Beginning Of Cycle Research Articles

Thorium application in the medium-sized sodium-cooled fast reactor

The report presents an analysis of the medium-sized Sodium-Cooled Fast Reactor (SFR) core with Thorium-based Mixed-Oxide fuel. The introduction of Transuranics (TRU) to the fuel was to allow long-lived nuclear waste incineration. The studied core is based on the modified Advanced Burner Reactor (ABR) 1000MWth core design, which was analysed in the OECD/NEA “Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core simulations with SERPENT 2.1.31 Monte Carlo computer code and ENDF library were performed, including static criticality and fuel burnup calculations for five fuel cycles. The core inventories at the Beginning of Cycle (BOC) and End of Cycle (EOC) were studied, and the impact of thorium fuel was assessed. The proposed core design is a burner reactor which uses thorium fuel. The excess core reactivity stays positive for long time despite large net consumption of transuranic elements as new fissile Uranium 233 is constantly breed from Thorium 232. Breeding of uranium allows longer fuel cycles.

Neutronic Study of Burnable Poison Materials and Their Alternative Configurations in Fully Ceramic Microencapsulated Loaded Pressurized Water Reactor Fuel Assembly

Use of FCM fuel in light water reactors is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. This work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and to keep the moderator temperature coefficient of reactivity (MTC) feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configurations including quadruple isotropic (QUADRISO), bi-isotropic (BISO), and Matrix Mix forms. The results obtained from the implicit random treatment of the double heterogeneity of tri-structural isotropic (TRISO), QUADRISO, and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5%. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at end of cycle (EOC); however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At the beginning of cycle (BOC), erbia-loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.

Neutronics analysis of CNPP-II loaded with 1/3rd MOX fuel

This paper performs the neutronics analysis of Chashma Nuclear Power Plant, Unit-II (CNPP-II) loaded with 1/3rd Mixed Oxide (MOX) fuel employing a burnup code Linkage of ORIGEN2.2 and OpenMC (LOOP). This is a PWR type reactor employed with UO2 fuel and is selected as a benchmark. The appropriate MOX composition which can give the same cycle length as that of pure UO2 core is found using linear reactivity model. The best suitable placement of MOX fuel is found to be at the periphery of the core. The important neutronics safety parameters for 1/3rd MOX fueled core at the Beginning of Cycle (BOC), Mid of Cycle (MOC) and End of Cycle (EOC) are compared with pure UO2 core of CNPP-II. The parameters include reactivity coefficients, control rods worth, shut down margin, peaking factors and effective delayed neutron fraction. It is observed that with slight decrease in the effective delayed neutron fraction, the MOX fuel can safely be used in the PWR type SMR reactor.

A deterministic method for PWR loading pattern optimization

A new deterministic approach to core loading optimization is presented. Selection of fuel enrichment and burnable poison (BP) arrangement in a pressurized water reactor (PWR) core are optimally determined subject to power peaking factor (PF) constraints. The deterministic approach is based on the method of Lagrange multipliers and the direct adjoining approach for the inequality power PF constraint. The optimality conditions are derived using calculus of variations resulting in the presence of jump conditions and a linear control Hamiltonian. The optimal control problems are addressed by simultaneously minimizing the Hamiltonian using the gradient method and performing Newton's method in different regions of the core to produce viable solutions for the controls. Without active control of the power profile in PWRs during depletion, our methodology is able to satisfy the power PF constraint at all times by proper loading selection at the beginning of cycle (BOC). This is achieved by promoting the proper burnup path forward in time during the calculation of the forward optimality conditions. This information is utilized in the adjoint burnup calculation, which acts to propagate the optimal control path back to the BOC during the calculation of the adjoint optimality conditions. The methodology has been developed into a new code DMCO (Deterministic Multi-Control Optimization) which can produce two-dimensional two-group depletion results in approximately seven minutes per optimization iteration on a personal computer. No application of heuristics is required and the code is capable of finding solutions from initial control distributions. A preliminary case study is presented for the Westinghouse AP600 first cycle fuel loading design which achieved an extended fuel cycle while satisfying power PF constraints.

Validation of LOOP through PWR measured data

This paper serves two purposes. First it demonstrates the validation of newly developed and verified burnup code, Linkage of ORIGEN and OpenMC (LOOP) against measured data of Post Irradiation Experiments (PIE) performed at Takaham-3 PWR. Secondly, it assesses the qualification of OpenMC model of Chashma Nuclear Power Plant unit-II (CNPP-II) PWR to predict different operating conditions of the reactor core. The LOOP code is then employed to perform the burnup simulations of CNPP-II first cycle with fresh UO2 fuel. The LOOP calculated results of power distribution of the first cycle at Beginning of Cycle (BOC), Middle of Cycle (MOC) and End of Cycle (EOC) are then compared with the operational data of the reactor. The maximum difference between the LOOP calculated radial and axial power distribution results and the measured data is 9.1% and 2.3% at BOC. At EOC, this difference further decreases to 7.2% and 1.5%. The depletion of burnable poison rods in the reactor was also compared with the FSAR predicted data which are in exact agreement with each other. The production of Cs-137 and Cs-134 was also investigated for the validity of LOOP burnup simulations.

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.

VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. These results provide confidence in VERA-CS's capability to perform high-fidelity calculations for practical PWR reactor problems.

Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

Analysis On The Change In Neutronic Parameters Due To Mispositioning Of Fuel In The Ap1000 Core

&lt;pre&gt;&lt;span&gt;AP1000 on the first cycle operation uses three types of UO&lt;sub&gt;2&lt;/sub&gt; fuel enrichments that are 2.35 w/o, 3.40 w/o and 4.50 w/o. To compensate excess reactivity, AP1000 uses an Integrated Fuel Burnable Absorber (IFBA) and a PYREX absorber as additional compensator to the boric solution in the moderator. IFBA is a burnable absorber made from ZrB&lt;sub&gt;2&lt;/sub&gt; which is integrated into the UO&lt;sub&gt;2&lt;/sub&gt; fuel. Human errors, such as fuel misposition, could happen when operators load fuel assemblies into the reactor core. For evaluating the design performance of AP1000, analysis on the change of neutronic parameters due to this fuel mispositioning need to be done. Analysis was performed on the reactor at hot zero power condition (HZP), beginning of cycle (BOC), and zero xenon&lt;/span&gt; &lt;span&gt;condition with several cases of mispositioning between two adjacent fuels. Neutronic parameters, mainly the k-eff and power factor distribution will be derived from SRAC2006 computer code module of CITATION. One of the inputs required is fuel lattice macroscopic cross-section data, which are generated by PIJ module. These calculations performed condensation energy group of 107 into 10 groups with JENDL - 3.3 library cross section data. &lt;/span&gt;&lt;span&gt;From the analysis, it can be concluded th&lt;/span&gt;&lt;span&gt;at misposition of the fuel in the first cycle of AP1000 core will result in very small change to the neutronic parameters. This very small change can not reduce the performance of the core.&lt;/span&gt;&lt;/pre&gt;

Comparative analyses of coated and composite UN fuel – Monte Carlo based full core LWR study

Longer continuous operation of a nuclear reactor leads to higher availability of nuclear power plant, entailing economic gain. In this context the option being explored, recently, is the use of higher density fuels as compared to current fuels i.e. UO2. Uranium mono nitride (UN) fuel is one of the options being explored in this regard. UN fuel has been used in nuclear industry for a long time with fast reactor option. More recently, studies have targeted its use in Light Water Reactor (LWR) environment. The main problem with using UN fuel in LWR is its potential reaction with water which produces hydrogen. Researchers have proposed different approaches to overcome this problem.One option is the use of coatings around UN fuel pellet to avoid the direct contact of water with fuel. The second option is use of a secondary phase (10 vol. percent) like ZrO2 which can make an oxide layer in case of contact with water and protect the main phase, Uranium mono nitride (80 vol. percent). Remaining 10 vol. percent is considered to be consumed by porosity. This study aims at comparison of neutron physics behavior of both these options in LWR environment. The upshot of the study is to quantify the impact of adding layers or secondary phase with respect to pure/complete UN fuel. To study these effects, two theoretical densities i.e. 95% and 80% for UN fuel are chosen for analyses. To avoid the problem of C-14 production from N-14, fuels studied are considered to be having 100% N-15.A validated model of Benchmark for Evaluation and Validation of Reactor Studies (BEAVRS) is used to perform all the analyses. Integral neutron physics parameters like neutron energy spectra, Radial Peaking Factors, Axial Peaking Factor, Doppler coefficient of reactivity, Isothermal Coefficient of reactivity and Temperature defect, Control Rod worth and excess reactivity for whole core are compared at Beginning of Cycle (BoC). Burnup obtained by different fuel option is also compared. Consistent with pin-cell based earlier findings, this full 3D analyses with UN based fuel shows noticeable spectral hardening leading to decrease in the value of control rod worth and less negative Doppler coefficient of reactivity while power peaking factors remain mostly unchanged. The economic advantage of switching to UN based fuels is expected when UN fuel above 80% TD is used. Approximately 19% increase in Equivalent Full Power Days (EFPDs) is witnessed by using 95% TD UN based fuels.

Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron adsorbent rods. 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 nuclear reactor. External coupling 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 the core in various 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. An external coupling 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 for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Burn up Effects on Kinetic Parameter’s Variations in an Electron Accelerator Driven Subcritical Reactor

One of the key components in nuclear reactors is calculation of the neutronic parameters and the burnup effect on the variation of these parameters. In this study, neutronic parameters such as neutron flux and power distributions, effective multiplication factor (Keff), neutron mean generation time (Λ), effective delayed neutron fraction (βeff) and fuel burn up are calculated in an Accelerator Driven Subcritical ALMR reactor to transmutation of TRUs by MCNPX code for the parabolic spatial distribution of electron beam. The results this study shows that with increasing in the fuel burn up the values of Keff and TRU decrease while the fission products rate increase. In addition, it observes that in comparison with the beginning-of-cycle (BOC) values, at end-of-cycle (EOC), the Λ parameter increases but the βeff parameter more or less constant as a function of burn up in the multiplication factor of 0.93. Also, according to the results, the average relative difference between the results of calculations with reference data is within 3%.

ICONE23-1611 NUMERICAL INVESTIGATION OF SUPERCRITICAL FLOW IN A VERTICAL TUBE UNDER NON-UNIFORM HEAT FLUX

We report the study of supercritical flow of carbon dioxide in a vertical tube under non-uniform heat flux. The investigation of supercritical fluid is crucial for advanced generation IV (GIF) nuclear reactors since both supercritical carbon dioxide and supercritical water are planned to be used in those designs. The supercritical carbon dioxide is going to be used in secondary loop of liquid metal fast reactors, the supercritical water will be used as a coolant itself. Thanks to scaling parameters for fluid-to-fluid modelling at supercritical conditions based on inlet conditions approach it is possible to predict response for different supercritical fluids at different scaled conditions. Hence, the results can be used for investigation of other supercritical fluids if it is needed. Since, the heat flux in particular fuel element varies axially it is imperative to investigate the influence of non-uniform flux in this direction. Current literature provides experimental results only when uniform heat flux is applied, whereas non-uniform heat flux can show the thermal response at different stage of fuel cycle. Hence, it is possible to simulate how the flow behaves from the beginning of cycle (BOC) when the shape of the curve is a cosine distributed to the end of cycle (EOC) moment when the peaking factor is skewed to the bottom of the fuel element. The numerical investigation is performed for a 2-D axisymmetric model of a tube with use of CFD code. First, the model is validated for a case with uniform heat flux against results found in the literature. Then, non-uniform heat flux is represented with two parameter equation to describe the variation along axial direction. Sensitivity study related to influence of pressure, temperature and average heat flux in order to better understand the phenomena are conducted. Obtained results provide information about heat transfer coefficient (HTC), heat transfer deterioration (HTD) in supercritical conditions and which parameters can influence it. With better understanding of this phenomena it is possible to prevent or to avoid it, since HTD may cause overheating of the fuel elements inside reactor core and lead to an accident.

Evaluation of Effective Prompt Neutron Lifetime for Pressurized Water Reactors

The lifetime of prompt neutrons is a basic characteristic of reactors since it determines the neutron kinetics of the reactor in all transient processes. This paper focuses on calculation of the prompt neutron lifetime for pressurized water reactors (PWRs). The calculation was performed using two independent methods. The first method uses the fundamental definition of the neutron lifetime with adjoint weighting that has recently been included in MCNPX. The second method is the 1/v absorber insertion method, where a 1/v absorber such as 10B is placed uniformly throughout a nuclear reactor and the change in reactivity is calculated. This prompt neutron lifetime is then extracted from the changes in the reactivity as the 10B concentration approaches zero.The results of the two methods are compared together at two points in the operation cycle [at beginning of cycle (BOC) and at end of cycle (EOC)]. The values of the prompt neutron lifetime as calculated with MCNPX are compared to values calculated with another independent method, and the results are in reasonable agreement with each other. Also, these results compared with the PWR final safety analysis report show good agreement. In the two methods of calculation, the prompt neutron lifetime was determined to be longer at EOC when compared to that at BOC.

Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor keff are 6.31mk for the SCWR at the beginning of cycle (BOC) and 6.99mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0mk and 0.9mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium.

Study of power distribution in the CZP, HFP and normal operation states of VVER-1000 (Bushehr) nuclear reactor core by coupling nuclear codes

In this research, the simulation of one-sixth of VVER-1000 (Bushehr) reactor core is carried out by WIMS-D4 nuclear code, based on symmetry of core and also by information obtained from FSAR. The cross sections of some nuclides are obtained by WIMS-D4 from the beginning of cycle (BOC) to the end of cycle (EOC), and they are transferred into the CITATION code as inputs. In the next stage, the amounts of neutron fluxes and power of reactor core are obtained by CITATION code in the CZP and HFP states. Then, the received products are returned again into the extended program cycle, thereby distributions of neutron fluxes and power are finally depicted. In the meantime, the space distribution of neutron fluxes and power throughout the core are presented during the normal operation by this simulation. It can be inferred that if the reactor operation continues, a flat power distribution will be made in the reactor core that might cause maximum power.

A neutronic pin cell and full-core design analysis of the 4S reactor

The 4S (Super-safe small, and simple reactor) core, based on the Toshiba Gen-IV 10MWe fast reactor design, is modeled and simulated with the Monte Carlo code MCNP to obtain useful neutronics design information. This paper carries out a preliminary pin-cell simulation to obtain the neutron spectrum, the infinite multiplication k∞ and the effect of temperature and moderator density followed by a full-core simulation for the effective multiplication factor keff with a central absorber control rod. Two absorber materials are considered viz boron carbide (B4C) and hafnium (Hf) to determine the fuel inventory to balance the beginning of cycle (BOC) excess reactivity. It is found that for a single central absorber rod of radius 9cm, the maximum fuel rod radius is 6mm for hafnium and 6.2mm for B4C. This work is of use for the core design of small multipurpose Gen-IV reactors.

IRRADIATION DEVICE FOR IRRADIATION TESTING OF COATED PARTICLE FUEL AT HANARO

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about 1.7×10 21 n/cm 2 , PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

Studi Sensitivitas Fraksi Packing Partikel TRISO dalam Desain Kritikalitas HTR Pebble Bed

HTR is a high temperature reactor used for electricity production and process heat applications such as hydrogen production, desalination of sea water, enhanced oil recovery and so on. HTR is designed based on the utilization of TRISO fuel particles that can prevent strongly the escape of fission products even at temperatures above 1600 o C. TRISO particles packing fraction is one of four key parameters that are essential in HTR design besides radius of the kernel, kernel density and fuel enrichment. This paper discusses the sensitivity of TRISO particles packing fraction that impacts to the loading of uranium in the fuel pebble, the long cycle of reactor operation and achievable maximum fuel burn-up. With the capability of Monte Carlo transport code MCNP5, all components of the reactor, starting from TRISO particles, were modeled in detail and explicit and calculated using the continuous energy nuclear data library ENDF/B-VI. The results show that the value of effective multiplication factor ( k eff) has a tendency to increase with decreasing particle TRISO packing fraction and to decrease with increasing fuel burn-up. K eff values decrease with increasing TRISO particle packing fraction both at the beginning of cycle (BOC) and at the end of cycle (EOC). Reactivity swing is also very sensitive on the TRISO particles packing fraction. From the analysis, it can be concluded that TRISO particles packing fraction greatly affects the neutronics performance of HTR pebble bed design. Packing fraction can change the effective multiplication factor ( k eff) and the swing reactivity with similar behavior.