not affected by the kinds of stripping acids, 3N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$. And the recovery percent of $^3H$ by distillation using the same apparatus was $81%{\sim}101%$ for the added activity range of $^3H,\;0.60\;Bq{\sim}435\;Bq$. Among the tested stripping acids, 3\;N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$, only the trapped $^3H$ solution by distillation in $3\;N-H_2SO_4$ was compatible with the 3H scintillator, Ultimagold XR. Neither of the $^{14}C\;and\;^3H$ trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 $N-H_2SO_4$ stripping contained gamma nuclides. However, some gamma nuclides, $^{60}Co,\;^{134}Cs,\;^{137}Cs\;and\;^{54}Mn$, were found in the trapped $^3H$ solutions of the spent resins by the wet oxidation-3 N-HCl stripping. It was the same for the $^3H$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). Meanwhile only two nuclides, $^{134}Cs,\;and\;^{134}Cs$, were found in the $^{14}C$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the $^{14}C$ in the spent resins existed as inorganic carbon form, more than about 70% of the total $^{14}C$ content. Among the analyzed 30 spent ion exchange resin samples, the average concentration of $^{14}C$ and $^3C$ for the high radioactive samples, 8 samples, was $19000\;Bq/g{\pm}41000\;Bq/g,\;670\;Bq/g{\pm}460\;Bq/g$ and that for the low radioactive samples, 22 samples, was $4.2\;Bq/g{\pm}4.3\;Bq/g,\;6.0\;Bq/g{\pm}5.3\;Bq/g$, respectively. And the average $^{14}C/^3H$ ratio for the high radioactive samples, was higher, 28, than that of low radioactive samples, 0.70. Some linear relationship trend was found between the activity concentrations of $^{14}C\;and\;^3H$.