Conceptual core design of Advanced Recycling Reactor based on mature technologies

This paper presents about comprehensive investigations into Advanced Recycling Reactor (ARR) based on existing and/or mature technologies (called “Early ARR”), aiming transuranics (TRU) burning and considering harmonization of TRU burning capability, technology readiness, economy and safety. The ARR is a 500 MW_e (1180 MW_th) oxide fueled sodium cooled fast reactor, which the low core height of 70 cm and the large structure volume fraction with 1.0 mm of cladding thickness to tube wall have been chosen among 14 candidate concepts to reduce the TRU conversion ratio (CR) and the void reactivity, taking technology readiness into account. As a result of nuclear calculation, the ARR has TRU burning capability from 19 to 21 kg/TW_thh and is sustainable in recycling. And the ARR can accept several kinds of TRU; the LWR uranium oxide fuels, LWR-MOX used nuclear fuel, and TRU recycled in this fuel cycle and the ARR is also flexible in TRU management in ways that it can transform from TRU CR of 0.56 to breeding ratio (BR) of 1.03. In addition, it has been confirmed that the ARR core conforms to the set design requirements; the void reactivity, the maximum linear heat rate, and the shutdown margin of reactivity control system. It has been confirmed that the closed fuel cycle with the ARR plants of 180 GW_th will not release TRU outside and generate more electricity by 65% compared with the present nuclear power system in the US, curbing the risk of nuclear proliferation. Thus the study can conclude that the Early ARR is able to close nuclear fuel cycle, using mature technologies and has features of the sustainability in recycling, and the accommodation of almost all the TRU at present and in the future, and the flexibility in TRU management with breakeven core.     

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR)^* and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).     

Design and analysis of HYPER

KAERI (Korea Atomic Energy Research Institute) has been developing an accelerator driven transmutation system called HYPER (hybrid power extraction reactor). It is designed to transmute long-lived TRU and fission products such as Tc-99 and I-129. HYPER is a 1000 MW_th system with k_eff = 0.98 which requires 17 mA proton beam for an operation at EOC (end of cycle). Pb–Bi is used as the coolant and target material at the same time. HYPER core has 186 ductless hexagonal fuel assemblies. The fuel blanket is divided into three TRU (transuranic elements) enrichment zones to flatten the radial power distribution. The core height of HYPER was compromised at 150 cm, and the power density was determined such that the average coolant speed could be about 1.64 m/s. The inlet and exit coolant temperatures are 340 and 490 °C, respectively, in the core. The cylindrical beam tube and spherical window is adopted as the basic window design of HYPER. We have also introduced an Lead–Bismuth eutectic injection tube to maximize the allowable proton beam current. A metallic alloy of U-TRU-Zr is considered as the HYPER fuel, in which pure lead is used as the bonding material. As a result, a large gas plenum is placed above the active core. TRU transmutation rate is 282 kg/yr. In the case of a FP transmutation, 28.0 kg of Tc-99 and 7.0 kg of I-129 are incinerated per year. The MACSIS-H (metal fuel performance analysis code for simulating the in-reactor behavior under steady-state conditions-HYPER) for an metallic fuel was developed as the steady-state performance computer code. The MATRA (multichannel analyzer for transient and steady-state in rod array) code was used to perform the thermal-hydraulic analysis of HYPER core.     

SFR with once-through depleted uranium breed & burn blanket

This paper assesses the feasibility of Sodium-cooled Fast Reactor (SFR) cores that have TRU recycled seeds and once-through depleted uranium blankets. The design objective of these Seed-and-Blanket (S&B) cores is to maximize the power generated by the blanket. As the blanket fuel cost is significantly lower than the cost of the seed fuel and does not need reprocessing, increasing the fraction of reactor power generated by the blanket will reduce the total fuel cycle cost and the fuel reprocessing capacity required per unit of electricity generated. The S&B core is designed to have a prolate (“cigar”) shape seed (“driver”) to maximize the fraction of neutrons that radially leak into the subcritical blanket and reduce neutron loss via axial leakage. Both seed and blanket contain multiple batches; the blanket batches are gradually shuffled inward, while one third of the fuel batches in the seed are recycled. The preliminary study found that it is possible to design the seed to accommodate a wide range of TRU conversion ratios (CR) without significantly penalizing the burnup reactivity swing. The relatively small burnup reactivity swing enables to design the S&B core to operate at longer cycles and discharge its fuel at a higher burnup relative to conventional TRU transmutation cores with identical CR. The S&B cores can generate 1000 MW_th and fit within the S-PRISM reactor vessel. The fraction of core power generated by the blanket is between 40% and 50% without exceeding the radiation damage constraint of 200 Displacements per Atom (DPA); this fraction increases when the seed is designed to have a smaller CR. These features are expected to improve the economics of SFR.     

TRU burning by dual tier system of LWR-SFR

This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10⁻³ Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.     

Transmutation-incineration potential of transuraniums discharged from PWR-UO2 spent fuel in modified PROMETHEUS fusion reactor

This study presents the potential of the burning and/or transmutation (B/T) of transuraniums (TRUs), discharged from the pressured water reactor PWR-UO₂ spent fuel, in the modified PROMETHEUS-H fusion reactor. Two different design shapes (Models A and B) were considered. The transmutation zone (TZ), which contains the mixture of TRU nuclides (10%), was located in the modified blankets. The volume fraction of Pu in the mixture is raised from 0 to 40% stepped by 10% to determine its effect on the B/T. The fuel spheres were cladded with SiC (5%) and cooled with high-pressured helium gas (85%) for nuclear heat transfer. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m². The results bring out that: (1) the Model B transmutes the TRUs more rapidly than the Model A; (2) the effective half-lives decrease about 20 and 40% with the increase of Pu fraction in the cases of Models A and B, respectively; (3) the M values are quite high with respect to the M value of the original PROMETHEUS fusion reactor; (4) the blankets can produce substantial electricity in situ.     

The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe – Part II: Thorium fuel cycle

Full recycling of transuranic (TRU) isotopes can in theory lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited timeframe, the radiotoxicity of the ‘final’ core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In Part I of this paper, TRU recycle over up to 5 generations of light water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. U-fuelled SFRs operating a break-even fuel cycle are not particularly effective at reducing repository radiotoxicity as the final core load dominates over a very long timeframe. In this paper, the analysis is extended to the thorium (Th) fuel cycle. Closed Th-based fuel cycles are well known to have lower equilibrium radiotoxicity than U-based fuel cycles but the time taken to reach equilibrium is generally very long. Th burner fuel cycles with SFRs are found to result in very similar radiotoxicity to U burner fuel cycles with SFRs for one less generation of reactors, provided that protactinium (Pa) is recycled. Th-fuelled reduced-moderation boiling water reactors (RBWRs) are also considered, but for burner fuel cycles their performance is substantially worse, with the waste taking ∼3–5 times longer to decay to the reference level than for Th-fuelled SFRs with the same number of generations. Th break-even fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years. While this is a very long timeframe, it is roughly half that required for waste from the Th or U burner fuel cycle to decay to the reference level, and less than a tenth that required for the U break-even fuel cycle. The improved performance over burner fuel cycles is due to a more substantial contribution of energy generated by ²³³U leading to lower radiotoxicity per unit energy generation. To some extent this an argument based on how the radiotoxicity is normalised: operating a break-even fuel cycle rather than phasing out nuclear power using a burner fuel cycle results in higher repository radiotoxicity in absolute terms. The advantage of Th break-even fuel cycles is also contingent on recycling Pa, and reprocessing losses are significant also for a small number of generations due to the need to effectively burn down the TRU. The integrated decay heat over the scenario timeframe is almost twice as high for a break-even Th fuel cycle than a break-even U fuel cycle when using SFRs, as a result of much higher ⁹⁰Sr production, which subsequently decays into ⁹⁰Y. The peak decay heat is comparable. As decay heat at vitrification and repository decay heat affect repository sizing, this may weaken the argument for the Th cycle.     

Advances in the conceptual design of a gas-cooled Accelerator Driven System (ADS) transmutation device for a sustainable nuclear energy development

The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been done in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In the paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made.Additionally, the heterogeneity of the TRISO particles inside the pebbles is not negligible. In the paper, a study of the power density distribution inside the pebbles using a detailed model of the TRISO particles and a homogeneous composition of the fuel is addressed.     

Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor

This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H fusion reactor. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor () under a neutron wall load (P) of 4.7 MW/m². In order to incinerate and transmute the MAs effectively, the transmutation zone (TZ) containing the mixture of MA nuclides, discharged from the pressured water reactor (PWR)-MOX spent fuel, was located in the modified blanket of the PROMETHEUS-H fusion reactor. Two different blanket modifications were considered, (the Models A and B). The MA mixture was spherically prepared and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium was selected for the nuclear heat transfer in the TZ. The effect of the MA volume fraction in the TZ on the B/T was also investigated. The results bring out that the MAs are converted by the transmutation rather than the incineration. Bothmodels can reduce significantly the effective half-lives of the MA nuclides by burning and/or transmuting these nuclides, and at the same time produce substantial electricity insitu.     

A design study on Pb-208 cooled compact CANDLE burning reactor for future nuclear energy supply

A compact pool-type Pb-208 cooled CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) with a thermal power rating of 125 MWth is considered for the future nuclear energy supply. Natural Pb consists of Pb-204, Pb-206, Pb-207 and Pb-208. Pb-208 has a small capture and inelastic-scattering cross-section, which makes it possible to reduce neutron capture by coolant and to make neutron spectrum harder. In case of Pb-208 coolant instead of natural Pb, the core height and radius are reduced to 1.5 m and 1 m, respectively. The effective multiplication factor of the core, k_eff, could be increased from k_eff = 0.984 of natural Pb up to k_eff = 1.006. For increasing natural circulation head, coolant velocities in each core zone are adjusted by orifice at the core inlet position. The reactor vessel height is equal to that of a typical loop-type demonstration FBR vessel to obtain natural circulation head.     

European Commission - 7th Framework Programme: The Collaborative Project on European Sodium Fast Reactor (CP ESFR)

Fast Reactors have a unique capability as sustainable energy source; the closed fuel cycle allows significantly improving the usage of natural resources and the minimisation of volume and heat load of high-level waste. Among the fast reactor systems, the sodium-cooled fast reactor has the most comprehensive technological basis, thanks to the experience gained internationally from operating experimental, prototype and commercial size reactors.The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named “Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR” further identifying, organizing and implementing a significant part of the needed R&D effort.The CP ESFR merges the contribution of 25 European partners; it is performed under the aegis of the 7th FP under the Area - Advanced Nuclear Systems with a refund from the European Commission of 5.8 M€ (11.55 M€ total budget). It is a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA).The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follows: an improved safety with in particular the achievement of a robust architecture vis à vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk similar to that of the other means of energy production; a flexible and robust management of nuclear materials and especially waste reduction through Minor Actinides burning.Within the paper relevant results of the first year of activity are succinctly presented; insights are given concerning the so called “working horses” cores and systems which have been provided by CEA and AREVA and that are used as a basis to test the performances and assess the pertinence of innovative solutions.     

Studies on LLFP transmutation in a pressurized water reactor

A systematic study on the long-lived fission product (LLFP) transmutation in a pressurized water reactor (PWR) is performed, aiming at an optimal transmutation strategy for present nuclear energy development. The LLFPs selected in the analysis include ⁹⁹Tc and ¹²⁹I discharged from light water reactors (LWRs). The isotope ¹²⁷I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, metallic technetium or MgI₂ target pins mixed with ZrH₂ are designed and investigated. Through the numerical analysis on equilibrium cycles, the transmuted amounts of ⁹⁹Tc and ¹²⁹I equal to the yields from 1.94 and 4.22 PWRs with a power of 1000 MWe, respectively. Numerical results indicate that both ⁹⁹Tc and ¹²⁹I can be transmuted conveniently in present PWRs in the form of target pins.     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

Enhancing BWR proliferation resistance fuel with minor actinides

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the ²³⁸Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (²³⁷Np and ²⁴¹Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of ²³⁸Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO₂ fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm³) to the top (0.35 g/cm³) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (²³⁷Np and/or ²⁴¹Am), significantly increases the ²³⁸Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.     

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B₄C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH_1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.     

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO₂-yttria stabilized zirconia (YSZ: (Zr,Y)O_2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as ²³⁸U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO₂ HTGR case by adding Er in the fuel. Burnup calculations on PuO_2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for ²³⁹Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO₂ HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.     

Using a modified CINDER90 routine in MCNPX 2.6.0 for the prediction of helium production in minor actinide targets

Fast reactors containing heterogeneous minor actinide (MA) target rods are now being modeled. When studying transmutation in these rods, helium production from α-decay must be considered since it is produced in substantial quantities. This research utilized an innovative method to calculate gas production by modifying the CINDER90 depletion code used by MCNPX 2.6.0 to include helium production from α-decay. The modified CINDER90 code was verified using the ORIGEN-ARP module of SCALE6. It was tested using the Sodium-Cooled Heterogeneous Innovative Burner Reactor model created at the University of South Carolina. It is recommended that the modified version of the cinder.dat file be distributed in subsequent MCNPX 2.6.0 releases for use in fast reactor calculations using heterogeneous MA target rods since it includes helium production otherwise not available from the current version.     

Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions

Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (P_ex = 7:2 MPa, T_in = 556 K) for mass velocity G = 400-800 kg/(m² s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The traditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles.