Discharge Burnup Research Articles

Design study of gas-cooled fast reactor with natural uranium as fuel employing modified CANDLE shuffling strategy in the axial direction

Abstract This study investigated gas-cooled fast reactor performance utilizing natural uranium as the fuel employs modified constant axial shape of neutron flux, nuclide densities, and power shape during life of energy production (CANDLE) shuffling strategy in the axial direction. The performance has been carried out on a reactor with various power levels in the range of 2,700–3,100 MWt and refueling every 10 years of burnup. The main importance of the modified CANDLE (MCANDLE) is that it can utilize natural uranium as fuel without needs reprocessing or enrichment plant. During this work, the core has been partitioned into 10 regions of similar volume along the axial direction. The fuel was filled into the first region, then transported to the second region after 10 years of burnup, and then to the third region 10 years later. This technique was practiced to all 10 regions, and the fuel in the tenth region was discharged in the reactor core. 10 % of natural thorium, natural uranium and enriched nitride (15 N) were used as fuel input. The calculations were performed employing a standard reactor analysis code (SRAC). The collision probability method (PIJ) module of SRAC was used to calculate cell burnup, and the reactor core design calculations were done with the CITATION module of SRAC, which used JENDL-4.0 as a nuclear data library. The result shows that a high power level causes an increasing rate of burnup level and effective multiplication factor. The highest average discharge burnup level is about 31.3 % HM for case E and the lowest average discharge burnup is about 27.2 % HM for case A.

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

Refueling Considerations for 99Mo Production in a CANDU Reactor

Operating CANDU reactors have the potential to produce significant quantities of molybdenum-99 (99Mo) because of their ability to be refueled online, high thermal neutron flux, and fuel design flexibility. A new molybdenum-producing fuel bundle (MPB), previously designed for CANDU reactors, has as its principal attribute that it is neutronically and thermal hydraulically equivalent to the standard 37-element fuel bundle typically used in CANDU reactors. Given that the typical irradiation time for MPBs is 20 days while the typical refueling period for a channel is on average 6 months, the refueling strategy needs to be adjusted to accommodate the shorter irradiation time of MPBs. This study evaluates a new refueling strategy suitable for employing the new MPBs in the core. A full-core, three-dimensional model is constructed in the diffusion code DONJON, and a fueling strategy for achieving the desired weekly yield of 99Mo is developed. The adequacy of the proposed refueling scheme is evaluated using a series of time-average calculations, which show that a small increase in the core reactivity (<0.4 mk) can be expected when irradiating a set of four MPBs in three different fuel channels in the inner region of the core. The small increase in the core reactivity can be managed by slightly increasing the discharge burnup in the non-MPB-bearing fuel channels, thus also improving slightly the fuel utilization in the reactor.

Long term reactivity operation for a high temperature testing reactor (HTTR) loaded with uranium nitride fuel

A High Temperature Testing Reactor (HTTR) is one of the fourth generation nuclear reactors, the reactor uses UO2 as typical fuel with different axial and radial enrichments. In the present research, uranium nitride (UN) is proposed as a fuel for the reactor for the same core enrichment distribution. The core multiplication factor, cycle length, discharge burnup are calculated and the results of UO2 fuel are compared with UN fuel results. Neutron spectrum, fissile isotopes behavior and Breeding capability are compared for the two fuel types. Kinetic parameters, Control rods and Burnable poisons worth are also evaluated. The results indicated that uranium nitride (UN) has a longer fuel cycle, higher breeding capability and harder thermal flux compared to UO2 fuel.

Capturing the run-in of a pebble-bed reactor by using thermal feedback and high-fidelity neutronics simulations

Modeling the run-in of a pebble-bed reactor (PBR) can be challenging as a result of changes in the power, fuel type, and temperatures that occur throughout the run-in period. Previous work utilized high-fidelity neutronics simulations or lower-fidelity coupled neutronics/thermal-hydraulics models to capture the general characteristics of the run-in process. The present work employs high-fidelity neutronics simulations (using Serpent) coupled with thermal-hydraulics simulations (using Griffin–Pronghorn) to capture the thermal feedback present during the run-in and approach to equilibrium for a PBR. Incorporating thermal feedback enables important distinctions to be made about conditions occurring inside the core,as the power distribution, discharge burnup, and isotopic compositions are all affected by the temperature distribution.

Practical Design of a Breed-and-Burn Lead-Cooled Fast Reactor with Rotational Fuel Shuffling Strategy

The purpose of this study is to demonstrate a practical core design for a lead-cooled, nitride fueled, rotational fuel shuffling breed-and-burn (RFBB) fast reactor. The core design is based on the Westinghouse Lead Fast Reactor (WH-LFR) and uses natural uranium nitride fuel with a sodium bond encased in oxide dispersion-strengthened steel cladding. Simulations confirmed the potential of the reactor to maintain criticality at the equilibrium state, with a reactivity swing of less than 200 pcm at every cycle interval and an average discharge burnup of 235 MWd/kg heavy metals (HM) for a 1050 effective full-power day refueling interval. Power profiles were maintained stable at the equilibrium state, while the cladding of the discharged fuel incurred over 650 displacements per atom over its entire residency in the core. From a nonproliferation perspective, the plutonium vector for the discharge fuel aligns with reactor-grade fuel standards, with over a 70% concentration of 239Pu and over 22% 240Pu, reducing the risk of weaponization. The adopted control rod system has been shown to offer sufficient negative reactivity of over 19 $ to bring the reactor into a subcritical state. Challenges such as the susceptibility of neutron balance to material thickness and neutron leakage have been addressed, emphasizing the necessity for meticulous design improvements. A steady-state thermohydraulic analysis confirmed the heat removal capacity from the hottest channel, ensuring operational safety. This study confirmed the feasibility of the RFBB strategy for a lead-cooled nitride-fueled fast reactor and sets a precedent for future research in enhancing fuel utilization and safety in nuclear reactors.

Potential of mixed oxide fuel for higher fissile conversion ratio in light water reactors

The prospect of adopting various mixed oxide fuels to increase the conversion ratio (CR) of conventional LWRs has been explored in this study. The Westinghouse AP1000 fuel assembly (FA) was taken as the reference design. While the geometry of the FA was kept the same, five assembly models were constructed utilizing different MOX fuels in the fuel rods. Depleted UO2 was mixed with PuO2 for the first model, the next two models featured a combination of UO2 with ThO2, the fourth one utilized a mixture of PuO2 and ThO2, and the final model employed a combination of UO2, PuO2, and ThO2. Depletion analysis of these fuel assembly models was then carried out using the lattice physics code Dragon 5. The burnup-dependent infinite multiplication factor and reactivity were determined and a significant improvement in cycle length was observed at the end of the burnup period because of fertile to fissile conversion. A higher fast fission rate was obtained for plutonium-based fuel models compared to the reference, whereas the thorium-based fuel models demonstrated higher thermal fission rates. The linear reactivity model (LRM) was utilized to calculate fuel cycle performance for a three-batch refueling scheme. The three fuel models utilizing PuO2 achieved improved discharge burnup and cycle length, and required lower charge fuel mass to obtain the same thermal power. All of the five proposed models attained higher conversion ratios compared to the reference, while the highest increment in CR was obtained for the last model utilizing all three oxides. The CR of conventional UO2 fuel in AP1000 was 0.64, whereas it was calculated to be 0.73 for Model 5. Key safety parameters including the fuel and moderator temperature coefficients of reactivity and the effective delayed neutron fraction of the models were also determined. Finally, the isotopic evolution of minor actinides and fission product poisons of the models were analyzed as a function of burnup and compared with the reference.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Fuel element microreactor integrating a square UO2 fuel rod with an internal heat pipe

In this study we propose the Fuel Element Micro Reactor (FEMR), with 308 square UO2 fuel rods, each one containing inside a heat pipe. We selected Mercury as the working fluid for the heat pipes and operate the microreactor at average temperature around 900 K or 637 °C. Heat pipes introduce empty regions into the core that increase neutron leakage and severely reduce core reactivity. Square fuel lattices allow greater fuel volume fractions and favor increasing core reactivity. The adoption of a 20 cm BeO reflector and Zircaloy as fuel cladding further increases core reactivity. The FEMR core thermal power is 2.3 MW, the power density distribution is rather flat with peaks occurring at the core-reflector interface and the peak fuel temperature is 1160 K. The reactor has a negative temperature isothermal coefficient of reactivity. The reactivity control system with 8 rotational drums and a central cruciform absorber rod meets the criteria of stuck rod and diversity of engineering principles. The average fuel discharge burnup is 8.96 MWd/kgU and the fuel cycle length without refueling is 8.7 years. The ratio between thermal power and total Uranium mass at beginning of life is 2.46 kW/kgU.

Prototyping of a Machine Learning–Based Burnup Measurement Capability for Pebble Bed Reactor Fuel

The pebble bed reactor is a unique reactor design due to its capacity for continuous multipass circulation of the fuel elements, without causing interruption to reactor operation, with the assistance of the burnup measurement system. Such a system necessarily requires an accurate knowledge of the burnup of each fuel pebble upon ejection from the core so as to inform the reloading decision and to ensure that no pebble exceeds the regulated discharge burnup limit at any point following reinsertion into the reactor core. In this work, we conceptualize, develop, and demonstrate a machine learning–based fuel burnup prediction framework leveraging advanced modeling and simulation capabilities. At its core, machine learning regression models are learned from simulated data to establish the correlation among the irradiated fuel composition (hence burnup), the gamma leakage spectrum, and the gamma spectroscopy results. Sensitivity analysis is conducted to quantify the impact of unknown design parameters, such as fuel enrichment, and irradiation environment, including power density, temperature, and neighboring materials, on the prediction accuracy of various supervised regression algorithms. The effects of a short cooldown period on machine learning prediction accuracy are also investigated. A test data set is used to validate that the data generation methodology proposed in this work successfully results in a machine learning model capable of interpolating its prediction of burnup onto a much wider range of irradiation conditions than were explicitly represented in the training database. The inclusion of a cooldown period of just 2 h leads to a prediction root-mean-square error of <5 MWd/kgU when the fuel enrichment is known and <9 MWd/kgU otherwise.

Neutronic analysis of high burnup thorium-HALEU fuels in PHWR

Recent developments have been made in a novel fuel for pressurised heavy water reactors: high burnup oxide fuels composed of a mixture of thorium and high assay low enriched uranium. Several claims are made regarding the performance of such fuels in terms of economics, waste and non-proliferation, among others. This article analyses this class of fuel using an infinite lattice study in SERPENT. First, a set of fuel compositions in the thorium/uranium enrichment design space that satisfy the high burnup target is produced. Then, a representative thorium-based fuel is compared to a uranium-only fuel with the same discharge burnup, as well as current standard fuel, in the context of the various performance claims.

Irregular, Backward-Compatible PWR Assembles with More Pins to Facilitate Uprates

Increasing the number of pins within a pressurized water reactor (PWR) assembly reduces pin temperature for a given assembly power. In conjunction with a core retrofit, this presents a potential route to PWR uprate, which is of growing interest given recent increases in electricity prices. However, most PWRs utilize regular lattice designs with fixed guide tube positions, such as the very common 17 × 17 lattice design with 25 guide/instrumentation tubes. These tubes are aligned with penetrations in the reactor pressure vessel, which presents a prohibitive obstacle to retrofit, and more widely, may “lock” many PWRs to this particular fuel configuration. In this paper, an irregular PWR fuel assembly is proposed. It is shown that a backward-compatible lattice with 324 fuel pins per assembly (BL324), uniform enrichment, and the same hydrogen-to–heavy metal ratio as a reference 17 × 17 assembly with 264 fuel pins can achieve within-assembly power peaking within 2% of the reference assembly under equivalent conditions while fixing the guide tube positions. Power peaking can be further reduced to reach that of the existing fuel assembly by reducing the enrichment of 36 of the pins by 0.2 wt%. The fuel assembly could potentially either support a significant uprate of up to ~20% in conjunction with low-enriched uranium plus (LEU+) fuel or a more aggressive cycle design, and hence, improved discharge burnup at the same power and batch strategy. A subchannel analysis shows that the coolant heat-up distribution is comparable to the reference assembly. However, the pressure drop is estimated to be 4% higher, which would challenge the performance of transition cores containing both 17 × 17s and BL324s. Further incremental changes to BL324 may be attractive, either to improve manufacturability or to slightly improve performance through formal optimization.

Sensitivity studies of PWR MOX fuel management to the plutonium initial vector using Artificial Neural Networks

This paper presents new metamodels based on artificial neural networks trained on full core 3D depletion simulations performed with APOLLO2 and CRONOS2. They are used to estimate the irradiation cycle length, discharge burn-up of each fuel assembly type and radial power factor of a PWR loaded with 30% of MOX fuels, as a function of the initial plutonium composition. They allow to explore the impact of the plutonium isotopic vector on the reactor characteristics and can be used for scenarios studied for future fuel cycle. Some exclusion domains in the plutonium isotopic vector phase space are identified as a function of the cycle length. As an example, the potentialities of such fuel management for plutonium recycling from MOX spent fuel are studied.

Preliminary conceptual design with neutronics and thermal-hydraulics assessments of FuSTAR

The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction.

An attempt to find the optimal loading pattern of minor actinides in the PWR fuel rods

Three approaches of minor actinide (MA) loading in the fuel rods of a pressurized water reactor were examined. They comprised inserting MAs and fuel mixture in the center of duplex fuel rods, coating of MAs and fuel mixture in the outer layer of fuel rods and lastly, mixing MAs homogeneously with the entire fuel rod elements. AP1000 fuel assembly was chosen as the reference design. Two fuel types namely conventional UO2 and UTh MOX fuel were selected as the base fuel models for the study. Although shorter cycle length than the UO2 fuel design was achieved by the UTh fuel design, UTh fuel exhibited more favorable FTC and yielded a lower plutonium stockpile at the end of the fuel cycle. When MAs were added in the outer layer of fuel rods, it resulted in higher rate of transmutation, however it had an adverse impact on the fuel reactivity, discharge burnup and pin power distribution. On the contrary, when MAs were loaded in the central region, it led to incomplete incineration of MAs. Therefore, based on the satisfactory transmutation rate, MTC, most negative FTC and uniform pin power distribution, the homogenous MA loading strategy can be concluded as the optimal loading pattern for the AP1000 nuclear reactor.

Fuel performance evaluation of two high burnup PWR core designs during normal operation, control rod withdrawal, and control rod ejection scenarios

There is interest among utilities to extend the current, 18-month operating cycle to 24 months. Economically, this extension would require greater than 5 % enrichment and peak rod average discharge burnup levels above 62 GWd/MTU. A notable challenge of increasing enrichment is the resulting additional excess reactivity encountered during the early stages of fuel life. To accommodate, burnable absorbers beyond soluble boron are introduced into the fuel system. In high burnup fuels, the possibilities of cladding lift-off and fuel melting increase due, in part, to increased rod internal pressures and limited fuel thermal conductivity, respectively. This work collaboratively employs PARCS, RELAP5-3D, and BISON to compare the fuel performance of two high burnup fuel candidates with higher than 5 % enrichment. The fuel performance parameters were compared to current NRC guidance. The results demonstrate an annular fuel design with homogenously blended gadolinium as a burnable absorber operates with greater safety margins during normal operation, allowing for additional operational flexibility. During normal operation, the core design utilizing Integral Fuel Burnable Absorber pins contained fuel pins which reached plenum pressures above 15.5 MPa by the end of the first fuel cycle and fuel pins experienced cladding hoop strains above 1 %. In the Gd core design, only two observed pins experienced plenum pressures above 15.5 MPa and no pins exceeded 1 % cladding hoop strain. During the control rod withdrawal scenario, plenum pressures for pins in both designs marginally exceeded system pressure, however neither experienced excessive hoop strain. The Gd core design experienced a maximum fuel temperature of 2418 K, which is significantly higher than the Integral Fuel Burnable Absorber design at 2157 K, but still within regulatory guidance. We predicted that the fuel in both could return to service after the CRW event. We also predicted that cladding would not fail during the Control Rod Ejection in either core design. Generally, the Integral Fuel Burnable Absorber core design performed with greater safety margin with regards to temperature during normal operation and the transient events. However, the Gd core design performed with greater safety margin regarding plenum pressure and hoop strain limits during normal operation and both transient events.

Multi-batch core design study for innovative small modular reactor based on centrally-shielded burnable absorber

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

Comprehensive feasibility assessment of cycle length extension for CNPGS Unit-3

The core cycle length is crucial parameter for effective and economic operation of a nuclear power plant. Nuclear industry has great emphasis on improving the feasibility and sustainability of nuclear power generation. The cost of the electricity can be reduced by increasing the cycle length without violating the safety margins. Various methods and strategies have been examined to extend a Pressurized Water Reactor (PWR) core cycle length with optimized loading patterns. The objective of this work is to analyse the feasibility of the system to extend the cycle length of Chashma Nuclear Power Generating Station (CNPGS) unit-3, currently operating in Pakistan from 12 months to ∼18 months. To achieve the objective, fuel enrichment is increased from 3.4% to 4.1% with Gd2O3 as burnable poisons to suppress the excess reactivity. The multi-cycle analyses showed promising results and the core cycle length has been extended to ∼18-month for CNPGS unit-3 keeping power peaking factor, temperature coefficients and discharge burnup within design limits. The method can be confidently applied at any cycle of the power plant. Additionally, fuel and moderator temperatures at average and hot channels at beginning of life (BOL) and the end of life (EOL) are predicted using the 1-D steady state thermal hydraulic analysis. The peak to average temperatures at BOL and EOL are found well below the reference value (1.6875).

High-fidelity simulations of the run-in process for a pebble-bed reactor

Pebble-bed reactors (PBRs) rely on a continual feed of fuel pebbles being cycled through the core. As a result, they require a “run-in” period in order to reach an equilibrium state. The run-in period for a PBR is a complex, time-dependent problem that requires the injection of new fuel, different types of fuel, and power increases. This complexity in the run-in makes it important to capture the physical processes in order to generate an accurate representation. The present work details the creation of a high-fidelity Monte Carlo methodology for analyzing the run-in and subsequent approach to equilibrium for PBRs. The methodology entails a Python module wrapped around Serpent so as to perform neutronics calculations, move pebbles, refuel the core, and discharge pebbles, thereby modeling the explicit behavior of the PBR run-in. Three run-in simulations (a constant temperature profile, a linear temperature profile, and a constant temperature profile using control rods) were examined in order to identify the key physical phenomena present in the run-in process. Utilizing kugelpy, we found the inclusion of a temperature profile to be important for accurately capturing a discharge burnup (around 141 MWd/kg), a consistent k-eff (around 1.005), and an average pebble power (around 2.5 kW/pebble) that all fall within acceptable limits.