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.
Research reactor is an important tool for supporting research and development in a wide range of scientific aspects such as neutron scattering, neutron activation analysis, material testing, medicine application, biological science, and education and training [
A number of MPRRs with the power output greater than 10 MW currently in operation, under construction, and planned were reviewed [
In the present work, a conceptual design of 10 MW MPRR core loaded with the Russian tube-type VVR-KN fuel assemblies has been conducted. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code. The VVR-KN fuel assembly contains low enriched uranium fuel with the density of 2.8 gU/cm3. The active core is surrounded by a beryllium reflector to attain high thermal neutron fluxes and high performance of irradiation channels. In the design process, the numbers of fuel assemblies and control rods and their arrangement in the core are determined for obtaining enough excess reactivity during the core operation for compensating a number of irradiation applications, while ensuring safety requirements. Thus, neutronic characteristics of the MPRR core including burnup performance, power distribution and power peaking factor, control rod worths, reactivity coefficients, sizes and types of irradiation holes for a variety of applications, neutron fluxes at irradiation channels, and so forth have been investigated. One of the design objectives is to attain high thermal neutron fluxes in the core and irradiation channels with maximum thermal neutron fluxes at the core center and at the beryllium reflector of about 2.0–3.0 × 1014 and 0.5–1.0 × 1014 n·cm−2·s−1, respectively. Thermal hydraulics analysis has also been performed using the PLTEMP4.2 code to evaluate the safety parameters at a steady state of the MPRR core.
The paper is organized as follows. Section
Figure
The VVR-KN fuel assemblies: (a) standard fuel assembly (FA-1) with eight fuel elements and (b) control fuel assembly (FA-2) with five fuel elements.
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 [
Main design parameters and design targets of the MPRR core.
Technical parameters | Description/value |
---|---|
Reactor type | Open swimming pool |
Thermal power | 10 W |
Coolant and moderator | Water |
Reactor materials | Beryllium |
Core height (mm) | 600 |
Fuel type | VVR-KN |
235U enrichment (wt.%) | 19.75 |
Direction of forced convection | Downward |
Maximum coolant temperature at inlet | 45°C |
Number of horizontal experimental channels | 4–6 |
Number of vertical experimental channels | 15–18 |
For NTD (8-inch ingots) | 3–4 |
For RI (2-3 channels with hydraulic transfer systems) | 10–12 |
For NAA with pneumatic transfer systems | 2–3 |
Number of control rods | 9–12 |
Shim rods (ShR), B4C | 6–9 |
Safety rods (SR), B4C | 2–3 |
Automatic regulating rod (AR), stainless steel | 1 |
Maximum thermal neutron flux at core center (n·cm−2s−1) | 2.0–3.0 × 1014 |
Thermal neutron flux at in-core vertical irradiation channels (n·cm−2s−1) | Up to 1.5 × 1014 |
Thermal neutron flux in beryllium reflector (n·cm−2s−1) | 0.5–1.0 × 1014 |
Thermal neutron flux at irradiation channel for NDT (n·cm−2s−1) | 1.0–2.0 × 1013 |
Thermal neutron flux at horizontal beam tubes (n·cm−2s−1) | 0.5–1.5 × 1013 |
Average discharged burnup | 50% |
Temperature coefficient (%∆ | −0.15 |
Maximum surface temperature of fuel elements | <98°C |
Onset nucleate boiling ratio | >1.3 |
Horizontal cross-sectional view of the MPRR core. HBT: horizontal beam tubes; SiD-1 to SiD4: neutron transmutation doping of silicon; S1-S9: small irradiation channels; B1-B2: big irradiation channels.
The MPRR core map with the hexagonal cell coordinates. SR: safety rod; ShR: shim rod; AR: automatic regulating rod.
The core is controlled by nine control rods: two safety rods (SR1 and SR2), six shim rods (ShR1-ShR6), and an automatic regulating rod (AR). The safety rods are normally withdrawn from the core during the reactor operation and are inserted into the core only in the case of reactor shutdown. The insertion of the shim rods is adjusted during the core operation to maintain criticality. The AR rod is partially inserted into the core during the reactor operation. In the case of emergency shutdown, the control rods fall into the core by the gravity. The safety and shim rods use B4C as a neutron absorbing material with the density of 1.69 g/cm3. The AR rod is made of stainless steel with the density of 7.8 g/cm3 for a smaller reactivity worth.
Neutronic calculations for designing the 10 MW MPRR core have been conducted using the MCNP6 and the ENDF/B-VII.0 nuclear data library [
PLTEMP4.2 is a thermal hydraulics code developed by Los Alamos National Laboratory for analyzing a steady-state flow and temperature solution for a fuel assembly or a reactor core in the subcooled boiling regime [
The first core configuration was established with 26 FAs (17 FA-1 assemblies and 9 FA-2 assemblies) and the power of 6 MW. Burnup calculations were then conducted to establish the next cycle with the increase of power level. After five cycles with the increase of power levels and the number of FAs loaded into the core, a final cycle was obtained. Figure
Configuration of the MPRR core with the power output of 10 MW.
Burnup reactivity of the MPRR core.
Figure
Radial relative power distribution of the MPRR core at the BOC.
Power rates in the fuel elements of the hottest fuel assembly. The fuel elements are numbered from outer to inner elements in the assembly.
Fuel element | FE 1 | FE 2 | FE3 | FE4 | FE5 | FE6 | FE7 | FE8 | Total |
---|---|---|---|---|---|---|---|---|---|
Fuel volume (cm3) | 92.481 | 82.114 | 71.747 | 61.38 | 51.013 | 40.646 | 30.278 | 18.868 | 448.527 |
Power (kW) | 106.21 | 80.28 | 61.79 | 47.67 | 36.69 | 27.84 | 20.24 | 12.78 | 393.49 |
Average power density (W/cm3) | 1148.46 | 977.67 | 861.24 | 776.59 | 719.16 | 684.94 | 668.30 | 677.34 | – |
Axial power profile of the hottest channel in the hottest assembly located at cell 8–4. The fuel elements are numbered from outer to inner elements in the assembly.
Figure
Burnup distribution in percent loss of 235U at the EOC of the MPRR core.
The reactivity worths of the control rods and the shutdown margin have been evaluated and met requirements with condition when all safety CRs were withdrawn from the reactor core. The reactivity worths were calculated with the full insertion of the control rods. Meanwhile, the shutdown margin is defined when all ShR rods are fully inserted, SR1 is completely withdrawn, SR2 is fully inserted, and AR is at the middle of the core. Table
Control rod worths in pcm unit of the MPRR core. Shutdown margin is determined at the condition that all ShR rods are fully inserted, SR1 is completely withdrawn, SR2 is fully inserted, and AR is at center line.
Control rod | pcm |
---|---|
Safety rod SR1 | 2051 |
Safety rod SR2 | 2102 |
Two safety rods | 4629 |
Automatic regulating rod (AR) | 395 |
Shim rod ShR1 | 1156 |
Shim rod ShR2 | 1996 |
Shim rod ShR3 | 2507 |
Shim rod ShR4 | 2056 |
Shim rod ShR5 | 2498 |
Shim rod ShR6 | 1298 |
Shutdown margin | −2330 |
In order to evaluate the safety parameters of the new MPRR core, the feedback reactivity coefficients associated with the change of fuel and coolant temperatures and coolant void fraction at the BOC have been calculated. The temperature of the reflector (water and beryllium) was assumed to be equal to room temperature. To calculate the fuel temperature coefficient (FTC), the fuel temperature was assumed to vary in the range of 294–600 K. Then, the FTC at the BOC is obtained as −2.467 pcm/K. The coolant temperature coefficients (CTC) were evaluated in two ranges of coolant temperature of 294–350 K and 350–400 K, where the linear change of the CTC is valid. It is found that the CTC values are about −11.594 and −12.327 pcm/K in the two ranges of coolant temperatures, respectively. The void reactivity coefficient was also evaluated when the coolant in the active core was voided from 0 to 5%. The void reactivity coefficient at the BOC is obtained as −259.9 pcm per one percent of coolant voided. The results show that the reactivity coefficients of the MPRR core are all negative. These values satisfy the design targets and the safety requirements recommended in IAEA guidelines [
Neutron fluxes in thermal energy range (
Neutron fluxes (n·cm−2·s−1) at irradiation positions in the reactor core and reflector.
Irradiation positions | Thermal | Epi-thermal | Fast |
---|---|---|---|
Neutron trap | |||
Average | 2.02 | 1.48 | 5.89 |
Maximum | 2.54 | 1.82 | 7.20 |
Cold neutron source | |||
Average | 1.23 | 1.59 | 4.02 |
Maximum | 1.57 | 2.07 | 5.12 |
SiD-1 | |||
Average | 1.02 | 6.99 | 1.29 |
Maximum | 1.30 | 9.33 | 1.65 |
SiD-2 | |||
Average | 5.65 | 2.97 | 6.29 |
Maximum | 7.16 | 3.94 | 8.04 |
SiD-3 | |||
Average | 6.16 | 3.04 | 5.25 |
Maximum | 7.82 | 4.06 | 6.70 |
SiD-4 | |||
Average | 1.01 | 6.18 | 1.03 |
Maximum | 1.29 | 8.31 | 1.32 |
B1 | |||
Average | 5.73 | 1.03 | 1.81 |
Maximum | 7.15 | 1.31 | 2.29 |
B2 | |||
Average | 5.94 | 1.12 | 2.02 |
Maximum | 7.60 | 1.46 | 2.58 |
S1 | |||
Average | 6.85 | 1.52 | 2.58 |
Maximum | 8.70 | 1.97 | 3.30 |
S2 | |||
Average | 6.97 | 1.58 | 2.68 |
Maximum | 8.90 | 2.06 | 3.42 |
S3 | |||
Average | 2.76 | 2.35 | 4.87 |
Maximum | 3.53 | 3.15 | 6.20 |
S4 | |||
Average | 2.85 | 2.66 | 5.56 |
Maximum | 3.62 | 3.51 | 7.17 |
S5 | |||
Average | 2.64 | 2.40 | 5.09 |
Maximum | 3.33 | 3.16 | 6.57 |
S6 | |||
Average | 1.55 | 7.99 | 1.40 |
Maximum | 2.08 | 1.27 | 1.85 |
S7 | |||
Average | 1.38 | 7.09 | 1.60 |
Maximum | 1.76 | 9.68 | 2.08 |
S8 | |||
Average | 1.28 | 6.91 | 1.64 |
Maximum | 1.62 | 9.15 | 2.10 |
S9 | |||
Average | 1.19 | 6.43 | 1.46 |
Maximum | 1.50 | 8.58 | 1.89 |
aRead as 2.02 × 1014.
Neutron fluxes (n·cm−2·s−1) at the horizontal beam tubes.
Beam port | At position close to the core | At position near reflector | ||||
---|---|---|---|---|---|---|
Thermal | Epi-thermal | Fast | Thermal | Epi-thermal | Fast | |
HBT1-1 | 7.91 | 2.74 | 3.19 | 1.26 | 9.22 | 6.50 |
HBT1-2 | 7.93 | 2.78 | 3.38 | 1.26 | 9.44 | 7.12 |
HBT2-1 | 4.46 | 1.51 | 1.97 | 1.92 | 1.39 | 9.61 |
HBT2-2 | 4.34 | 1.47 | 1.88 | 1.85 | 1.36 | 9.58 |
HBT3-1 | 1.16 | 6.10 | 1.08 | 9.46 | 1.94 | 3.27 |
HBT3-2 | 1.15 | 5.98 | 1.05 | 9.40 | 1.91 | 3.23 |
aRead as 7.91 × 1013.
Thermal hydraulics analysis at the steady state was performed for the hottest FA using the PLTEMP4.2 code to determine the safety parameters of the newly designed MPRR core. The hottest fuel assembly with the highest power rate of 393.49 kW is located at cell 8–4 as shown in Figure
Figures
Axial profile of coolant temperature in the channels of the hottest fuel assembly. The channels are numbered from outer to inner channels of the FA.
Axial profile of cladding temperature of the FEs in the hottest assembly. The FEs are numbered from outer to inner FEs in the FA.
Conceptual design of a 10 MW MPRR core using the VVR-KN fuel has been conducted using the REBUS-MCNP6 linkage code and the ENDF/B-VII.0 data library. The core is loaded with 36 LEU fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies. The cycle length is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the BOC and the EOC, respectively. The highest discharged burnup of the fuel assembly is 56.74% loss of 235U. The highest power rate per assembly is 393.49 kW. The maximum power density per fuel volume is about 1442.59 W/cm3, which corresponds to a power peaking factor of 2.2. Feedback reactivity coefficients regarding the change of fuel and coolant temperatures and coolant void fraction are all negative. The reactivity worths of the nine control rods ensure the safe operation of the core and provide sufficient shutdown margin. The results show that the neutronics performance of the newly designed MPRR core meets the design targets. Thermal hydraulics analysis at the steady state has been performed for the hottest fuel assembly using the PLTEMP4.2 code. The maximum coolant and cladding temperatures at the hottest channel are 66.0°C and 83.0°C, respectively, without the application of a hotspot factor. The minimum ONBR and DNBR are about 1.80 and 7.02, respectively. These values satisfy the safety limits. For the purposes of application, the neutron fluxes at the vertical and horizontal irradiation channels have been analyzed. The maximum thermal neutron flux of 2.5 × 1014 n·cm−2·s−1 is obtained at the neutron trap located in the core center, whereas this value at the beryllium reflector is about 8.9 × 1013 n·cm−2·s−1.
Data will be made available upon request.
The authors declare that they have no conflicts of interest regarding the publication of this paper.
The authors are grateful to the administrative staffs of VINATOM and DNRI for their kind support. The staffs at Reactor Physics and Engineering Department of DNRI are acknowledged for their valuable technical discussions. Collaborations with the staffs of Argonne National Lab. and Institute of Nuclear Physics of Kazakhstan are greatly appreciated. This research was supported by Ministry of Science and Technology of Vietnam under Grant no. DTDL.CN-50/15.