Burnup Reactivity Research Articles

The boundary problem in the 1D-CANDLE burn-up reactor

Abstract We have analytically solved the boundary problem for a solitary CANDLE-Burn Up wave in a slab reactor configuration using the affine approach which can be considered as the analogon to the Buckling-Concept for spatially/stationary critical zones in linear reactor physics. It sheds some light on the problems, we are confronted with, when we have to deal with nonlinear mathematics in this new field of reactor physics.

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

Dynamic Characteristics of Compact Fast-Spectrum Reactors

This paper focuses on some of the unique dynamic characteristics of compact fast-spectrum reactors. The study is limited to the characteristics that are relatively independent of how the reactor is integrated into a complete power system. Some of the well-established characteristics of compact fast-spectrum reactors are that point kinetics is generally very accurate for these systems and that temperature and burnup reactivity feedback mechanisms are relatively small and simple. Beyond this, there are two unique aspects of highly reflected fast reactors (e.g., space reactors) that do not occur in more traditional reactors. First, the neutron reflector has a very important impact on dynamic performance, and in some cases the temperature coefficient of the radial reflector is higher than that of the fuel. The thermal time constant of the reflector is much longer than that of any component in the core, which requires all reflector temperature and expansion effects to be modeled individually. Second, reflected neutrons have a much longer fission life span than in-core neutrons. In effect, this creates additional delayed neutron groups, referred to as geometric delayed neutron groups. These groups can have life spans orders of magnitude longer than neutrons that do not leave the core, and have much higher worth due to moderation. For compact beryllium reflected reactors there is also a measurable delayed group of photo-induced neutrons that result from delayed gammas. Another characteristic of compact fast-spectrum reactors is simplified control and the ability to passively handle a wide range of transients without control. Various transient analyses are presented that were performed by the Fission Reactor Integrated Nuclear Kinetics (FRINK) code, which facilitates near-term compact reactor design and development by providing a transient analysis tool. In its current state FRINK is a very simple system model, and the “system” only extends to the primary loop power removal boundary condition; however, this allows the simulation of simplified transients (e.g., loss of primary heat sink, loss of flow, etc.).

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ( ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

Burnable poison for reactivity management in a very high temperature reactor

A physics study has been performed to find the optimal application of burnable poison (BP) for an excess reactivity management in a 600 MWth block-type very high temperature reactor (VHTR). Six potential BP materials (B, Gd, Er, Eu, Sm, Dy) were evaluated for an equilibrium cycle in terms of the major core performance parameters such as the burnup reactivity swing, discharge burnup, fuel and moderator temperature coefficients, and fuel temperature. In addition to the conventional 6-hole BP loading in a fuel block, a 12-hole BP application was also considered in order to accelerate the depletion rate of the BPs. The self-shielding effect of the BP was optimized for a successful management of the excess reactivity by changing the effective diameter of the BP region. Additionally, a mixed BP application of Gd and Er was proposed to make better use of the two BP materials in terms of the reactivity swing and the core temperature coefficients. It has been demonstrated that the burnup reactivity swing over a long cycle period can be reduced from ∼15,000 pcm to 3000–5000 pcm with only a small reduction of the discharge fuel burnup.

Core design studies for a 1000 MW th Advanced Burner Reactor

This paper describes the core design and performance characteristics of 1000 MW th Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of ∼0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from ∼0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

Recycling-Independent Core Design for the ENHS Fuel Self-Sustaining Reactor

The feasibility of indefinite recycling in the Encapsulated Nuclear Heat Source (ENHS) core without changing the pitch-to-diameter (P/D) ratio, while maintaining a nearly zero burnup reactivity swing, is investigated. The P/D ratio required to achieve a nearly burnup-independent keff over the life of the ENHS core was found sensitive to the initial composition of the transuranium (TRU) loaded and to the number of recycles this fuel underwent. The longer the cooling time is of the TRU from light water reactor (LWR) spent fuel, the larger the optimal P/D ratio becomes. Whereas the optimal P/D ratio of the reference ENHS core that is fueled with TRU from LWR spent fuel discharged at 50 GWd/t heavy metal (HM) and cooled for 10 yr is 1.36, it is 1.54 for the equilibrium core that features a substantially smaller concentration of 241Pu as well as of 242Pu, a larger concentration of 239Pu, and a substantially larger concentration of minor actinides. It was found that by increasing the cooling period of the above LWR TRU to ~32 yr, the optimal first core P/D ratio is that of the equilibrium core. The burnup reactivity swing of the subsequent cores fueled with successive recycling of the ENHS discharged HM is satisfactory. There is no need to adjust the core P/D ratio from recycle to recycle. The power level that can be removed by natural circulation from the P/D = 1.54 core is ~36% higher than that of the reference ENHS core. The physical phenomena affecting the observed trends are discussed, and the neutronic characteristics of the equilibrium cores identified are summarized.

Application of Covariances to Fast Reactor Core Analysis

In the present paper, the current status of covariance applications to fast reactor analysis and design is summarized with actual examples. The covariance applications are classified into three fields. First, covariances are used to quantify the uncertainty of nuclear core parameters such as criticality, control rod worth, reaction rate ratio, power distribution, sodium void reactivity, etc. The special features of the Japanese case are to include both the burnup-related parameters such as burnup reactivity loss or fuel composition changes, and the temperature-related parameter, that is, the Doppler reactivity. Second, covariances are used to select the important nuclides, reactions and energy ranges which are dominant to the uncertainty of core parameters, and to give nuclear scientists recommendations to improve the accuracy of the data. Finally, covariances are used to improve the accuracy of core design values by adopting the integral data such as the critical experiments and the power reactor operation data. The way toward improvement is classified into the conventional E/C (Experiment/Calculation) bias method and the more comprehensive cross section adjustment based on the Bayesian theorem and the generalized least square method. In Japan, an adjusted group-constant set, ADJ2000R, is now being used in the design work for future fast reactors.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn’s f-parameter during its fabrication process.

New 237Np Burning Strategy in a Supercritical CO2-Cooled Fast Reactor Core Attaining Zero Burnup Reactivity Loss

A new 237Np burning strategy in a supercritical CO2-cooled fast reactor core has been proposed: consuming 237Np as fuel and burnable poison to attain zero burnup reactivity loss. Addition of 237Np at content of 6.5 wt% in fuel engenders nearly zero burnup reactivity loss of 0.02% Δk/k during 10 yr. The burning rate of 237Np in the core is ~69 kg/yr, which is equivalent to the quantity produced in a year from about 20 light water reactors of equivalent electrical output. The zero burnup reactivity loss enables reduction of the control rod number to half that of a typical sodium-cooled mixed-oxide fuel MONJU core without added 237Np and no need for rod operation with fuel burning to compensate for the burnup reactivity loss. Void reactivity is 0.72% Δk/kk’, which is three-fourths that of a typical Na-cooled core, although 237Np is added and the active core length is elongated to 1.2 m. The power density is reduced to ~20% of that in a Na-cooled core. The hot-spot temperature of cladding is below its maximum permissible temperature of 700°C.

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.

Optimization of Burnable Poison Loading for HTGR Cores with OTTO Refueling

A burnable poison (BP) loading principle has been proposed for once-through-then-out refueling of a high-temperature gas-cooled reactor (HTGR) core with pebble fuel. The principle holds that an axial core power peaking factor can be minimized when k∞ of the fuel pebbles is kept constant during their axial movement from the top to the bottom of the core by adding BP. This principle has been confirmed numerically using B4C with 10B enrichment of 90% and Gd2O3 with natural content as BP. Spherical particles of B4C and Gd2O3 are distributed uniformly in the fuel pebble. The respective optimal radius and number of BP particles are 90 μm and 1650 for B4C and 950 μm and 16 for Gd2O3. Through addition of B4C and Gd2O3, the power peaking factors are reduced from 4.4 to 1.61 and 1.64, respectively. Burnup reactivity swings are reduced from 38% to about 2% in both BP loadings. Because of reduction of the power peaking factors, the maximum fuel temperatures are respectively lower than the maximum permissible values of 1250 and 1600°C for normal operation and depressurization accident.

Multi-recycling in the ENHS and the equilibrium core

This work investigates the effect of initial fuel composition, power density and number of recycles on the pitch-to-diameter ( P/ D) ratio and TRans-Uranium isotopes (TRU) loading required for attaining one of the most important design goals of the Encapsulated Nuclear Heat Source (ENHS) – nearly zero burnup reactivity swing over the ∼20 years of core life. It is found that the required P/ D ratio is sensitive to, primarily, the initial concentration of the short-lived isotope 241Pu in the fuel loaded into the first core and to the core power density. The longer is the cooling time of the TRU from LWR spent fuel the smaller becomes the relative 241Pu concentration and the smaller becomes the fraction of 241Pu lost via radioactive decay and, hence, the smaller needs be the conversion ratio required for nearly zero burnup reactivity swing and the larger can be the P/ D ratio. Likewise, the higher is the ENHS power density, the smaller becomes the fraction of 241Pu lost via radioactive decay and the larger becomes the P/ D required for the first core. The optimal P/ D ratio tends to increase with the number of times the fuel is recycled from one ENHS core to the next one. The optimal P/ D ratio for the equilibrium composition core is in between 1.53 and 1.59. For a given discharge burnup it tends to somewhat increase with the equilibrium core power density. However, if structural materials will be developed to enable a 20 years core life at elevated power densities, the higher the power density the smaller is the required equilibrium P/ D ratio.

Design study on small CANDLE reactor

A new reactor burn-up strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burn-up reactivity control mechanisms such as control rods are not required. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal during power operation. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burn-up. Therefore, the operation of the reactor becomes much easier than the conventional reactors especially high burn-up reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. In our previous works it appeared that application of this burn-up strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burn-up of the spent fuel is about 40% of total charged fuel. It is equivalent to 40% utilization of the natural uranium without the reprocessing and enrichment. This reactor can be realized for large reactor, since the neutron leakage becomes small and its neutron economy becomes improved. In the present paper we try to design small CANDLE reactor, whose performance is similar to the large reactor, by increasing its fuel volume ratio of the core, since its performance is strongly required for local area usage. Small long-life reactor is required for some local areas. Such a characteristic that only natural uranium is required after second core is also strong merit for this case. The core with 1.0 m radius, 2.0 m height can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as a coolant. From equilibrium analysis, we obtained the burning region velocity, power density distribution, core temperature distribution, etc. The burning region velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE ( Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.

Feasibility of Designing the Encapsulated Nuclear Heat Source Reactor with Negative Void Reactivity Feedback

The Encapsulated Nuclear Heat Source (ENHS) is a small lead-bismuth-cooled Generation IV reactor designed to have a once-for-life core with a nearly zero burnup reactivity swing and heat removal by natural circulation. All the ENHS cores designed so far have positive coolant void reactivity. This work searches for ENHS core designs having negative coolant void reactivity feedback and quantifies the penalty associated with a design for negative void reactivity. The approaches tried for turning the positive void coefficient negative are as follows: (a) enhancing the neutron leakage probability by reducing the fuel length, introducing neutron absorbers at the core boundary, using a gas-lift pump to introduce gas bubbles throughout the coolant in the core and fission gas plenum regions, and incorporating neutron-streaming channels in and adjacent to the core; (b) introducing into the core materials, such as Ca3N2, having enhanced absorption cross section at high energy; and (c) introducing into the core materials, such as CaH2, that will keep the neutron spectrum softer in the case of coolant voiding.The preferred negative void reactivity core design consists of 100-cm-long fuel rods, a 20-cm-thick B4C layer below the core, and a voided channel around the core radial boundary. The reactivity effect is negative when the coolant is voided from the entire core and even from only the central region of the core. In order to maintain a nearly zero burnup reactivity swing over at least 20 effective full-power years (EFPY), the core pitch-to-diameter ratio (P/D) has to be reduced from the reference value of 1.36 to 1.20. Correspondingly, for a given ENHS module dimensions the power level that can be removed by natural circulation from this core is ~75% of the reference core. Allowing a burnup reactivity swing of ~0.1% over 20 EFPY enables attaining a negative void reactivity core having P/D of 1.27 that can deliver ~93% of the P/D = 1.36 core power.

Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, 148Nd, 145Nd+146Nd, total of the Nd isotopes, 133Cs and 137Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (233U, 242Pu, 150Nd, and 133Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.