Full Power Days Research Articles

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.

Calculation and measurement of kinetic parameters of Pakistan Research Reactor-1 (PARR-1)

Rossi-Alpha ( β eff / Λ) for critical reactor measured experimentally by noise analysis technique at PARR-1 core at 35.26 full power days burn up. In noise analysis technique the inherent reactivity fluctuations are taken as input to reactor system and the neutron density population fluctuations are considered as output of the reactor system. The auto power spectral density of the linear channel is taken and used to find out the break frequency by non-linear least square fitting method, which leads to β eff / Λ = 161.45 s −1. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. The calculated β eff / Λ = 161.07 s −1 at 35.26 full power days burn up. The measured and calculated values for Rossi-Alpha are in good agreement within 0.235% of error.

Source term evaluation for the upgraded LEU Pakistan Research Reactor-1 under severe accidents

Evaluation of source term has been carried out for the upgraded LEU PARR-I system taken as a typical material test reactor (MTR). The modeling and simulation of release of radioactivity has been carried out by developing a Matlab based computer program which uses the ORIGEN2 code for core inventory calculations. For post 180 full-power days continuous operation, various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. For noble gases, iodine and for aerosols, the release rate studies have been carried out for the normal, emergency and for the isolation states of containment. The values of source term as well as that of containment retention factor show rapid increase followed by an approach towards saturation values as the exhaust rate values are increased. The isotope-dependency of the containment retention factor has been studied and the results indicate strong sensitivity for 85Kr, 137Xe, 138Xe and 138Cs towards exhaust rate values.

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity—Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL—

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

A class-based search for the in-core fuel management optimization of a pressurized water reactor

The In-Core Fuel Management Optimization (ICFMO) is a prominent problem in nuclear engineering, with high complexity and studied for more than 40 years. Besides manual optimization and knowledge-based methods, optimization metaheuristics such as Genetic Algorithms, Ant Colony Optimization and Particle Swarm Optimization have yielded outstanding results for the ICFMO. In the present article, the Class-Based Search (CBS) is presented for application to the ICFMO. It is a novel metaheuristic approach that performs the search based on the main nuclear characteristics of the fuel assemblies, such as reactivity. The CBS is then compared to the one of the state-of-art algorithms applied to the ICFMO, the Particle Swarm Optimization. Experiments were performed for the optimization of Angra 1 Nuclear Power Plant, located at the Southeast of Brazil. The CBS presented noticeable performance, providing Loading Patterns that yield a higher average of Effective Full Power Days in the simulation of Angra 1 NPP operation, according to our methodology.

CANDU refuelling optimization by successive two-step mathematical programming – I: MILP models for refuelling channel selection

A CANDU refuelling optimization method based on successive two-step mathematical programming (MP) is developed. The first step is for the selection of weekly refuelling candidates, while the second step is for the determination of the detailed weekly refuelling scheme with the prescribed candidates. To search for the overall optimal solution with the aftereffects of refuellings being considered, a zero- dimensional core physics model and a three-dimensional core physics model are directly incorporated into their corresponding MP model. The formulated large-scale MP problems are solved by use of the powerful generic MP problem solver CPLEX® (commercial product of IBM ILOG). The potential for engineering application of the method is demonstrated by its on-site test application to Qinshan CANDU 6 reactor. Refuelling simulation against the operation history of 392 full power days demonstrates that in addition to the good regional power and core reactivity ripple control, higher average fuel discharge burnup of 176.54 MWh/kgU is achieved. No violation of reactor operating limits is found. Regarding the computational cost of the method, though for most tested cases the performance of the method satisfies the stringent on-site time requirement, some exceptional cases are observed. However this occasional speed problem is within a factor of two or three from the targeted requirement, which is realistically optimistic in the near future considering today’s rapid advancement in hardware and software.

Out of pile tritium release behaviour and microscopic investigation of lithium metatitanate irradiated in the High Flux Reactor in Petten

The irradiation experiment EXOTIC 9/1 is the latest in a series of EXOTICs that has been irradiated in the high flux reactor in Petten and studies tritium release and inventory in ceramic pebbles containing lithium, which are key properties for the fusion fuel cycle. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The pebbles, produced by an extrusion–spheroidisation–sintering process at CEA in 2004, differ in porosity and density from the Li 2TiO 3 ceramics tested in the previous EXOTIC 8 programme. The pebbles have a diameter in the range from 0.6 to 0.8 mm, a high density (93% TD) and an open porosity of 1.7%. The pebbles mainly differ in enrichment (7.5% and 30% 6Li-enrichment). During irradiation of 300 Full Power Days in the HFR in Petten, the temperature in EXOTIC 9/1 was varied between 340 and 580 °C. The final irradiation (EOL) temperature is 480 °C and the average total 6Li burn-up is 3%. The total measured activity from tritium during irradiation is 220.42 Ci. After the completion of the irradiation, the out of pile release behaviour was studied by temperature programmed desorption (TPD). More specifically, the remaining tritium inventory in the pebbles and the activation energy of different release mechanisms were studied. The TPD was carried out for Li 2TiO 3 samples with 7.5% (natural) and 30% (enriched) 6Li-abundance and revealed very different curves, indicating that the materials have different release mechanisms. Dissolving the irradiated samples in hydrogen peroxide indicates that no tritium is left in the pebbles after TPD. Results from optical microscopy show it is likely that more intense bubble formation has taken place in the enriched samples, compared to the natural samples.

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity — Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL —

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

Global Depletion Analysis of Korean Helium-Cooled Solid Breeder TBM Model for Testing in ITER

The Korean HCSB (Helium Cooled Solid Breeder) TBM (Test Blanket Module), whose breeding zone is composed of lithium ceramic, beryllium and graphite in pebble form, was designed based on LOCAL assumption. In this research, we establish a simple preliminary GLOBAL neutronics model for the Korean HCSB TBM and perform neutronics analyses including depletion (transmutation) calculation during 500EFPDs (Effective Full Power Days) using the modified MONTEBURNS code. The neutronics characteristics for D-D plasma phase are investigated in the preliminary GLOBAL neutronics model, and the results are compared with those of D-T plasma phase. Moreover, we also establish the A-lite based GLOBAL neutronics model for more reliable neutronics calculation, and the results are compared with each other.

Reactivity control options of space nuclear reactors

This paper compares two ex-core control options of the gas-cooled Submersion Subcritical Safe Space (S^4) reactor with a fast neutrons energy spectrum: (a) rotating BeO drums with 120° thin segments of enriched B 4C in the BeO radial reflector; and (b) sliding segments in the BeO radial reflector. Investigated are the effects on the beginning-of-life (BOL) excess reactivity, reactivity depletion rate and operation life, and the spatial neutron flux distributions and fission power profiles in the core. Also investigated is the effect of reducing the thickness of the enriched B 4C segments in the control drums on the BOL excess reactivity, when one or two of the 6 drums are stuck in the shutdown position. Reducing the thickness of the B 4C segments from 0.5 mm to 0.238 mm, with one drums stuck in the shutdown position, increases BOL cold and hot-clean excess reactivity from +$1.71 and +$0.47 to +$2.38 and +$0.89, respectively. These reactivity values are almost identical to those of the reactor with one of the six reflector segments stuck open in the shutdown position. Results also showed that the control options made little difference in the reactor performance. The power peaking in the reactor core with sliding reflector segments is slightly lower and the spatial power profiles are relatively flatter. The operation life of the reactor with a sliding reflector segments control, when operating at a nominal thermal power of 471 kW, is only 22 full power days longer than with rotating drums control.

A Sensitivity Study for Determining the RCS Volume through an Evaluation of Specific Activities of the SMART Primary Coolant

The specific activities of the SAMRT primary coolant due to fission products were evaluated and they will be used as the source terms of the BOP shielding design. The level of a specific radioactivity is also one of the important factors for determining the RCS volume. SMART reactor which is in the process of being developed consists of 69 fuel assemblies with a 660Mw thermal output. The reactor core has 2 batch fuel assemblies with a 30 month cycle and the effective full power days of the equilibrium cycle are 822 days. The specific activities of the SMART reactor primary coolant due to fission products were evaluated by using the SAEP code. The SAEP code can be applied to an integral reactor such as SMART as well as the current commercial PWRs in existence. The ORIGEN2 code is integrated into the SAEP as a subroutine, to solve the inhomogeneous differential equations which are used to determine the coolant activities. From the evaluation results, the specific activities of the SMART primary coolant due to fission products were compared with those of UCN 5&6 primary coolant with the same purification rate condition; they are averagely 25% of those of UCN 5&6. They are also inversely proportional to the RCS volume for almost all radioisotopes except the isotopes with a long half-life. Therefore, it is found that there is a possibility of reducing the RCS volume.

Modelling of the HFR-EU1BIS experiment and thermomechanical evaluation

Numerical modelling has been successfully applied during the design and evaluation of the HFREU1bis HTR fuel pebble irradiation experiment conducted by the Joint Research Centre Institute for Energy (JRC) in the HFR Petten, The Netherlands. HFR-EU1bis contains 5 HTR fuel pebbles that are placed in a graphite shroud. This assembly is placed in a containment tube, which is again placed in another containment, which is in contact with the HFR cooling water. The experimental requirement is to maintain a central temperature of 1250 °C for all pebbles throughout the irradiation time of 250 effective full power days (efpd). This was achieved by tailoring the gas gaps in the sample holder such that the axial heat generation profile is compensated. Determining the dimensions of the components of the experiment has been done using a thermomechanical finite element model. A combination of numerical and analytical techniques has been applied to minimise calculation times. The thermomechanical finite element model includes the thermal influence of changing gas gap dimensions due to expansion. Heat transfer by radiation through the gas gap has been modelled as well. Results of the model have been compared with measured temperatures. After some adjustments of the model parameters, good agreement has been found between calculated and measured temperatures. Additionally, the influence of irradiation on graphite thermal conductivity and thermal expansion has been included in the model. Due to irradiation damage the thermal conductivity of graphite decreases and thermal expansion generally increases, leading to increasing thermal stresses. Due to burn-up however, the heat flux decreases during irradiation, which leads to a decrease in thermal stresses. To evaluate the influence of these competing mechanisms, the thermal stress evolution during irradiation has been calculated and will be discussed.

AC-3-irradiation test of sphere-pac and pellet (U,Pu)C fuel in the US Fast Flux Test Facility

The objective of the AC-3 bundle experiment in the Fast Flux Test Facility (FFTF) was to evaluate a fuel fabrication method by ‘direct conversion’ of nitrate solutions into spherical uranium–plutonium carbide particles and to compare the irradiation performance of ‘sphere-pac’ fuel pins prepared at Paul Scherrer Institute (PSI) with standard pellet fuel pins fabricated at Los Alamos National Laboratory (LANL). The irradiation and post test examination results show that mixed carbide pellet fuel produced by powder methods and sphere-pac particle fuel developed by internal gelation techniques are both valuable advanced fuel candidates for liquid metal reactors. The PSI fabrication process with direct conversion of actinide nitrate solutions into various sizes of fuel spheres by internal gelation and direct filling of spheres into cladding tubes is seen as more easily transferable to remote operation, showing a significant reduction of process steps. The process is also adaptable for the fabrication of carbonitrides and nitrides (still based on a uranium matrix), as well as for actinides diluted in a (uranium-free) yttrium stabilized zirconium oxide matrix. The AC-3 fuel bundle was irradiated in the Fast Flux Test Facility (FFTF) during the years 1986–1988 for 630 full power days to a peak burn up of ∼8 at.% fissile material. All of the pins, irradiated at linear powers of up to 84 kW/m, with cladding outer temperatures of 465 °C appeared to be in good condition when removed from the assembly. The rebirth of interest for fast reactor systems motivated the earlier teams to report about the excellent, still perfectly relevant results reached; this paper focusing on the sphere-pac fuel behaviour.

Automatic refueling simulation of a DUPIC fuel transition core

This paper presents an automatic refueling simulation of a 713MWe Canada deuterium uranium (CANDU-6) reactor transition to Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC) fuel system. The simulation searches for a set of refueling channels based on the zone controller unit water level, the reactivity insertion tilt in two axial zones and the discharge burnup. Compared to the natural uranium core, the transition core has two distinctive features such as an axial power peaking and an inconsistent reactivity insertion into two axial zones with a fueling direction, which are accommodated by adjusting the reference power of the central zone and complying with the reactivity insertion tilt in the axial direction. Starting from the natural uranium equilibrium core, it took 730 full power days to completely fill the core with the DUPIC fuel, and the characteristics of the transition core represented by the maximum channel and bundle powers were far below the license limits.

Irradiation damage to the beam window in the 800MWth accelerator-driven system

Irradiation damage to the beam window in the concept of 800MWth accelerator-driven system is evaluated. Heat produced in the window is also evaluated. Transport of proton and neutron up to 3.0 GeV is calculated by both PHITS that is the Monte Carlo code for particles and heavy ions and TWODANT that is two-dimensional deterministic transport code. The beam window is irradiated at the center of the accelerator-driven system with 20 MW proton beam power and neutron from the core during 300 full power days. Heat, displacement per atom, production rate of hydrogen and helium isotopes, and neutron and proton fields are estimated, assuming the Gaussian and flat beam profiles.