Void Reactivity Coefficient Research Articles

Conceptual Engineering of CARA Fuel Element with Negative Void Coefficient for Atucha II

Experimentally validated void reactivity calculations were used to study the feasibility of a change in the design basis of Atucha II Nuclear Power Plant including the Large LOCA event. The use of CARA fuel element with burnable neutronic absorbers and enriched uranium is proposed instead of the original fuel. The void reactivity, refuelling costs, and power peaking factors are analysed at conceptual level to optimize the burnable neutronic absorber, the enrichment grade, and their distribution inside the fuel. This work concludes that, for the considered plant conditions, either a void reactivity coefficient granting no prompt critical excursion on Large LOCA or negative void reactivity is achievable, with advantages on refuelling cost and linear power density.

Nonlinear stability analysis of a reduced order model of nuclear reactors: A parametric study relevant to the advanced heavy water reactor

In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.

Improvements in the procedure of determining the void reactivity coefficient of a reactor of the RBMK type

An alternative method of processing experimental data in measurement of the void reactivity coefficient of an RBMK reactor is discussed. The method proposed allows one to reduce substantially the dependence of measurement results on the sign of perturbation of the feedwater flow rate and makes it possible to reduce many times the relative error of measurement of the void reactivity coefficient.

Design & construction of a digital real time reactivity meter for Tehran research reactor

Reactivity measurement is one of the challenges of monitoring, control and investigation of nuclear reactors. In this paper design and construction of a reactivity meter for continuous monitoring of reactivity in research reactors are described. The device receives amplified output of the fission chamber, which is in mA range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in Tehran research reactor (TRR) are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. The maximum relative error in several experiments is calculated to be 13%.

Reactivity feedback calculation of a conceptual TRISO fueled compact PWR core

Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together.

Core and sub-channel analysis of SCWR with mixed spectrum core

The SCWR core concept SCWR-M is proposed based on a mixed spectrum and consists of a thermal zone and a fast zone. This core design combines the merits of both thermal and fast SCWR cores, and minimizes their shortcomings. In the thermal zone co-current flow mode is applied with an exit temperature slightly over the pseudo-critical point. The downward flow in the thermal fuel assembly will provide an effective cooling of the fuel rods. In the forthcoming fast zone, a sufficiently large negative coolant void reactivity coefficient and high conversion ratio can be achieved by the axial multi-layer arrangement of fuel rods. Due to the high coolant inlet temperature over the pseudo-critical point, the heat transfer deterioration phenomenon will be eliminated in this fast spectrum zone. And the low water density in the fast zone enables a hard neutron spectrum, also with a wide lattice structure, which minimizes the effect of non-uniformity of the circumferential heat transfer and reduces the cladding peak temperature. The performance of the proposed core, including the neutron-physical and thermal-hydraulic behavior in sub-channel scale, is investigated with coupled neutron-physical/thermal-hydraulic simulation tools, which at the same time enables multi-scale analysis. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a multi-channel code and the neutron-physical performance is computed with a 3-D diffusion code. Based on the core results, the pin-power reconstruction is carried out for each fuel assembly to predict the local pin-power distribution. Moreover, the sub-channel calculation is performed to obtain the thermal-hydraulic parameters for each sub-channel and fuel rod. Based on the coupled analysis, measures to improve the performance of the SCWR-M core design are proposed and evaluated in this paper. The results achieved in this paper have shown that the mixed spectrum SCWR core concept (SCWR-M) is feasible and promising. One reference SCWR-M design is proposed for future analysis.

Stability Analyses of the Uprated BWR Using the Large Assembly with Small-Pins Concept

The Large Assembly with Small Pins (LASP) concept is an evolutionary fuel design proposed to enable a higher power density in boiling water reactors while maintaining the same operating conditions, such as power-to-flow ratio, core inlet conditions, and fuel-to-moderator ratio. It is based on replacing four traditional assemblies and the large water gap regions between them with a single large assembly having a 22 × 22 square fuel pin lattice. Twenty-five water rods within the assembly help maintain neutron moderation and accommodate as many finger-type control rods. It was previously shown that the LASP core allows operation with 20% higher power density than the core with traditional 9 × 9 fuel assemblies. However, the void reactivity coefficient of the LASP core is 25% more negative. In this study, the stability performance of the LASP core has been evaluated.The characteristics of density wave oscillations in the LASP core and their sensitivity to the operating parameters have been investigated. Although the perturbation decay ratios for the LASP core were found to be greater than those of the reference core, the stability criteria are sufficiently satisfied. Sensitivity studies were performed on the effects of design and operating parameters. It can be concluded that the LASP and the reference core have similar sensitivity to operating parameters. Furthermore, the calculated decay ratios were much smaller than the stability criterion for all the considered parameter ranges.

Safety Evaluation of the Uprated BWR Using the Large Assembly with Small Pins Concept

An advanced design of a Large Assembly with Small Pins (LASP) has been proposed at the Massachusetts Institute of Technology to increase the power density of boiling water reactors (BWRs) while keeping most of the operating conditions of current BWRs. LASP is based on replacing four traditional assemblies and the large water gap regions with a single large assembly having a 22 × 22 square lattice. In-assembly water rods accommodate control rods as well as provide help to the moderation of neutrons. Previous steady-state analysis showed that the LASP core allows for operation with 20% higher power density than the core with traditional 9 × 9 fuel assemblies. However, the void reactivity coefficient of the LASP core is 25% more negative and the steam flow rate is 20% higher than that of the reference core. In this study, the performances of the LASP core and reference core are compared for selected design-basis accidents and transients. Generally, the LASP design is found to behave in a manner similar to the traditional assemblies. First, the clad peak temperature during a large-break loss-of-coolant accident analysis satisfies regulatory criterion, and it is possible to preserve peak cladding temperature margin of the reference design if the capacity of the low-pressure core injection system is increased by 20%. Second, the generator load rejection with bypass failure and feedwater controller failure analyses show a decrease in dryout margin for the LASP core because of the combination of more negative void coefficient and increased steam load. However, this problem could be remedied by increasing the steam line flow area or allowing an additional flow restrictor in the steam line to attenuate the back propagating pressure wave in the main steam pipe following the turbine stop valve closure. Finally, the LASP core preserved the same level of margin to dryout as the reference core in the cases of four other evaluated events.

Coupled thermal-hydraulics and neutron-physics analysis of SCWR with mixed spectrum core

The new SCWR conceptual design (SCWR-M) is proposed on the basis of a mixed spectrum core consisting of a thermal spectrum zone and a fast spectrum zone. This new core design is considered to be the hybrid of the existing thermal SCWR and fast SCWR cores. It combines the merits of both thermal and fast SCWR cores, at the same time minimizes their shortcomings. For the thermal zone, the difficulties in the mechanical design and the maximum cladding temperature can be reduced as far as possible by the co-current flow mode; and for the fast zone, a sufficiently large negative coolant void reactivity coefficient and breeding ratio can be achieved by the multi-layer arrangement of fuel rods. The performance, including the burn-up behavior, of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. The results obtained so far have shown that the mixed spectrum SCWR concept (SCWR-M) is feasible and promising.

Small PWR core design with coated particle based fuel with a novel composition

Abstract A conceptual design study for a small-pressurized water reactor (PWR) core has been presented in this research paper that utilizes carbon coated fuel particles with a novel composition. The utilization of novel tristructural-isotropic (TRISO) fuel particles can completely eliminate the use of soluble boron system and requirement of burnable poison if adequate number of control rods is used. Apart from the simplification of the design by excluding such as complex systems from the design, the utilization of TRISO fuel particles in PWR technology can enhance the design integrity by confining the radioactive fission products within fuel itself during reactor operation. Reactor simulation computer codes WIMS-D/4 and CITATION were used for neutronics design and RELAP5 was used for thermal hydraulic analysis. The designed core is small which has only 25 MWth power out put with more than 2 EFPY operating life. The values of Doppler, moderator and void reactivity coefficients were found negative. The Soluble Boron Free (SBF) and Burnable Poison Free (BPF) concepts are more viable in small and medium size reactors (SMRs) because it makes the plant simpler and uniform burnup can also be achieved throughout the core life.

Exploring new coolants for nuclear breeder reactors

Abstract Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the cancelled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F 2 Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive RD the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F 2 Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients, in the case of both critical and subcritical reactors.

Study on Characteristics of Void Reactivity Coefficients for High-Conversion-Type Core of FLWR for MA Recycling

The investigation on the characteristics of void reactivity coefficients for the high-conversion-type core of the FLWR (HC-FLWR) concept for MA recycling has been performed. Void reactivity coefficients are the major restrictions for the core design of HC-FLWR for MA recycling, because the loaded MA makes void reactivity coefficients worse. Therefore, it is important to investigate the characteristics of void reactivity coefficients as a mechanism of reactor physics. Thus, in this study, the investigation of void reactivity coefficients has been performed using the exact perturbation calculations. In the exact perturbation theory, the reactivity is related to the variation in the cross section, and divided into scattering, leakage, absorption, and fission terms. Then, the worsening of the void reactivity coefficient caused by the MA loading mainly via the scattering term is found. Moreover, the void reactivity coefficient becomes better via the scattering term for the smaller fuel rod diameter, and via the leakage term for the lower core height. In addition, the 100% void reactivity coefficient, which is the restriction for the core design of HCFLWR for MA recycling, cannot be negative only by using the effect of the scattering term by reducing the fuel rod diameter. Therefore, the mechanism of achieving the negative 100% void reactivity coefficient by using the effect of the leakage term through the core height reduction is quantitatively verified.

Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal–hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U–ZrH 1.6 fuelled PWR cores are investigated and compared against those of the currently used UO 2 fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k ∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U–ZrH 1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 v/ o of the ZrH 1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.

Void reactivity coefficient benchmark results for a 10 × 10 BWR assembly in the full 0–100% void fraction range

A boiling water reactor SVEA-96+ fresh fuel lattice has been used as the basis for a benchmark study of the void reactivity coefficient at assembly level in the full voidage range. Results have been obtained using the deterministic codes CASMO-4, HELIOS, PHOENIX, BOXER and the probabilistic code MCNP4C, combined for almost all cases with different cross section libraries. A statistical analysis of the results obtained showed that the void reactivity coefficient tends to become less negative beyond 80% void and that the discrepancies between codes tend to increase from less than 15% at voidages lower than 40% to more than 25% at voidages higher than 70%. The void reactivity coefficient results and the corresponding differences between codes were isotopically decomposed to interpret discrepancies. The isotopic decomposition shows that the minimum observed in the void reactivity coefficient between 80% and 90% void is largely due to the decrease in the relative importance of the 157Gd(n, γ) rate with increasing voidage, and that the fundamental discrepancies between codes or libraries are mainly governed by the different predictions of the 238U(n, γ) variation with voidage.

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

Study on Characteristics of Void Reactivity Coefficients for High-Conversion-Type Core of FLWR for MA Recycling

The investigation on the characteristics of void reactivity coefficients for the high-conversion-type core of the FLWR (HC-FLWR) concept for MA recycling has been performed. Void reactivity coefficients are the major restrictions for the core design of HC-FLWR for MA recycling, because the loaded MA makes void reactivity coefficients worse. Therefore, it is important to investigate the characteristics of void reactivity coefficients as a mechanism of reactor physics. Thus, in this study, the investigation of void reactivity coefficients has been performed using the exact perturbation calculations. In the exact perturbation theory, the reactivity is related to the variation in the cross section, and divided into scattering, leakage, absorption, and fission terms. Then, the worsening of the void reactivity coefficient caused by the MA loading mainly via the scattering term is found. Moreover, the void reactivity coefficient becomes better via the scattering term for the smaller fuel rod diameter, and via the leakage term for the lower core height. In addition, the 100% void reactivity coefficient, which is the restriction for the core design of HCFLWR for MA recycling, cannot be negative only by using the effect of the scattering term by reducing the fuel rod diameter. Therefore, the mechanism of achieving the negative 100% void reactivity coefficient by using the effect of the leakage term through the core height reduction is quantitatively verified.

New Estimation Method for Void Reactivity Coefficient Using the TRACY Transient Data

The new estimation method for void reactivity coefficient using the measured transient data of the power and pressure was proposed. The present estimated void reactivity coefficient is less than the calulated values because of the distribution of gas voids is concentrated in the center region of the core due to the distribution of neutron flux

Breeding Capability and Void Reactivity Analysis of Heavy-Water-Cooled Thorium Reactor

The fuel breeding and void reactivity coefficient of thorium reactors have been investigated using heavy water as coolant for several parametric surveys on moderator-to-fuel ratio (MFR) and burnup. The equilibrium fuel cycle burnup calculation has been performed, which is coupled with the cell calculation for this evaluation. The η of 233U shows its superiority over other fissile nuclides in the surveyed MFR ranges and always stays higher than 2.1, which indicates that the reactor has a breeding condition for a wide range of MFR. A breeding condition with a burnup comparable to that of a standard PWR or higher can be achieved by adopting a larger pin gap (1–6 mm), and a pin gap of about 2mm can be used to achieve a breeding ratio (BR) of 1.1. A feasible design region of the reactors, which fulfills the breeding condition and negative void reactivity coefficient, has been found. A heavy-water-cooled PWR-type Th-233U fuel reactor can be designed as a breeder reactor with negative void coefficient.