Effective Full Power Days Research Articles

Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

In this work, we present results of fuel depletion analyses performed for a potential LEU core of Ghana’s Miniature Neutron Source Reactor (GHARR-1) using the Monte Carlo N-particle extended (MCNPX) neutron transport code. Depletion calculation was carried out for the reactor core from the Beginning of Life (BOL) to the End of Life (EOL) which corresponds to 10years of reactor operation. The amounts of uranium and plutonium actinides were estimated at BOL and EOL of the core. Decay heat removal rate for the MNSR after reactor shutdown was investigated due to its significance to reactor safety. Inventory of fission products produced as a result of burnup was also calculated. The results show that a maximum discharge burnup equivalent to 0.568% of U-235 was consumed at EOL equivalent to operating the reactor for 200 Effective Full Power Days (EFPD), while the amount of Pu-239 produced was not significant. Also, the decay heat decreased exponentially after reactor shutdown confirming that decay heat will be removed in the system by natural circulation after shutdown and thus guaranteeing the safety of the reactor.

Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

IRRADIATION DEVICE FOR IRRADIATION TESTING OF COATED PARTICLE FUEL AT HANARO

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about 1.7×10 21 n/cm 2 , PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part I: Lattice benchmarking, cycle length, and reactivity coefficients

The fully ceramic microencapsulated (FCM) fuel concept is based on the tri-isotropic (TRISO) carbon coated fuel particles. These particles were developed and demonstrated for use in high temperature gas reactors. It has been proposed to use these particles in light water reactors to provide potential operational and safety benefits. The reference fuel in this case assumes TRISO-like particles with a ∼20%-enriched uranium-nitride kernel embedded in a silicon carbide (SiC) matrix. The fuel particles are contained in a “compact” which is then inserted into a cladding. The fuel assembly features the same dimensions as a standard 17×17 Westinghouse fuel assembly. FCM fuel requires fission products to traverse several barriers in the proposed fuel design before reaching the cladding. FCM fuel may also reduce fuel-cladding interaction and fuel pellet swelling while enabling higher fuel burn-up. This study is a neutronic evaluation of the use of FCM fuel in an advanced pressurized water reactor (PWR). On the lattice level, the SERPENT Monte Carlo and TRITON deterministic tools were used, while the whole core simulation was based on the three-dimensional PARCS nodal code.This paper presents the results of the lattice-level neutronic study of doubly heterogeneous FCM fuel. Strong agreement was found between the SERPENT and TRITON codes in terms of k-infinity as a function of burn-up, actinide build-up, and “pin” powers. The impact of several simplifying geometric assumptions was considered, such as the use of a square particle lattice within the FCM fuel pins. It was determined that the linear reactivity model does not provide a good estimate of the fuel cycle length, due primarily to non-linear reactivity behavior at high burn-up (>800 effective full power days). To determine cycle length, higher order reactivity models were applied to the lattice results. The calculated cycle lengths are slightly reduced versus a reference uranium oxide case. Finally, the assembly-level reactivity coefficients were calculated as a function of burn-up. The fuel and moderator temperature coefficients were negative for FCM fuel, but reduced in magnitude by approximately 50% versus a reference uranium oxide case.

Distribution characteristics of fission gas along the axial direction for an irradiated fuel rod of a pressurized water reactor (PWR)

In this work, the fission gas distribution of an irradiated oxide fuel was investigated by measuring the fission gas release (FGR) and retained gas from and in an oxide fuel along the axial height of the fuel rod. In addition to the fission gas measurements, other destructive and non-destructive post irradiation examination tests such as gamma scanning, eddy current testing, density and oxide layer thickness measurements, and ceramography/metallography were conducted to review their impact on the FGR and gas retention. A lead rod was extracted from a lead test assembly, which was irradiated to 56.9 GWd/tU during three irradiation cycles for a total exposure of 1,406 effective full power days. Considering the water-side oxide layer thickness and local burnup of the fuel rod, the rod was sectioned into four positions along the axial height and four samples were prepared from these positions. A sample fragment of around 0.1–0.2 g was individually melted to measure its retained krypton and xenon concentration. According to the measurement results, the retained krypton and xenon concentration ranges were 0.114–0.139 cc/gU and 1.073–1.338 cc/gU, respectively, and their retention percentage after normalization of the sample local burnup showed a decreasing trend as the axial height of the fuel rod raised. The water-side oxide layer thickness scope of the tested fuels measured by an observation of the optical microscope images was 13–42 μm, and the thickness was increased along the axial height of the rod. The amount of released krypton and xenon into the rod-free volume was measured as 5.7 and 54.6 cc, respectively, by a rod puncturing/collection procedure, which corresponds to a 1.94 % fractional fission gas release referring to fission gas generation by a code calculation.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs) is made along with a design change in CSRs to increase the reactivity worth of absorber rods for ensuring safe reactor operation. This new enhanced burnup core can be used in the future with advanced structural materials for clad and hexcan, which can withstand higher neutron fluence and gamma radiation.

A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 (237Np) or americium-241 (241Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237Np and 241Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237Np and 241Am, CSDT was short by approximately 40 and 50years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses.

PREDICTION OF DIAMETRAL CREEP FOR PRESSURE TUBES OF A PRESSURIZED HEAVY WATER REACTOR USING DATA BASED MODELING

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

APPLICATION OF A GENETIC ALGORITHM FOR THE OPTIMIZATION OF ENRICHMENT ZONING AND GADOLINIA FUEL (UO2/Gd2O3) ROD DESIGNS IN OPR1000s

A new effective methodology for optimizing the enrichment of low-enriched zones as well as gadolinia fuel (<TEX>$UO_2/Gd_2O_3$</TEX>) rod designs in PLUS7 fuel assemblies was developed to minimize the maximum peak power in the core and to maximize the cycle lifetime. An automated link code was developed to integrate the genetic algorithm (GA) and the core design code package of ALPHA/PHOENIX-P/ANC and to generate and evaluate the candidates to be optimized efficiently through the integrated code package. This study introduces an optimization technique for the optimization of gadolinia fuel rod designs in order to effectively reduce the peak powers for a few hot assemblies simultaneously during the cycle. Coupled with the gadolinia optimization, the optimum enrichments were determined using the same automated code package. Applying this technique to the reference core of Ulchin Unit 4 Cycle 11, the gadolinia fuel rods in each hot assembly were optimized to different numbers and positions from their original designs, and the maximum peak power was decreased by 2.5%, while the independent optimization technique showed a decrease of 1.6% for the same fuel assembly. The lower enrichments at the fuel rods adjacent to the corner gap (CG), guide tube (GT), and instrumentation tube (IT) were optimized from the current 4.1, 4.1, 4.1 w/o to 4.65, 4.2, 4.2 w/o. The increase in the cycle lifetime achieved through this methodology was 5 effective full-power days (EFPD) on an ideal equilibrium cycle basis while keeping the peak power as low as 2.3% compared with the original design.

Long-life fast breeder reactor with highly protected Pu breeding by introducing axial inner blanket and minor actinides

A feasibility study on simultaneous approaches to the extension of core life-time and the high protected plutonium (Pu) breeding by introducing the axial inner blanket and doping minor actinides (MAs) in a large-scale sodium-cooling fast breeder reactor (FBR) has been performed for mix-oxide (MOX) and metallic fuel. The maximum available effective full power days (EFPDs) in MOX-fueled FBR with introducing the axial inner blanket and MA was extended from 1700 to 2900 compared with the conventional MOX-fueled FBR. The maximum available EFPDs in the case of metallic-fueled FBR with introducing the axial inner blanket and MA was extended to 5900. Conventional Attractiveness (ATTR) to evaluate the proliferation resistance of Pu based on isotopic material barriers such as decay heat (DH) and spontaneous fission neutron rate (SN) was modified by taking into account bare critical mass (BCM) as Attractivenessmod (ATTRmod), which was applied to evaluate the proliferation resistance of Pu generated in the axial inner blanket and axial/radial outer blankets. It was found that if only a small amount (4–5wt.%) of MA are doped into the axial inner blanket and axial/radial outer blanket in both MOX and metallic fuel, the proliferation resistance of Pu was increased to satisfy the criteria of “practically unusable for an explosive device” proposed by Pellaud and “technically unfeasible for a high-technology hypothetical nuclear explosive devices (HNEDs)” proposed by Kessler and Kimura. The fissionable Pu inventory ratio (FPIR), defined as the ratio of inventory of 238Pu, 239Pu and 241Pu at the beginning of cycle (BOC) to that at the given irradiation time, was introduced to survey the breeding characteristics of FBR. FPIR at the end of cycle (EOC) in MOX-fueled FBR with doping MA into the axial inner blanket and axial/radial outer blankets was increased from 1.008 to 1.168 compare with the conventional MOX-fueled FBR. In addition, FPIR at EOC in metallic-fueled FBR with doping MA into the axial inner blanket and axial/radial outer blankets was increased to 1.263. In the present study, the feasibility of simultaneous approaches to the extension of core life-time and the high protected Pu breeding has been confirmed by the introduction of the axial inner blanket and MA doping.

TRU self-recycling in a high temperature modular helium reactor

Abstract The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again. In this study, a single recycling of the self-generated TRUs is considered. The UO 2 fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO 2 fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO 2 fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO 2 fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO 2 -only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).

A fuel performance analysis for a 450 MW th deep burn-high temperature reactor

A performance analysis for a 450 MW th deep burn-high temperature reactor (DB-HTR) fuel was performed using COPA, a fuel performance analysis code of Korea Atomic Energy Research Institute (KAERI). The code computes gas pressure buildup in the void volume of a tri-isotropic coated fuel particle (TRISO), temperature distribution in a DB-HTR fuel, thermo-mechanical stress in a coated fuel particle (CFP), failure fractions of a batch of CFPs, and fission product (FP) releases into the coolant. The 350 μm DB-HTR kernel is composed of 30% UO 2 + 70% (5% NpO 2 + 95% PuO 1.8) mixed with 0.6 moles of silicon carbide (SiC) per mole of heavy metal. The DB-HTR is operated at the constant temperature and power of 858 °C and 39.02 mW per CFP for 1395 effective full power days (EFPD) and is subjected to a core heat-up event for 250 h during which the maximum coolant temperature reaches 1548.70 °C. Within the normal operating temperature, the fuel showed good thermal and mechanical integrity. At elevated temperatures of the accident event, the failure fraction of CFPs resulted from the mechanical failure (MF) and the thermal decomposition (TD) of the SiC barrier is 3.30 × 10 −3.

Swelling under irradiation of MgO pellets containing americium oxide: The ECRIX-H irradiation experiment

The ECRIX-H irradiation experiment studied the behaviour of pellets containing americium dispersed in MgO. The purpose of the irradiation was to demonstrate the capacity of magnesia to provide an efficient support matrix. After fabrication, the sintered pellets contained 16.65 wt.% of Am microdispersed in the inert matrix. The ECRIX-H pellets were irradiated under a locally moderated neutron flux in the Phénix sodium-cooled fast reactor (SFR) for 318 Effective Full Power Days (EFPD). Post-test calculations indicated that the fission and transmutation rates of americium at the maximum flux plane reached 33.9% and 92.6% respectively at the end of the irradiation phase. The results of the post-irradiation examinations – both non-destructive and destructive – are discussed in this paper. These results indicate a satisfactory behaviour of the MgO matrix. Particularly, a moderate swelling occurs in the pellets under irradiation even with significant quantities of helium generated and at high transmutation rate.

Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

Approach to equilibrium fuelling scheme of 500 MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on fuel and blanket pins it is found that the nominal core achieves full power only in mid-cycle. A novel interpolation scheme is used to find the peak LHR in any axial section of a fuel/blanket sub-assembly. Breeding ratio is adequate for self-sufficient Pu generation in a closed fuel cycle with Pu from axial blankets and two rings of radial blanket sub-assemblies.

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.

Irradiation Test of MOX Fuel Rods Fabricated by Attrition-Milling and Analysis of In-Pile Data with COSMOS Code

Attrition-milling technology for fabricating mixed oxide (MOX) fuel was developed to mix the plutonium in UO2 fuels as homogeneously as possible. The fabricated MOX fuels were instrumented with temperature and pressure gauges that enabled one to measure the fuel temperature and rod internal pressure online. An irradiation test in the Halden reactor was performed to investigate the in-pile behavior of the fabricated MOX fuel. The irradiation of 1020 effective full-power days was successfully accomplished with good integrity of the test fuel rods. The rod average burnup reached ~50 MWd/kg HM, and the measured fuel centerline temperature was ~1000°C for the MOX fuels. A significant fission gas release was observed due to the high power level. The online measured in-pile performance data of the two attrition-milled MOX fuel rods were analyzed and compared with the fuel performance code COSMOS. COSMOS simulated the fuel centerline temperature and rod internal pressure for both MOX fuel rods. The analysis by COSMOS showed good agreement with the online measured in-pile behavior of MOX fuel.

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50–80% relative to the current design, with only a modest increase in the fissile material loading (15–20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once – through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10–15% over that of current MHR design.

Criticality and burnup analyses of a PBMR-400 full core using Monte Carlo calculation method

In this study, the criticality and burnup analyses have been performed for full core model of Pebble Bed Modular Reactors, such as PBMR-400, using the computer codes MCNP5.1.4 and MONTEBURNS 2.0. Three different pebble distributions, namely; Body Centered Cubic (BCC) (packing fraction = 68%), Random Packing (RP) (packing fraction = 61%) and Simple Cubic (SC) (packing fraction = 52%) were selected for the analyses. The calculated core effective multiplication factor, k eff, for BCC, RP and SC came to be 1.2395, 1.2357 and 1.2223, respectively. The core life for these distributions were calculated as ∼1200, 1000, and 800 Effective Full Power Days (EFPDs), whereas, the corresponding burnups came out to be ∼99,000, ∼92,000 and ∼86,000 MWD/T, respectively, for end of life k eff set equal to 1.02.

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k eff ≤ 0.97 and produces less fission power of 1.5 GW th. Its main performance parameters turn out substantially better.

Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li 6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.