SRAC Code Research Articles

Design Study of Gas Cooled Fast Reactor (GFR) with Uranium Plutonium Carbide (UC-PuC) as Fuel with Addition Protactinium (Pa-231)

Analysis performance of uranium plutonium carbide (UC-PuC) as fuel in gas cooled fast reactor (GFR) with addition of protactinium as a burnable poisons has been done. Neutronic analysis in this research was carried out using the SRAC code from JAERI with a nuclear library based on JENDL 4.0. The calculation is carried out by two steps, the first step is the PIJ calculation which calculates the fuel cell and the second step is the CITATION calculation which calculates the various configurations of the reactor core. The first calculation determines the k-eff value in a homogeneous core configuration. The results obtained show that the percentage of 10% is the sloping result with a k-eff value of 1%. The second calculation determines the k-eff value in the heterogeneous core configuration. The results obtained indicate that the fuel variation 8% -10% -12% is the most critical percentage with a peak power density value of less than 100 Watt/cc. Furthermore, the addition of protactinium with a variation of 0% to 5%. At a protactinium 4% percentage and 63% fuel fraction, the excess reactivity value is 1.02% or close to 1% which indicates that the reactor is in a critical condition.

Swing Reactivity Calculation of Accelerator Driven System Subcritical Reactor 100 MWth

Accelerator driven system (ADS) applied subcritical reactor which can effectively transmute plutonium, minor actinide, and thorium as nuclear waste. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Accelerator Driven System ( ADS) system lead store duce of the amount of the Pu and Th, so it’s not only can create the addition of energy source, but also high safety reactor. The fuel used in this reactor is obtained from LWR’s spent fuel and thorium - based fuel. The use of thorium-based fuel can lower k-eff of the core and generates U-233, which is a nuclear fuel for critical reactor. Neutron obtained from spallation reaction of Pb-Bi target is placed in the center of the core. This ADS design produces subcritical k-eff during the operation of the reactor and can decrease the amount of fission products of LWR’s spent fuel and produce U-233. The objective of the calculation are criticality as a function of burn up (swing reactivity) in the subcritical assembly fuel. this ADS fuel is originated from UO2 PWR spent fuel with enrichment of 3,4 % and burn up level 64 GWD or 100 MWt for 1825 days calculation of accelerator driven system subcritical reactor 100MWth has been performed. This research d one by using Origen2.1, SRAC2006 and MCNP6 Code. Based on the results swing reactivity of the calculation by using Origen2.1 Code, indicates that Kinf whose value be low subcritical condition (k-eff<1) is Kinf resulted fromTh-232 with percentage from 10% to 5%(as U-238 substitution).the calculation by using SRAC2006 Code indicates that at the beginning of life (BOL), the k-eff of ADS design with 50 cm thickness is 0.9821491 and at the end of life (EOL) k-eff is 0.9396284 and the calculation by using MCNP6 Code. At beginning of life (BOL), the k-eff of ADS design is 0.9847 and the end of life (EOL) is 0.9823.The presence of plutonium and mayor actinides makes k-eff get slightly lower. Arrangement at this ADS model is performed to produce subcritical condition, in which an increase in fuel volume cause k-eff>I.

Study on the Effects of Enrichment and Fraction of Coated Fuel Particles on Fissile Utilization of 100 MWt Prismatic-type of High Temperature Gas Reactor

High Temperature Gas-cooled Reactor (HTGR) is one of Generation-IV reactor technology that has graphite moderate, Helium gas-cooled, and Coated Fuel Particle (CFP) layered by Tristructural-Isotropic (TRISO). The HTGR has an outlet temperature of around 1000°C that can be utilized for many co-generation processes other than to generate electricity. Due to characteristics of the CFP and TRISO, utilization of fissile material during reactor operation becomes important. This study aims to analyze the effects of enrichment and fraction of CFP on the fissile utilization of 100 MWt HTGR for two different fuels; UO2 and (Th-U)O2. Fissile enrichment is analyzed from 1-20% while the fraction of CFP from 10-60%. Using SRAC2006 code and JENDL-4.0 as nuclear data library, calculation and analysis are performed to find the optimal values of several important neutronic parameters.

Conceptual Analysis of (Th,U-233)O2 Fueled Small Long-Life PWR with Np-237 and Pa-231 as Burnable Poison

Conceptual analysis of (Th,U-233)O2 fueled small long-life PWR with Np-237 and Pa-231 as burnable poison has been performed. The fuel used is (Th,U-233)O2 with Th-232 and U-233 as fertile-fissile nuclides pair with 8% U-233 and addition of Pa-231 and Np-237 as burnable poison (BP). This study aims to investigate optimum design of fuel and core configuration and the performance of reactor. The calculation is done by SRAC2006 code developed by JAEA and is based on JENDL-4.0 library. Cell calculation conducted by PIJ module shows that optimum design is reached at 60% of fuel volume fraction and 6% of burnable poison. Core optimization calculation by CITATION module for two dimensional (R – Z) core geometry shows that reactor could reach 60 years of lifetime at 150 MWt power level which correspond to 25.36 power density. Calculation for heterogen reactor shows that reactor with BP combination of 8% of Pa-231 at center zone and 6% of Pa-231 at outer zone has better performance with 60 years of operation period than reactor with combination of 8% of Np-237 at center zone and 6% of Pa-231 at outer zone that reaches 46 years lifetime.

Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor

This study has made a calculation of Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor. The purpose of this study was to determine the characteristics of minor actinides when incorporated into plutonium nitride uranium fuel. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some steps of parameter survey to obtain the ultimate optimization results. The initial calculation was calculation of fuel cell (PIJ calculation) by using hexagonal cell and then followed by calculation of core reactor (CITATION calculation) by using program code SRAC2006. The addition of minor actinide has been calculated in this research. The minor actinides which were added were americium (Am-241 and Am-243) and neptunium 237. They were put into the uranium plutonium nitride material to decrease the k-eff value. The addition of minor actinide into the reactor aimed to reduce the number of minor actinides in the world. Minor actinide is a nuclear waste fuel or often called spent nuclear fuel (SNF) which has high toxicity. Neptunium 237 is a minor actinide with the highest percentage of SNF.

Neutronic Analysis of Small Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

The requirement for electricity increases with the growth of the human population. The existing power plants have not been able to fulfill all electricity requirements, especially in remote areas. The small long-life pressurized water reactor (PWR) is one of the solutions and innovations in nuclear technology that can produce electrical energy for a long time without refueling. This study aimed to analyze the neutronic of small long-life PWR that using Thorium-Uranium dioxide ((Th-U)O2) fuels with enriched Uranium-235 (U-235) and the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. The SRAC Code with the JENDL-4.0 nuclear data library had been used for the calculation method. In this study, the geometry of the two-dimensional (R-Z) reactor core with different fuel volume fraction was analyzed. Moreover, variations of the Uranium-235, Gadolinium, and Protactinium-231 fractions in the fuels were carried out. The result in this study was a PWR 420 MWt design using 60% Uranium dioxide fuel with enriched Uranium-235 of 10%-11%-12% and the addition of 0,0125% Gadolinium and 1,0% Protactinium-231 as the burnable poisons that could operate for thirteen years without refueling. The small long-life PWR design could produce a power density of 85,1 watts/cc with the reactivity for less than 4,6% dk/k.

Study on Transmutation of Minor Actinides as Burnable Poison in VVER-1000 Fuel Assembly

Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feasibility study on transmutation of minor actinides as burnable poison in the VVER-1000 LEU (low enriched uranium) fuel assembly. The VVER-1000 LEU fuel assembly was modeled using the SRAC code system, and the SRAC calculation model was verified against the MCNP6 calculations and the available published benchmark data. Two models of minor actinide loading in the LEU fuel assembly have been investigated: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The consequent negative reactivity insertion by minor actinides was compensated by reducing the gadolinium content and boron concentration. The reactivity of the LEU assembly versus burnup and the transmutation of minor actinide nuclides were examined in comparison with the reference case. The results demonstrate that transmutation of minor actinides as burnable poison in the VVER-1000 reactor is feasible as minor actinides could partially replace the functions of gadolinium and boric acid for excess reactivity control.

Conceptual design of a small-pressurized water reactor using the AP1000 fuel assembly design

This paper presents the conceptual design of a 300 MWt small modular reactor (SMR)using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 data library. The analysis showed that Doppler, moderator temperature, void, and power reactivity coefficients are all negative over the core lifetime. Semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel andclad surface temperature. The minimum departure from nucleate boiling ratio (MDNBR) is also calculated. The results indicate that a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

Homogenous (Th,Pu)O2 Fuel Utilization on High Temperature Test Reactor 30MWt

HTTR (High Temperature Test Reactor) is one of the Generation IV nuclear reactor that has been built in Japan. This study will discuss about several neutronic aspect on HTTR 30 MWt such as effective multiplication factor (k-eff) whole of active core, conversion ratio, spectrum of neutrons and the power distribution of the reactor core at the beginning and at the end of the period by using homogenous (Th,Pu)O2 fuel assembly. 3D triangular geometrical type of active core has been selected for neutronic calculation by using SRAC2006 code with nuclear data bases JENDL4.0 which is developed by JAERI. The results shows that core optimum when (Th,Pu)O2 with PuO2 fraction is 6,7% and will give the optimum k-eff during the operation time. The spectrum neutron result shows that (Th,Pu)O2 is dominant at fast energy. The power distribution results show that it can be stable for long operation period, both at beginning of life (BOL) and end of life (EOL).

HTTR 30MWth Reactor with Homogenous (Th,U)O2 Fuel

(Th,U)O2 fuel has been utilized for neutronic analysis of HTTR 30 MWt which is a high-temperature test reactor with helium gas as a coolant that has been built in Japan. HTTR has a Triple-isotropic Coated Fuel Particle (TRISO CFP) type of fuel. Homogenous fuel of (Th,U)O2 in the active core reactor give some different results for neutronic analysis such as effective multiplication factor (k-eff) each block fuel and whole of active core, conversion ratio, and spectrum of neutrons. The calculation was calculated by using neutronic SRAC2006 code and based on JENDL4.0 nuclear data which has been developed by JAERI. The results show that the optimum core may be obtained when (Th,U)O2 has 60% of ThO2 and 16.8% enrichment of UO2. The spectrum neutron shows that (Th,U)O2 fuel results in a softer spertrum.

Neutronic Comparison Study Between Pb(208)-Bi and Pb(208) as a Coolant In The Fast Reactor With Modified CANDLE Burn up Scheme.

Neutronic study of Pb(208)-Bismuth as a coolant in the Lead Fast Reactor (LFR) with Modified CANDLE burn up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with 500 MW thermal power output. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled with fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The neutronic calculations were performed by SRAC code using nuclear data library based on JENDL 4.0. Level burn up of Pb(208)-Bi cooled fast reactor is 530.688 GWD/MTU at BOC and 433.051 GWD/MTU at EOC whereas 190.790 GWD/MTU at BOC and 433.051 GWD/MTU at EOC for Pb(208) . The effective multiplication factor of Pb(208)-Bi Cooled Fast Reactor is 1.0554 at BOC and 1.05958 at EOC whereas 1.06703 at BOC and 1.06816 at EOC for Pb208.

Design Study of 600 MWt Long Life Modular Gas Cooled Fast Reactors

Design study of long life 600 MWt gas cooled fast reactors which can be operated for 25 years without refueling or fuel shuffling has been performed. A small Gas-cooled Fast Reactors (GFRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN) have been calculation by SRAC code system and FI-ITB CH1 code. The results shows that it has been obtained some designs which can be operated for 25 years with relatively excess reactivity. The maximum excess reactivity is 2% dk/k. The reactor core is divided into three regions, the first regon is filled with highest content of plutonium and located in the most outer region. The second part of the core is filed with second highest plutonium content and located in the inner part of first region. The third region is filled wiith the lowest content of plutonium and located in the central part of the core. At the begnning of life the first reactor is the most dominant, but as time proceeds the second and the third regions of the core become more important. The average burn-up is around 10% HM

Reactivity feedback effect on loss of flow accident in PWR

Abstract In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Neutron Radiation Damage Estimation in the Core Structure Base Metal of RSG GAS

Radiation damage in core structure of the Indonesian RGS GAS multi purpose reactor resulting from the reaction of fast and thermal neutrons with core material structure was investigated for the first time after almost 30 years in operation. The aim is to analyze the degradation level of the critical components of the RSG GAS reactor so that the remaining life of its component can be estimated. Evaluation results of critical components remaining life will be used as data ccompleteness for submission of reactor operating permit extension. Material damage analysis due to neutron radiation is performed for the core structure components made of AlMg3 material and bolts reinforcement of core structure made of SUS304. Material damage evaluation was done on Al and Fe as base metal of AlMg3 and SUS304, respectively. Neutron fluences are evaluated based on the assumption that neutron flux calculations of U3Si8-Al equilibrium core which is operated on power rated of 15 MW. Calculation result using SRAC2006 code of CITATION module shows the maximum total neutron flux and flux >0.1 MeV are 2.537E+14 n/cm2/s and 3.376E+13 n/cm2/s, respectively. It was located at CIP core center close to the fuel element. After operating up to the end of #89 core formation, the total neutron fluence and fluence >0.1 MeV were achieved 9.063E+22 and 1.269E+22 n/cm2, respectively. Those are related to material damage of Al and Fe as much as 17.91 and 10.06 dpa, respectively. Referring to the life time of Al-1100 material irradiated in the neutron field with thermal flux/total flux=1.7 which capable of accepting material damage up to 250 dpa, it was concluded that RSG GAS reactor core structure underwent 7.16% of its operating life span. It means that core structure of RSG GAS reactor is still capable to receive the total neutron fluence of 9.637E+22 n/cm2 or fluence >0.1 MeV of 5.672E+22 n/cm2.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Fuel Loading Pattern Optimization with Constraint on Fuel Assembly Inventory Using Quantum-Inspired Evolutionary Algorithm

Nuclear fuel management was done by optimizing fuel loading pattern in a reactor core. Practically, performing fuel loading pattern optimization was difficult because of its combinatorial problem complexity which needed to be solved. Therefore, Quantum-inspired Evolutionary Algorithm (QEA) which could solve the combinatorial problem faster than conventional method was used. The main purpose of this research was to obtain an optimum fuel loading pattern of KSNP-1000 reactor core without altering fuel assembly inventories. KSNP-1000 core was modeled in SRAC code package using PIJ module for fuel pins and fuel assemblies’ lattices and CITATION module for fuel assemblies’ pattern in a quarter core symmetry. Optimization problem adaptation using QEA was made by presenting 52 fuel assemblies in Q-bit individuals with the length of 8 Q-bits. Q-bits were converted to corresponding bit values and then given weight which would be used as consideration to optimize the pattern. The optimization program was coupled with the SRAC neutronic code to obtain the values of effective multiplication factor (keff) and power peaking factor (PPF). The optimization was calculated based on fitness value which was a function of keff and PPF values with the particular weight factor. Using a rotation gate angle of Δθ=0.02π and a weight factor of w=0,041, fuel loading pattern optimization was done on 360 days burnup level. The optimization resulted in keff and PPF value of 1.11233 and 1.944 respectively. By calculating keff value on various burnup levels for the chosen core loading pattern, reactor cycle length obtained was 659 days with PPF at BOC was 2.19. Compared to the standard KSNP-1000 core which had 560 days of cycle length, the optimized core configuration increased 17.67% in cycle length.

Comparison of uranium plutonium nitride (U-Pu-N) and thorium nitride (Th-N) fuel for 500 MWth gas-cooled fast reactor (GFR) long life without refueling

Summary Comparison of uranium plutonium nitride and thorium nitride fuel for 500 MWth gas-cooled fast reactor has been done. Gas-cooled fast reactor is one type of generation IV reactor that can be operated in high temperature. Due to the high temperature, it can be used in hydrogen production. In this study, we compare the neutronic analysis of two fuel types, ie, uranium nitride fuel (U,Pu)N and thorium nitride fuel (Th,U233)N. The neutronic calculation uses SRAC2006 code system, and the data libraries use JENDL4.0. First, the fuel pin calculation (PIJ calculation) has been done to take the macro data that are used in CITATION calculation. Both uranium and thorium fuel use heterogeneous configuration with 3 variation fuel in the core. F1 is located in the central core, F2 middle core, and F3 outer core. The variation of fuel fraction is 40% until 65%, cladding 10%, and coolant 25% until 40%. For (U,Pu)N fuel, the diameter of the active core is 220 cm, and the height of the active core is 110 cm. And for (Th,U233)N, the diameter of the active core is 250 cm and the height of the active core is 150 cm. The reflector radial-axial width is 50 cm. For uranium plutonium nitride fuel, the type of the fuel in one core is varied; ie, F1 is 8%, F2 is 10%, and F3 is 12%. For thorium nitride fuel, the type of the fuel in one core also varied; ie, F1 is 7.8%, F2 is 8%, and F3 is 8.8%. The optimum value of thorium nitride is when fuel fraction of region F1 = 60%, F2 = 57.5%, F3 = 60%, the burn up time up to 20 years without refueling, max k-eff value is 1.0109, and max excess reactivity value is 1.08%. Neutronic analysis shows that both uranium and thorium fuel have excess reactivity value less than 2% but thorium fuel has excess reactivity less than uranium nitride fuel. Uranium fuel has better breeding capability than thorium fuel. Therefore, it is better to use uranium fuel for fast reactor like GFR, which has high breeding capability.

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D<H). The reflector radial-axial width is 50 cm. The power is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

Automated generation of burnup chain for reactor analysis applications

Abstract This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.