Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

Hanan, N. A.; Garner, P.L.

doi:10.2172/1214274

Cited by 4 publications

(8 citation statements)

References 2 publications

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“…Maximum coolant temperature does not exceed 74 • C, being much lower than start of the water nucleate boiling, whereas maximum temperature of the «hot» section of the hottest fuel element doesn't exceed 97.1 • C, being much lower than the SAV melting point. Thus, with one operating standby pump, which provides 45 It should be mentioned that the authors are engaged in the analysis like the presented here as applied to the WWR-K reactor conversion for many years [6][7][8][9][10][11][12][13][14], when various candidates for WWR-KN FA were under study, in view of proper choice for the WWR-K reactor core. ANL was a reviewer of the WWR-K RR Safety Analysis Report prepared to physical start up for the flow rate in the core 700…”

Section: Analysis Of Transientsmentioning

confidence: 99%

Safety analysis for the WWR-K research reactor converted to LEU fuel

Koltochnik

Shaimerdenov

2019

Eurasian j. phys. funct. mater.

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Recently in the WWR-K water-water research reactor the former HEU fuel, enriched to 36% in Uranium-235, was changed to LEU one, enriched to 19.7%, with substantial change of the core configuration. In view of reactor conversion, a new Safety Analysis Report (SAR) was developed for the WWR-K reactor. Substantiation of reactor safe operation under both normal operation and emergency conditions was done under thermal-hydraulic approach. In the analysis developed prior to physical start up it was assumed that a main circulation pump (MCP) provides the coolant flow rate in the core equal to 350 m 3 /h (a certified value for the pump CB-321). However, in course of the reactor physical start up it was found that it is only 250 m 3 /h. The reason was a decision of the reactor staff to reduce the primary pump power consumption, prolonging its life time.Therefor the thermal hydraulic analysis was revised, and the SAR was renewed. Safety analysis implies also consideration of some potential initiating events capable to develop into an accident. So, several typical initiating events are subject to thermal-hydraulic analysis to substantiate observance of nuclear and radiation safety in emergency situations. It is shown that, owing to proper operation of safety systems, the initiating events under consideration don't result into accidents, if two primary pumps provide not more than 585 m 3 /h.

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Section: Analysis Of Transientsmentioning

confidence: 99%

Safety analysis for the WWR-K research reactor converted to LEU fuel

Koltochnik

Shaimerdenov

2019

Eurasian j. phys. funct. mater.

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“…In the present work, the WWR-K reactor with the power of 6 MW in Kazakhstan is taken as a reference for designing a new MPRR with the power of 10 MW [17]. e reactor type is open swimming pool.…”

Section: Mprr Core Designmentioning

confidence: 99%

“…e reactor core has the height of 180 cm and the radius of 113.5 cm. Table 1 presents the main design parameters and the design targets of the MPRR core using the VVR-KN FAs [17,18]. Figure 2 shows the horizontal crosssectional view of the MPRR core.…”

Section: Mprr Core Designmentioning

confidence: 99%

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Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

Nguyen

Huynh

et al. 2020

Science and Technology of Nuclear Installations

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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/cm3 and the 235U 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 235U, 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 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

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“…Practically, the PLTEMP/ANL code was applied for steady-state thermal-hydraulic analysis for fuel conversion of research reactor cores using Russian tube fuel-types of VVR-KN and IRT-4M. A comparison of calculation results of the Kazakhstan WWR-K reactor core loaded with LEU VVR-KN fuel, using the PLTEMP/ANL and RELAP5 codes, was performed, and a good agreement was found between cladding surface temperatures calculated using PLTEMP/ ANL and those calculated using RELAP5-3D [9,10]. e PLTEMP/ANL was also used for steady-state analysis of the Uzbekistan WWR-SM and IRT-Sofia reactors loaded with LEU IRT-4M [11,12].…”

Section: Introductionmentioning

confidence: 99%

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

Nguyen

Huynh

et al. 2021

Science and Technology of Nuclear Installations

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This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

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Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

Cited by 4 publications

References 2 publications

Safety analysis for the WWR-K research reactor converted to LEU fuel

Safety analysis for the WWR-K research reactor converted to LEU fuel

Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

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