Burnup Reactivity Research Articles

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Abstract In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

High burnup calculation characteristics of traveling wave reactor

The traveling wave reactor (TWR) is an innovative nuclear system, whose discharge burnup is 400 GWd/tHM, about three to four times of existing fast reactor and six to eight times of pressurized water reactor. High discharge burnup forces the present codes to face the great challenges in precision.This paper studies the high burnup calculation characteristics of traveling wave reactor from energy spectrum, importance of fission production and calculation error accumulation of fuel consumption by KYLIN-1 code. The analysis results of typical traveling wave reactor hexagonal assembly show that the low enriched uranium assemblies have different energy spectra at beginning and end of life, which cause big calculation error ofkinf by existing code system. To ensure the correctness of the traveling wave reactor burnup calculations, the burn chain should contain 70 kinds of important fission product nuclides. The calculation error accumulation of fuel consumption is small when the burnup step is increased. The calculation error ofkinf is about 0.001% each burnup step.

ZrC COATING ON FUEL ELEMENT CLADDING ZIRCALOY-2

ZrC COATING ON FUEL ELEMENT ZIRCALOY-2 CLADDING. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR) fuel element were performed due to the extension of fuel element’s utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. This technique is supposed to improve the corrosion resistance of cladding without changing the dimension of fuel cladding. In current research, the ZrC film was coated on the zircaloy-2 cladding surface by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens then undergone heating process at 700oC, 900oC and 1100oC respectively in Argon gas atmosphere for 1 hour. The microstructure and crystal structure of the coated cladding were characterized by optical microscope and XRD respectively. The optical microscope showed the growth of the grains with increasing temperature. XRD examination on the specimens revealed that the ZrC crystal structure on the cladding surface occurred only at 1100oC, but it did not appear at 700oC and 900oC. It can be concluded that dipping process of specimen in colloidal graphite with subsequent heating at 1100oC provided ZrC film coated on zircaloy-2 cladding. The heating process at this temperature allowed carbon atoms to diffuse into zircaloy surface to form ZrC film.PELAPISAN ZrC PADA KELONGSONG ELEMEN BAKAR NUKLIR ZIRKALOI-2. Riset yang intensif pada elemen bakar reaktor berpendingin air dengan fraksi bakar tinggi terus dilakukan dalam rangka memperpanjang umur operasi elemen bakar. Salah satu riset tersebut berupa proses untuk mengubah kelongsong berbasis zirkonium yang ada saat ini dengan cara pelapisan. Cara ini diharapkan akan memperbaiki ketahanan korosi kelongsong tanpa mengubah dimensi kelongsong tersebut. Pada riset ini, lapisan tipis ZrC dilapiskan pada permukaan kelongsong zirkaloi-2 melalui proses pencelupan (dipping) spesimen zirkaloi-2 dalam koloid grafit pada temperatur kamar. Spesimen yang telah dicelup dipanaskan masing-masing pada temperatur 700oC, 900oC dan 1100oC dalam media gas Argon selama 1 jam. Kelongsong yang telah dilapisi dan mengalami proses pemanasan, dikarakterisasi menggunakan mikroskop optik dan uji XRD untuk mengetahui mikrostruktur dan struktur kristalnya. Pengujian dengan mikroskop optik memperlihatkan pertumbuhan butir seiring dengan naiknya temperatur pemanasan. Pengujian dengan XRD mengungkapkan bahwa struktur ZrC terbentuk pada permukaan spesimen kelongsong zirkaloi-2 yang dipanasi pada temperatur 1100oC. Struktur ZrC tidak muncul pada spesimen hasil pelapisan dengan pemanasan pada temperatur 700oC and 900oC. Berdasarkan analisis data dari riset ini dapat disimpulkan bahwa pembentukan lapisan tipis ZrC pada permukaan spesimen kelongsong zirkaloi-2 dapat dilakukan melalui proses pencelupan spesimen dalam koloid grafit yang diikuti dengan pemanasan pada temperatur 1100oC yang memungkinkan atom-atom karbon mempunyai cukup energi untuk mendifusi ke permukaan kelongsong membentuk struktur ZrC.

Robust PID control of power in lead cooled fast reactors: A direct synthesis framework

Output core power control for an innovative design of lead cooled fast reactors (BREST) is investigated applying a PID design approach. The reactor core model is developed through a lumped parameter thermo-neutronic coupled layout of the plant. Control design is carried out within a direct synthesis framework whereby desired closed loop characteristics are incorporated. The PID coefficients are calculated employing a Laurent expansion of the intended controller and the resultant control scheme is employed on the uncertain nonlinear reactor plant. A worst case robust stability analysis is thereby conducted for the extreme hypothetical working point of the reactor subject to a near prompt criticality reactivity insertion. Results confirm that the control practice proposed may quite satisfactorily be implemented on the original plant as reasonable robustness characteristics are preserved for this ultimate operational point of the system. Nyquist stability criterion has been resorted for robust stability analysis with cycle burnup reactivity swing taken as the source of uncertainty and further recast into a multiplicative layout. The overall control method for desired power trajectories provides a safe reactivity insertion rate and a feasible rod speed.

New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between 235U and 239Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller.

Design Study of Thorium Cycle Based Long Life Modular Boiling Water Reactors

Design study of long life (cycle) Boiling Water Reactor, which can be operated for 20-30 years of operation time without the necessity of refuelling during that period, has been performed. In shown in the previous study Th232-U233 cycle based fuel has potential for longer operation time in the thermal reactor domain. To increase reactor operation time and minimize excess-reactivity burnable poison may be used. Protactinium (Pa-231), Np-238, or gadolinium has good properties for such purpose. Here based on the previous study we select Protactinium and Gadolinium poisons to be used in the present study. Optimizations the content of 231Pa in the core enables the BWR core to sustain enough reactivity for long period of time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation compositions can be achieved reactor core with longer operation time, 20 ~ 30 years operation without fuel shuffling or refuelling. Similarly Gadolinium has been successfully used to extend refuelling time and reduce excess reactivity during burnup period for both oxide fuel and nitride fuel.

Core design options of an ultra-long-cycle sodium cooled reactor with effective use of PWR spent fuel for sustainable energy supply

In this work, 350MWe ultra-long-cycle sodium-cooled reactor cores are designed to supply electric energy over ~60 Effective Full Power Years (EFPYs) without refueling and with an effective use of Transuranics (TRU) and uranium from large pressurized water reactor (PWR) spent fuel stocks. The core employs the axial blanket-driver-blanket (ABDB) burning strategy, which was recently proposed by the authors to achieve an ultra-long-cycle length with self-controllability under unprotected accidents. In particular, a thorium–uranium fuel cycle is considered to remove the heterogeneity of the fuel assemblies for design simplification and to improve the core performance parameters by selectively adding thorium into both blanket and driver fuels. The results show that the use of TRU nuclides from PWR spent fuel leads to significant extension of the fuel cycle length, but considerable increase of burnup reactivity swing. In addition, these results also indicate that the uranium–thorium mixed fuels both in the lower blanket and driver considerably improve the inherent safety of the ultra-long-cycle core by reducing burnup reactivity and sodium void worth; this makes it possible to simplify the previous heterogeneous fuel assembly design with improved core performances. Copyright © 2016 John Wiley & Sons, Ltd.

Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository.As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

Automatic construction of a simplified burn-up chain model by the singular value decomposition

Nuclear reactor design analysis often requires a simplified, or reduced-order, burn-up chain model to reduce computation time. It is difficult to construct the reduced-order burn-up chain because it requires engineers to have highly skilled techniques and in-depth knowledge into burn-up processes. This paper develops an algorithm for automatically constructing a reduced-order burn-up chain model from a detailed model using the singular value decomposition (SVD). In our approach, we prepare a detailed burn-up chain matrix A, and an extraction matrix C, which extracts important nuclides for specific purposes such as the evaluation of neutron multiplication factor. First, the nuclides extracted by C are specified as the first candidate nuclides of the reduced-order burn-up model. Then, by applying the SVD to C, we can obtain the first information transfer matrix F12(1), which defines the relationship between the first candidate nuclides and remaining nuclides. In the next place, by applying SVD to F12(1), we can obtain additional candidate nuclides for the reduced-order burn-up chain model from the remaining nuclides. We repeat this process until the norm of the information transfer matrix is sufficiently close to zero. Finally, all candidate nuclides chosen through these simplification processes are adopted as a reduced-order burn-up chain model. As a test case, we reduce a detailed burn-up chain model consisting of 1421 nuclides to a model of 204 nuclides. We can use the resulting reduced-order model to calculate the burn-up of light water reactor fuel cells with a high degree of accuracy.

A Core Physics Study of Advanced Sodium-Cooled TRU Burners with Thorium- and Uranium-Based Metallic Fuels

In this work, 400-MW(electric) sodium-cooled fast reactor cores using thorium- and uranium-based metallic fuels for high burning rates of light water reactor spent-fuel transuranics (TRUs) are neutronically designed and analyzed based on equilibrium cycles with a focus on consistent comparative analysis of the differences in performance between thorium- and uranium-based fueled cores. Axial uranium and thorium blankets are introduced in thorium- and uranium-based driver fueled burner cores to improve TRU burning rates without considerable increases of burnup reactivity swing. For this core configuration, it was shown that cores using thorium and depleted uranium blankets can be designed to have a high TRU burning rate, a low sodium void reactivity (SVR) worth, and a low burnup reactivity swing. In particular, the use of uranium or thorium blankets without recycling in the thorium-based driver fueled cores led to significant reductions of burnup reactivity swing with considerable increases of the TRU burning rate and small increases of SVR. In addition, the core configuration having central nonfuel regions was considered to show the effects of the thorium-based driver metallic fuel versus the uranium-based metallic fuel coupled with moderator rods. The core configuration with thorium-based fuel led to a negative SVR without moderator rods, and the use of moderator rods further improved the Doppler coefficient and reduced SVR. Also, a decomposition analysis of SVR was performed to better understand the differences in the contributing factors between the uranium- and thorium-based fueled cores, and a quasi-static reactivity balance analysis was performed to show the inherent safety of the cores in terms of self-controllability.

An optimization study on the excess reactivity in a linear breed-and-burn fast reactor (B&BR)

This work is concerned with minimization of the excess reactivity in a compact linear breed-and-burn fast reactor (B&BR). Neutronic studies are performed to reduce the core excess reactivity below 1$ in order to prevent possibility of the prompt criticality accident. A reference B&BR core design is introduced as a starting point for the optimization, which is loaded with vented annular metallic (U–Zr) fuels. The initial criticality of the B&BR core is achieved by using an LEU (low-enriched U) fuel and a spent nuclear fuel is reutilized as the blanket fuels. In order to minimize the transitional excess reactivity of the B&BR core, this study proposes a splitting of the Zr content in the U–Zr metallic fuel in combination with a geometrical optimization of the initial LEU core to flatten the radial power distribution simultaneously. A Zr-zoning is first applied to the initial LEU core to reduce excess reactivity and then a concave initial LEU core is also adopted to achieve both small excess reactivity and long lifetime of the B&BR core. In addition, a similar Zr-zoning strategy is also applied to the blanket region to optimize the excess reactivity behavior and flatten the radial power as well. It is demonstrated that a Zr-zoning combined with a concave core configuration can provides a promising neutronic performances. Finally, a recommended core has been characterized in terms of the burnup reactivity change, conversion ratio, power profiles, and reactivity coefficients.

A sodium-cooled ultra-long-life reactor core having improved inherent safety with new driver–blanket burning strategy

In this paper, two sodium-cooled ultra-long-life reactor cores are designed by using new axial blanket-driver burnup strategy such that they have ultra-long-life longer than 40 EFPYs (Effective Full Power Years), high average fuel burnup of ∼180 MWD/kg, small burnup reactivity swing, and improved inherent safety in terms of self-controllability under unprotected accidents. An exploration on the arrangement of driver and blanket fuels revealed that axially blanket-driver-blanket (B-D-B) configuration with different thicknesses in inner and outer core regions is effective to achieve the design goals. The final core was determined through the reduction of sodium void worth with an upper sodium plenum above the core and with partial use of dummy rods in the lower blanket of the inner core, and through the reduction of burnup reactivity swing with the adjustment of driver and blanket heights and with partial use of dummy rods in outer driver region. The neutronic analysis shows that the final core has desirable features such as long life of 53 EFPYs, high average burnup of 182 MWD/kg, small burnup reactivity swing of 773 pcm, and satisfaction of all the conditions for self-controllability under unprotected accidents.

Parallel GPU implementation of PWR reactor burnup

This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

Thorium utilization in existing and advanced reactor types

A comparative study of the capability of some of the existing power reactor designs to produce fissile material that is vital for sustenance of fission nuclear power is made. The fissile conversion rate is less in reactors using enriched fuel compared to the natural uranium fueled ones. Even in fast reactors the net fissile breeding is feasible only due to blanket regions. A case is made for reconfiguring the existing reactor designs to accommodate internal blankets or exclusive fissile breeding zones. For maximizing the net fissile accumulation it is suggested to use suitably large volume/mass of fertile breeding zones and extend the fuel cycle duration to as long as 2 or 3 years. The burnup reactivity swing of such modified core designs can be as little as ±5 mk over 2 to 3 full power years, thus obviating the need for large excess reactivity control through either burnable absorbers like boron, Gd or control absorbers. The safety characteristics of the new core designs improve with minimal reactivity changes due to temperature, Xe, Sm, burnup and voiding.

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 ) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

3D in-core fuel management optimization for breed-and-burn reactors

Breed-and-burn (B&B) reactors are a special class of fast reactors that are designed to utilize low grade fuel such as depleted uranium without fuel reprocessing. One of the most challenging practical design feasibility issues faced by B&B reactors is the high level of radiation damage their fuel cladding has to withstand in order to sustain the B&B mode of operation – more than twice the maximum radiation damage cladding materials were exposed to so far in fast reactors. This study explores the possibility of reducing the minimum required peak radiation damage by employment of 3-dimensional (3D) fuel shuffling that enables a significant reduction in the peak-to-average axial burnup, that is, more uniform fuel utilization. A new conceptual design of a B&B core made of axially segmented fuel assemblies was adopted to facilitate the 3D shuffling. Also developed is a Simulated Annealing (SA) algorithm to automate the search for the optimal 3D shuffling pattern (SP). The primary objective of the SA optimization is to minimize the peak radiation damage while its secondary objective is to minimize the burnup reactivity swing, radial power peaking factor and maximum change of fuel assembly power over the cycle. Also studied is the sensitivity of the 3D shuffled core performance to the number of axially stacked sub-assemblies, core height and power level.It was found that compared with the optimal 2-dimensional (2D) shuffled core, the optimal 3D shuffled B&B core made of four 70 cm long axially stacked sub-assemblies and 12 radial shuffling batches offers a 1/3 reduction of the peak radiation damage level – from 534 down to 351 displacements per atom (dpa), along with a 45% increase in the average fuel discharge burnup, and hence, the depleted uranium utilization, while satisfying all major neutronics and thermal-hydraulics design constraints. For the same peak dpa level, the 3D shuffling offers more than double the uranium utilization and the cycle length relative to 2D shuffling. The minimum peak radiation damage is increased to 360 or to 403 dpa if the core is made of, respectively, three – 70 cm or two – 140 cm long axially stacked subassemblies. Reducing the length of the subassemblies of B&B cores made of three-segment assemblies from 70 cm to 60 or 50 cm results in an increase in the peak radiation damage from 360 dpa to, respectively, 368 and 397 dpa.

Physical characteristics of large fast-neutron sodium-cooled reactors with advanced nitride and metallic fuels

Mixed nitride uranium–plutonium fuel is the most attractive and advanced type of fuel for fast-neutron sodium-cooled reactors, and is considered as the base fuel for future commercial fast-neutron power reactors. However, a substantial increase in breeding achieved thanks to the use of this fuel instead of oxide fuel is not enough to meet new requirements the major of which, minimization of the burn-up reactivity margin, is defined by the reactor core breeding. This paper presents the results of computational studies aimed at choosing the best possible layout for the core of a large fast-neutron sodium-cooled reactor meeting modern requirements.Metallic fuel has been considered as the fuel for fast-neutron reactors since their early designs due to high density and heat conductivity and the smallest possible number of the diluent nuclei which provides for the highest possible breeding efficiency. The paper presents the features of the fuel under consideration and the results of computational studies into its application in large fast-neutron sodium-cooled reactors as compared to nitride fuel. A conclusion is made that, with the same design of the reactor core, metallic fuel is inferior to nitride fuel from the point of view of ensuring the safety of the reactor facility.Most of the calculations were performed in a diffusive approximation based on the TRIGEX software package.

Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.