Temperature Reactivity Feedback Research Articles

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g.for projects such as G4-STORK).G4-STORK derives reactor physics parameters (e.g.keff) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g.recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g.fuel temperature reactivity feedback).The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes.

Stiffness treatment of differential equations for the point reactor dynamic systems

An original methodology based on the Padé and Chebyshev rational approximations for the solution of the non-linear point kinetics equations with temperature reactivity feedback is described and investigated. Piecewise constant approximations of the reactivity and source function are made. The technique is enhanced by explicitly accounting for the feedback and the reactivity variation within a time step through an iterative cycle. An important feature of the Chebyshev rational method is that good numerical approximations to the solutions of the stiff coupled kinetics differential equations can be obtained in a single time step, as opposed to several time steps required for the conventional methods. The CPU time required for the Chebyshev rational method is less than that time required for the conventional method (Padé approximations) by 72.64%, which is one advantage of the presented method. The cases of approximations which combined with its A(α)-stability, leads to a better reduction of the errors when intermediate and large times are reached after series of small time steps if the inserted reactivity is positive and sufficiently large. Numerical studies are presented for different benchmark problems of various reactivity insertions, time varying reactivity and temperature feedback reactivity. The results confirm the theoretical analysis and indicate the range of applicability of the methods presented. The computational results indicate that the method is efficient and accurate.

The ATWS analysis of one control rod withdraw out of the HTR-10GT core in addition with bypass valve failure

The 10MW high temperature gas cooled test reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basic accidents of HTR10-GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism, in spite of the failure of bypass valve. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1283.3°C. It was a litter than 1230°C – the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation will be raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM for they have the same outlet temperature of the core.

Temperature coefficients calculation for the first fuel loading of NPP Mochovce 3–4

Due to the planned first criticality start-up of the Nuclear Power Plant (NPP) Mochovce units 3 and 4 in the near future, detailed analyses of core parameters are required to support safe operation of the nuclear facility. Since the previous calculations were performed by the deterministic nuclear codes, the stochastic approach was used to produce the independent results. The paper introduces determination of temperature feedback reactivity coefficients, especially summary (isothermal) and moderator (density) reactivity coefficients between 200°C and 260°C with step of 2°C. The evaluation of the critical boron acid (H3BO3) concentrations for different positions of the 6th control assembly group was performed for the given coolant temperatures as a precondition of the thermal coefficients calculation. All calculations were performed by computational code MCNP5 1.60 supported by NJOY99.364 microscopic cross sections processing system and our control scripts. For the validation of the criticality calculations (critical boron acid concentration), the experimental data based on the first critical conditions of the NPP Mochovce 2 start-up were used.

Confirmation of accuracy of generalized power series method for the solution of point kinetics equations with feedback

In this paper, generalized power series method (GPWS) is developed to obtain approximate solutions to point kinetics equations with feedback. The stiffness of the kinetics equations restricts the time interval to a small increment, which in turn restricts the traditional power series method (PWS) within a very small constant step size especially when the generation times are very small. The GPWS method has introduced time intervals that are much longer than time intervals used in the conventional numerical integrations like Generalized Runge–Kutta or power series methods, and it is thus useful in reducing computing time. Convergence of both the power series and the partial sums are discussed and the time step has been restricted within a circle of convergence by using the convergence conditions. Local truncation errors and some other constraints are used to produce the largest step size allowable at each step while keeping the error within a specific tolerance. The accuracy of the method is examined using five different cases of temperature reactivity feedback for step and ramp impressive reactivities with one and six groups of delayed neutrons. Supercritical (prompt and delayed) processes of a nuclear reactor with temperature feedback are discussed while inserting large and small reactivities. Results obtained by GPWS method attest the effectiveness the theoretical analysis, they demonstrate that the convergence of the iteration scheme can be controllable. The proposed method is accurate when compared to the analytical and numerical methods.

Solution of Point Reactor Neutron Kinetics Equations with Temperature Feedback by Singularly Perturbed Method

The singularly perturbed method (SPM) is proposed to obtain the analytical solution for the delayed supercritical process of nuclear reactor with temperature feedback and small step reactivity inserted. The relation between the reactivity and time is derived. Also, the neutron density (or power) and the average density of delayed neutron precursors as the function of reactivity are presented. The variations of neutron density (or power) and temperature with time are calculated and plotted and compared with those by accurate solution and other analytical methods. It is shown that the results by the SPM are valid and accurate in the large range and the SPM is simpler than those in the previous literature.

Monte Carlo Calculations Applied to SLOWPOKE Full-Reactor Analysis

Monte Carlo simulations are applied to the full-reactor analysis of the SLOWPOKE design. The temperature reactivity feedback calculated by using the MCNP code for either the high enriched uranium (HEU) or low enriched uranium (LEU) core is in good agreement with the experimental data, with a k-eff bias of +3.3 mk for a HEU core and +6 mk for a LEU core. Two methods that are based on existing third-party codes have been developed for use in core following: 1) MCNP (for the transport calculation) in conjunction with WIMS-AECL (for fuel burnup advancement), and 2) SERPENT (that combines both transport and burnup capabilities). Both methods show very good agreement with the experimental data for core excess reactivity and detailed power distributions versus burnup and reactivity shim.

Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

The molten salt reactor (MSR) is one of the Generation IV reactors. The fuel is dissolved in the carrier salt and circulates in the loop. The technologies are different from that in the solid-fuel reactors. In this work, the attention is focused on development of safety analysis tool for the MSRs. A single channel model, a heat transfer model, a heat sink model and a liquid-fuel point kinetic model that takes into account the effect of circulation are employed. The validation of this code is done with the experimental data of pump coastdown and startup in MSRE. The unprotected loss of heat sink (ULOHS), combination of ULOHS and unprotected loss of flow (ULOF) are performed on the Molten Salt Actinide Recycler and Transmuter (MOSART). This work aims to study temperature fluctuation, corresponding power change, effect of flow delayed neutron precursors and temperature reactivity feedback in transient accident to examine the inherent safety design of MOSART. The transient results reveal that the large negative temperature feedback coefficients guarantee MOSART inherent safety and the range of temperature is within the safety margin in case of combination of accidental events.

Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available.Currently, at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the DYN3D code is being extended and adopted for the application to block-type High Temperature Gas-Cooled Reactors (HTGRs). In this paper, we give an overview of the latest developments of DYN3D concerning block-type HTGR.The simplified P3 (SP3) transport approximation is implemented into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method previously implemented into DYN3D for square fuel element geometry of LWR is being extended for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes to be able to perform a systematic mesh refinement.One of the main challenges in cross section generation for the HTGR core calculations is the treatment of the so-called “double heterogeneity”. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. The main steps of the RPT method are described. The use of the method for the cross section generation of a simplified HTGR core including its verification is presented.A 3D heat conduction module coupled with a channel-type coolant flow model is implemented to take the temperature reactivity feedback to neutronics physically correctly into account. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Taylor’s series method for solving the nonlinear point kinetics equations

Abstract Taylor’s series method for solving the point reactor kinetics equations with multi-group of delayed neutrons in the presence of Newtonian temperature feedback reactivity is applied and programmed by FORTRAN. This system is the couples of the stiff nonlinear ordinary differential equations. This numerical method is based on the different order derivatives of the neutron density, the precursor concentrations of i-group of delayed neutrons and the reactivity. The r th order of derivatives are derived. The stability of Taylor’s series method is discussed. Three sets of applications: step, ramp and temperature feedback reactivities are computed. Taylor’s series method is an accurate computational technique and stable for negative step, negative ramp and temperature feedback reactivities. This method is useful than the traditional methods for solving the nonlinear point kinetics equations.

ICONE19-43309 ANALYSIS ON ATWS CAUSED BY EARTHQUAKE OF THE HTR-10GT

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700℃. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750℃ and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into a initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223℃. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 ℃. It was a little higher than 1230℃--the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620℃, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

Inhomogeneous point kinetics equations and the source contribution

Inhomogeneous point reactor kinetics equations with one-group of delayed neutrons are solved analytically for linear reactivity insertion as well as for step reactivity insertion in the presence of external neutron source using the prompt jump approximation. The solution is obtained as an infinite series. The methodology is found to be a promising tool for analyzing nuclear reactor kinetics with positive or negative ramp reactivity insertion on a sub-critical or a zero power delayed critical reactor, where the temperature reactivity feedback is negligibly small. To check the consistency and the accuracy of the analytical solution, the results are compared with the numerical solution for different sub-critical and delayed critical states. The comparison is found to be good for all kinds of positive and negative step and ramp reactivity insertions. The analytical solution is arranged into two terms, one as a function of source contribution the other without that. Using the newly rearranged solution, the importance of the source term and the contribution to the error while neglecting source term to the reactor kinetics analysis, can be realized. Contribution to the error is small (less than 0.1%) when the equilibrium power is more than about one megawatt for a medium sized LMFBR. Similarly, the importance of source contribution to the total reactor period as a function of initial equilibrium power is also realized with the newly rearranged analytical solution. The total reactor period is over predicted (larger period in place of smaller period) which is not conservative, if the source contribution is not considered, for considerably small initial equilibrium power. The percentage of error in not considering the source term in period calculation varies as a function of net reactivity and ramp rate. The percentage of error in period determination without considering the source is comparatively high for small ramp rates.

Dynamic Simulation of a Space Reactor System with Closed Brayton Cycle Loops

A dynamic simulation model for space reactor power systems with multiple closed Brayton cycle loops for energy conversion is developed and demonstrated for a startup transient. The simulated power system employs a submersion-subcritical safe space S ^ 4 reactor with a negative temperature reactivity feedback and has no single-point failures in reactor cooling and energy conversion. The S ^ 4 reactor core is divided into three hydraulically independent sectors, and each sector has a separate closed Brayton cycle loop that is thermal-hydraulically coupled to a circulating liquid NaK-78 secondary loop with two water heat pipe heat rejection radiator panels. Each closed Brayton cycle loop has a Brayton rotating unit designed and optimized for high thermal efficiency and low specific mass (1.16 kg/kW e ). The reactor coolant and the closed Brayton cycle working fluid is a He-Xe binary gas mixture with a molecular weight of 40 g/mol. Results are presented for a startup transient of the S ^ 4 closed Brayton cycle power system to full-power operation at a reactor thermal power 471 kW th , a Brayton rotating unit shaft speed of 45 krpm, and turbine and compressor inlet temperatures of 1149 and 400 K, respectively. At these conditions, the nominal electrical power and thermal efficiency of the power system are 130.8 kW e and 27.8%.

Performance and radiological analyses of a space reactor power system deployed into a 1000–3000 km earth orbit

This paper presents the results of performance and radiological analyses of a space reactor power system to support space-based, radar satellites in a 1000–3000 km orbit for global civilian air and ocean traffic control. The power system with six primary and secondary loops to avoid single point failures in reactor cooling and energy conversion employs a sectored, liquid NaK-78 cooled fission reactor that has a negative temperature reactivity feedback and SiGe thermoelectric energy conversion modules, nominally generating 37.1 kW e for up to 6 years. The reactor nominally generates 1183.6 kW th at an exit temperature of ∼992 K. Thermoelectric–electromagnetic (TEM) pumps circulate the liquid NaK-78 in the primary and secondary loops and passively remove the decay heat from the reactor core after shutdown. The system parameters during the startup and shutdown transients and nominal steady-state operation are calculated. The effects of the period of incrementally inserting external reactivity on the system parameters during nominal operation are also investigated. The radioactivity buildup in the reactor during nominal operation up to 6 years and the decay in radioactivity after shutdown and as a function of storage time in orbit up to 1000 years are calculated.

Dynamic Characteristics of Compact Fast-Spectrum Reactors

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

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx–Al) containing 4.40gU/cm3 of an MTR was replaced with silicide (U3Si–Al and U3Si2–Al) and oxide (U3O8–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38°C to 50°C and 100°C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8–Al was about 2% more than the original UAlx–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

Prospects of using different clad materials in a material test research reactor – Part 2 – The reactivity feedback coefficients

The reactivity feedback coefficients of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced with stainless steel-316 and zircaloy-4. Calculations were carried out to find the fuel temperature reactivity feedback coefficient, clad temperature reactivity feedback coefficient, moderator temperature reactivity feedback coefficient and moderator density reactivity feedback coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 38 °C to 50 °C, at the beginning of life, were maximum in magnitude for stainless steel-316 cladded fuel, followed by aluminum and least for the zircaloy-4 cladded fuel. The fuel temperature feedback coefficient increased in magnitude by 47.37% for stainless steel-316 and decreased by 4.72% for zircaloy-4 clad. The moderator temperature feedback coefficient increased in magnitude by 60.41% for stainless steel-316 and decreased by 3.03% for zircaloy-4 clad, while the moderator density feedback coefficient showed an increase in magnitude of 59.18% for stainless steel-316 and a decrease of 7.63% for zircaloy-4 clad. Zircaloy-4 gave a positive value for clad temperature feedback coefficient, while the others two did not have any clad temperature feedback coefficient.

Reactivity feedback coefficients of a material test research reactor fueled with high-density U 3Si 2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U 3Si 2 dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U 3Si 2 LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δ k/ k × 10 −5 K −1) −2.116 − 0.118 ρ U, 0.713 − 37.309/ ρ U and −12.765 − 34.309/ ρ U, respectively for 4.0 ≤ ρ U (g/cm 3) ≤ 6.0.

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S ∧4) reactor

The He–Xe gas-cooled, S ∧4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S ∧4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S ∧4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.

Reactivity feedback and control margins in natural uranium fueled and heavy water moderated nuclear research reactors

Three-dimensional neutronics analyses were performed to find the reactivity feedback coefficients and control margins under normal and presumed accidental conditions for the novel proliferation resistant, safer and economical natural uranium fueled and heavy water moderated nuclear research reactor cores [Annals of Nuclear Energy 31 (2004) 1331–1356 & 32 (2005) 612–620, Progress in Nuclear Energy (2006a,b) in press]. The results were compared with the reference core similar in design to National Experimental Reactor (NRX) and Canadian Indian Reactor (CIR). Standard reactor physics simulation codes WIMS-D/4 and CITATION were employed for this study. It was found that as opposed to the reference design, the new proposed cores are inherently stable as far as temperature reactivity feedback is concerned. Voids formation (loss of coolant) adds positive reactivity in all the cores. However, in the case of proposed cores, the reactivity added would be about 10% of that added in the reference core for loss of a given fraction of coolant. Moreover, as opposed to the NRX/CIR core, the shut-off rods, under the one stuck rod criterion, can shutdown and maintain the proposed cores subcritical in the accidental conditions.