Void Reactivity Research Articles

The use of zirconium hydride blankets in a minor actinide/thorium burner sodium-cooled reactor for void coefficient control with particular reference to UK's plutonium disposition problem

The use of zirconium hydride (Th–ZrH1.6) blankets in a thorium-fuelled sodium-cooled reactor for void reactivity control with particular reference to UK's plutonium disposition problem is proposed and considered. It is shown that, with the use of such blankets, a mild moderation effect is produced during voiding which compensates for the general hardening of the spectrum, enabling a net negative void coefficient at pin level to be attained without the need to rely on traditional neutron leakage enhancement techniques or neutron poisons, and with negligible impact on transmutation capabilities. One important difference in comparison with the traditional methods is that the void coefficient is obtained at the pin level, eliminating or mitigating substantially the spatial dependencies on the location of the void. Combining the use of such blankets with a suitable n-batch fuelling scheme yields a negative void reactivity coefficient throughout the life of fuel. Additional research and development are required to explore further this concept's potential.

Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR

The natural circulation boiling type SMR can experience flow instability during the startup transients due to the void reactivity feedback. A BWR-type natural circulation test loop has been built to perform the nuclear coupled startup transient tests for Purdue Novel Modular Reactor (NMR). This test loop is installed with different instruments to measure various thermal hydraulic parameters. The testing process can be monitored and controlled through PC with the assistance of LabVIEW procedure. The effects of power ramp rate on the flow instability during the nuclear coupled tests were investigated by controlling the power supply based on the point kinetics model with coolant void reactivity feedback. Two power ramp rates were investigated and the results were compared with those of the thermal hydraulic startup transients without void reactivity feedback. The time trace of power supply, system pressure, natural circulation rate, and void fraction profile are used to determine the flow stability during the transients. The results show that nuclear coupled startup transients also experience flashing instability and density wave oscillations. The power curves calculated from point kinetics model for startup transients show some fluctuations due to void reactivity feedback. However, the void reactivity feedback does not have significant effects on the flow instability during the startup procedure for the NMR.

Single pass core design for a Super Fast Reactor

Single pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward coolant flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern significantly simplifies the core concept in terms of upper core structure, coolant flow scheme as well as refueling procedure. In single pass core design, supercritical-pressure water at approximately 25.0MPa enters the core at 280°C and heated up in one through upward flow to the average outlet temperature at 500°C. Great coolant density change in vertical direction will cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO2 also introduces great power increase in blanket assemblies during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of blanket assemblies to mitigate the power change in blanket assembly and to increase the coolant outlet. With exclusive refueling and shuffling scheme as well as partial length control rod design, neutronics & TH coupled calculation shows satisfactory results that can fulfil the requirement of design for both 500°C outlet temperature and negative coolant void reactivity.

Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

Abstract An analysis of the influence of addition of minor actinides (MA) to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL) state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW) decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Study on the recently proposed direct S–CO2 and helium-cooled GCFR: Size of vessel, power conversion, and reactivity coefficients

This article contains a review on the history, progress, and objectives of the GCFR as well as a review on the newly proposed TID-type and block-type fuels of a GCFR, together with a research on their selected physical parameters. We have roughly estimated and compared size of core vessel of a suggested GCFR for S–CO2 and helium cooled as well as power conversion system efficiencies for a direct super-critical CO2 cooled cores. Moreover, the Coolant Void Reactivity (CVR) and Doppler coefficient for a reference GCFR with two different coolant types (supercritical CO2 and helium) and either different matrix materials (blended: fuel + diluents) in the block type core is calculated. MCNP 5.0 was used for our neutronic simulation and MAKXSF module was used to Doppler broaden the cross sections. Finally, the burn up comparison for two coolants and also for suggested diluents is also performed herein.

Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

Seed and blanket ADS using thorium–reprocessed fuel: Parametric survey on TRU transmutation performance and safety characteristics

Conceptual designs of accelerator driven systems (ADS) that utilize thorium fuel as blanket and reprocessed fuel as seed for fission reaction in order to transmute the transuranic elements in spent nuclear fuel and produce energy from thorium utilization was proposed. In order to analyze the TRU transmutation performance of the system while the safety margin during the operation is satisfied, a parametric survey is done in this study. The impact of minor actinide (MA) amount and the reactor size on transmutation efficiency and safety characteristics of the ADS is investigated using MCNPX code using the ENDF/B-VII. As the results, with increasing MA content, TRU transmutation rate is increased while void reactivity of the system becomes less negative. Besides, increasing the core size by introducing more thorium and reprocessed fuel assemblies into the system reduces the TRU transmutation rate and reactivity swing due to burnup of the system. In order to match the efficient transmutation purpose and safety features during operation, higher MA content is suggested to introduce into the system.

Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2and MOX fuels. UO2fuel enrichment is increased above existing design limitations for LWR fuels (&gt;5%). MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.

Primary Breeding Ratio Analysis of an Improved Supercritical Water Cooled Fast Reactor

The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and fuel nuclide components to the breeding ratio are analyzed. In assembly parameters, the seed fuel rod diameter has higher sensitivities to the conversion ratio (CR) than the coolant tube diameter in blanket. Increasing heavy metal fraction is good to CR improvement. The CR of SCFR also increases with a reasonable core arrangement and Pu isotope mass fraction reduction in fuel, which can achieve more negative coolant void reactivity coefficient at the same time. The breeding ratio of SCFR is 1.03128 with a new core arrangement. And the coolant void reactivity coefficient is negative, which achieves a fuel breeding in initial fuel cycle.

Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Abstract Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B4C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes.In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly.The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor

Abstract To cope with increasing spent fuel disposals and limited domestic spent fuel storages in Korea, the Korea Atomic Energy Research Institute has developed an advanced sodium-cooled fast reactor for TRU transmutation with an electricity output of 600 MWe (called the KALIMER-600 TRU burner). The design philosophy of the KALIMER-600 TRU burner concept is highly focused on inherent safety mechanisms, i.e., passive responses to abnormal and emergency conditions, and thereby minimizes the need for engineered safety systems. Accordingly, the main concern is on the sodium coolant void reactivity, a very important safety parameter of the KALIMER-600 TRU burner. This study was therefore performed to analyze the sodium void reactivity of the KALIMER-600 TRU burner to the finest resolution possible, e.g., contributions from any isotope in each core region. Such detailed analysis could be valuable and applicable to further optimization of the core passive safety characteristics against severe coolant voiding accident conditions.

Alternative reflectors for a compact sodium-cooled breed-and-burn fast reactor

The breed-and-burn fast reactor (B&BR) is a unique fast reactor concept that offers attractive characteristics in terms of core performance and non-proliferation aspects. The B&BR has the ability to breed its own fuel and use it in situ to achieve an extremely long life. In order to achieve the breed-and-burn condition, the neutron economy should be very good. In this work, a compact sodium-cooled B&BR is investigated from the physics point of view. In a compact size B&BR which usually has a higher neutron leakage, a good neutron reflector is essential to maintain the neutron economy. In the conventional sodium-cooled B&BRs, a steel reflector such as HT-9 is usually adopted as reflector material. In the current work, various alternative reflector materials such as pure lead, lead–bismuth eutectic (LBE), lead–magnesium eutectic (LME), PbO, MgO, Ni, and HT-9 are investigated from the neutronics perspectives. Several characterizations including the reflecting performance, core neutron spectrum, core leakage, power distribution, and sodium coolant void reactivity (CVR) have been performed to understand the behavior of each reflector material. The impact of the coarse lattice reflector configuration on the reflector performance and CVR has also been analyzed. In addition, the impact of the reflector material on the gas expansion module (GEM) worth has been investigated, too. Finally, it was concluded that lead-based reflectors such as LBE, LME, and PbO, are promising alternative reflector candidates for a compact B&BR. The calculations were all done by using a continuous energy Monte Carlo code.

Moderator Displacers for Reducing Coolant Void Reactivity in CANDU Reactors: A Scoping Study

When the coolant is voided in the lattice of a Canada deuterium uranium (CANDU) reactor, the net reactivity change is positive, due primarily to the fact that the coolant and moderator are separated and the coolant volume is much smaller than the moderator volume. The modest loss in moderation occurring when coolant is lost is not sufficient to offset the positive reactivity contributions of increased fast fission rate and reduced epithermal absorption. A way to achieve a negative net reactivity effect on coolant voiding is to increase the importance of moderation in the coolant by decreasing the moderator-to-coolant volume ratio. This work proposes reducing the moderator-to-coolant volume ratio in existing CANDU reactors by packing the moderator with displacers in the shape of hollow spheres in a close-packed pattern. Several materials and shell thickness values are investigated for different fuel enrichments. Calculations are performed using the lattice code DRAGON. Results show that it is possible to reduce the coolant void reactivity in a CANDU lattice with spherical moderator displacers arranged in a hexagonal closed-packed array, albeit at a cost in discharge burnup.

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.

Direct reuse of spent nuclear fuel

In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO2 fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the “finite neutron multiplication factor”, k*, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and k* on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty.The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

Heterogeneous Cores for Implementation of Thorium-Based Fuels in Heavy Water Reactors

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWRs) are highly advantageous for implementing thorium-based fuels because of their high neutron economy and online refueling capability. The use of heterogeneous seed/blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize fissile utilization (FU) and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor-grade PuO2 (~67 wt% fissile) and ThO2, with a central zirconia rod to reduce coolant void reactivity. Several annular and checkerboard-type heterogeneous seed/blanket core concepts with plutonium-thorium–based fuels in a 700-MW(electric)–class PT-HWR were analyzed, using a once-through thorium cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, FU, power distributions, and other performance parameters. WIMS-AECL Version 3.1 was used to perform lattice physics calculations using two-dimensional, 89-group integral neutron transport theory, while RFSP Version 3.5.1 was used to perform the core physics and fuel management calculations using three-dimensional two-group diffusion theory. Among the different core concepts investigated, there were cores where the FU was up to 30% higher than that achieved in a PT-HWR using natural uranium fuel bundles. There were cores where up to 67% of the Pu was consumed, cores where up to 43% of the energy was produced from thorium, and cores where up to 363 kg/year of 233U was produced in the discharged fuel.