Reactivity Feedback Coefficients Research Articles

Under irradiation issues of the CSZ–based inert matrix fuels from IFA-652 Halden experiment

The renewal of interest in Fast Reactors (FR) deployment as possible answer to the future perspective of natural uranium shortage has led to reconsider the plutonium as an energy source with a consequent radical change in its management. The transition to a closed fuel cycle is estimated to occur in a period ranging from 2030 to 2050. In the near term, Inert Matrix Fuels (IMF) are currently viable options, capable to contribute in reducing the proliferation risk associated to nowadays separated plutonium stockpiles in a business-as-usual nuclear energy scenario where a prominent role is still played by Light Water Reactors (LWR) in a once-through fuel cycle strategy. Double-strata strategy with Accelerator Driven Systems (ADS) machines to transmute minor actinides also support the hypothesis of inert matrix heterogeneous fuels. These reasons justify the current interest in IMF. A high burning efficiency achieved by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Amongst other options, Calcia-Stabilised Zirconia (CSZ) and Yttria-Stabilised Zirconia (YSZ) fulfil these criteria standing as the most promising matrices to host plutonium. ENEA has conceived an in-pile testing (IFA-652 experiment) of viable inert matrices concepts: CSZ and thoria. This experiment was performed in the Halden Heavy Boiling Water Reactor (HBWR), a joint project of the OECD Nuclear Energy Agency. The test-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria matrix, the fissile phase is High Enriched Uranium (HEU) oxide (UO 2 93% 235U enriched). IMT is a ternary fuel composed of CSZ + thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel mainly to improve the low IM reactivity feedback coefficients. The discharge burnup, under typical LWR conditions, was, for all pins, 90–95% of the planned 45 MWd/kgU. Pins are instrumented providing fuel centreline temperature, pin inner pressure and fuel stack elongation measurements. On the basis of the experimental dataset under power ramping, a possible correlation of the FGR onset of CSZ-based fuels with the Halden Vitanza curve apparently exists. The TRANSURANUS (TU) fuel performance code predictions proved to be in nice agreement with the experimental findings concerning the thermal performance of IMF whilst, regarding FGR and densification, a satisfactory modelling is still needed.

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ( ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

Core design studies for a 1000 MW th Advanced Burner Reactor

This paper describes the core design and performance characteristics of 1000 MW th Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of ∼0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from ∼0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

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.

Inert Matrix Fuel Analysis by Means of the TRANSURANUS Code: The Halden IFA-652 In-Pile Test

Inert matrix fuels (IMFs) are a possible option to reduce separated plutonium stockpiles by burning it in light water reactor (LWR) fleet. A high burning efficiency targeted by preventing new plutonium buildup under irradiation (U-free fuel), a proved high radiation damage, and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Among other options, both calcia-stabilized zirconia (CSZ) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. While several in-pile tests concerning thoria fuels are found, calcia-stabilized zirconia under-irradiation performance is still to be fully assessed; with this regard the thermal conductivity, markedly lower than the uranium oxide (UOX) and mixed oxide (MOX) cases, plays a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing of three different U-free inert matrix fuel concepts, which have been performed in the OECD Halden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90–97% of the 45MWd∕kgUeq target under typical LWR irradiation conditions. The test rig is a six-rod bundle loaded with IM, IMT, and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria as matrices, the fissile phase being the high enriched uranium (HEU) oxide (UO2 93% U235 enriched). IMT is a ternary fuel composed by CSZ+thoria matrix and HEU oxide as a fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centerline temperature, pin inner pressure, and fuel stack elongation measurements. Our purpose is to investigate the key processes of IMF under-irradiation behavior by means of the TRANSURANUS fuel performance code. Thermal conductivity and its degradation with burnup, densification-swelling response, and fission gas release (FGR) are tentatively modeled in the burnup range of IFA-652. In particular, the effects of pellet geometry and fuel microstructures in the IM and IMT cases are pointed out. The consistency of our results is discussed aiming at understanding the in-pile response, as a fundamental step, in the perspective of future deployment of the nuclear fuels we are dealing with. Notwithstanding this ambitious objective, it is clear, however, that these results rely on a limited data set and that, as TRANSURANUS is a semi-empirical code mostly tailored for commercial fuels, the modeling of the IMF is still a work in progress.

Communications orales – Recherche fondamentale/recherche clinique

Previous neutronic/thermal-hydraulic (TH) coupled numerical simulations using full-core TRACE/PARCS and SIMULATE-3K BWR models have shown evidence of a specific “rotating mode” behavior (steady rotation of the symmetry line, i.e. constant phase shift of approximately 90° between the first two azimuthal modes) in out-of-phase limit cycle oscillations, regardless of initial conditions and even if the first two azimuthal modes have different natural frequencies. This suggests a nonlinear coupling between these modes; otherwise, the phase shift between these modes would change at a constant rate during the limit cycle. The goal of the present work is to gain further insights on the rotating mode behavior using a simplified mathematical model which contains all of the important physics for this application while providing sufficient flexibility and simplicity to allow for in-depth understanding of the underlying phenomena. This was accomplished using a multi-channel, multi-modal reduced-order model, using a modification of the fixed pressure drop boundary condition to simulate channel coupling via the inlet and outlet plena, in order to destabilize the out-of-phase mode over the in-phase mode. Examination of the time-dependent solution of the nonlinear system showed a clear preference for rotating mode behavior in the four-channel model under stand-alone TH conditions and for conditions with weak neutronic feedback. When neutronic feedback was strengthened (i.e., larger reactivity feedback coefficients), the side-to-side mode (stationary symmetry line) was favored instead. Additional analyses using higher-fidelity numerical modeling, as well as a physical explanation for the rotating behavior seen in both sets of analyses, will be provided in a companion paper (“Part 2”).

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.

Reactivity calculation in the frequency domain for BWR stability codes

This paper deals with the problem of computing the feedback reactivity in the frequency domain codes as the LAPUR code. First, we explain how to calculate the feedback reactivity in the frequency domain using slab-geometry (1D) kinetics, also we show how to perform the coupling of the 1D kinetics with the thermal–hydraulic part of the LAPUR code in order to obtain the density to reactivity feedback coefficients, the power to reactivity feedback coefficients and the inlet temperature to reactivity feedback coefficients for the different channels. Also we explain how to solve the lambda eigenvalue equation in the frequency domain by the spectral nodal method and the comparison between the reactivity obtained by solving the eigenvalue equation in the frequency domain and the reactivity obtained from first order perturbation theory. Because LAPUR works in the linear regime, it is assumed that in general the perturbations are small. There is also a section devoted to the reactivity weighting factors used to couple the reactivity contribution from the different channels to the reactivity of the entire reactor core in point kinetics and 1D kinetics. Also we display a comparison of the reactivity obtained using point kinetics and 1D kinetics for the different kind of perturbations and we check the consistency of the results.

Fertile-Free Annular Fuel for Plutonium Recycling

Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-type reactors can be achieved by the use of internally and externally cooled annular fuel geometry. As a result, the accumulated stockpiles of Pu, especially if incorporated in fertile-free inert matrix, can be burnt at a substantially higher rate as compared with the conventional mixed oxide-fueled reactors operating at standard power density. In this work, we explore the basic feasibility of a PWR core fully loaded with Pu incorporated in fertile-free fuel of annular internally and externally cooled geometry and operating at 150% of nominal power density. We evaluate basic burnable poison designs, fuel management strategies, and reactivity feedback coefficients. The three-dimensional full core neutronic analysis performed with Studsvik Core Management System showed that the design of such a Pu-loaded annular fuel core is feasible but significantly more challenging than the Pu fertile-free core with solid fuel pins operating at nominal power density. The main difficulty arises from the fact that the annular fuel core requires at least 50% higher initial Pu loading in order to maintain the standard fuel cycle length of 18 months. Such a high Pu loading results in hardening of the neutron spectrum and consequent reduction in reactivity worth of all reactivity control mechanisms and, in some cases, positive moderator temperature coefficient (MTC). The use of isotopically enriched Gd and Er burnable poisons was found to be beneficial with respect to maximizing Pu burnup and reducing power peaking factors. Overall, the annular fertile-free Pu-loaded high-power-density core appears to be feasible, although it still has relatively high power peaking and potential for slightly positive MTC at beginning of cycle. However, we estimate that limiting the power density to 140% of the nominal case would assure acceptable core power peaking and negative MTC at all times during the cycle.

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ( ρ ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).

Feedback reactivity coefficients and their coupling

Coupled reactivity feedback coefficients which accounts for variation in fuel temperature and moderator void simultaneously, have been determined for swimming pool type research reactor namely Pakistan Research Reactor PARR-1. The state of art is core criticality calculations, employing lattice cell code WIMS-D/4 and application of Taylor series expansion for core reactivity up to third order, involving two variables, i.e. fuel temperature and coolant void. The spectral effects in one region due to change of parameter in other region have also been studied. When spectral changes in moderator region due to 20 K change in fuel temperature have been incorporated in the calculation of fuel temperature coefficient, the results seems to be improved by 4.12%. Further, the results of void coefficient of reactivity show the improvement of 0.1% when the spectral effect in fuel region due to 5% change in void in moderator region is taken into account. These differences seem to be an improvement in the results, as physically any change in one region is accompanied by change in the other region.

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.

Consistent generation of point feedback reactivity parameters for stability and thermalhydraulic codes

This paper shows a consistent methodology to obtain the point kinetics feedback reactivity parameters to be used by stability codes, like LAPUR-5, or transient codes, like TRAC-BF1. This methodology has been implemented in the code PAPU that generates the point kinetic parameters and feedback reactivity coefficients for the LAPUR and TRAC-BF1 codes. The results of the nodal reactivities obtained with the PAPU methodology have been compared with the results of other codes for different types of perturbations. Also, the reactivity tables generated by PAPU have been used in the LAPUR-5 code obtaining good results when the DR computed by LAPUR with these reactivity tables have been compared with the experimental DR obtained from signal analysis of Cofrentes NPP.

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.

A Combined Nonfertile and UO2 PWR Fuel Assembly for Actinide Waste Minimization

A new COmbined NonFertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept, ~20% of the UO2 fuel pins are replaced with fertile free fuel hosting the transuranic elements (TRUs) generated in the previous cycle. This leads to a fuel cycle sustainable with respect to net TRU generation, and the amount and radiotoxicity of the nuclear waste can be significantly reduced in comparison with the conventional once-through UO2 fuel cycle. It is shown that under the constraints of acceptable power peaking limits, the CONFU assembly exhibits negative reactivity feedback coefficients comparable in values to those of the reference UO2 fuel. Feasibility of the PWR core operation and control with complete TRU recycle has been shown based on full-core three-dimensional neutronic simulation. However, gradual buildup of small amounts of Cm and Cf challenges fuel reprocessing and fabrication due to the high spontaneous fission rates of these nuclides and heat generation by some Pu, Am, and Cm isotopes. Feasibility of the processing steps becomes more attainable if the time between discharge and reprocessing is 20 yr or longer.

Sodium cooled small fast long-life reactor “4S”

CRIEPI and Toshiba Corp. have been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. A conceptual design of 4S ( Super- Safe, Small and Simple) reactor is proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void reactivity are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core lifetime is more than 10 years. The 4S reactor is a metallic fueled sodium cooled fast reactor. A target of an electrical output is 10–50 MW. A remarkable feature of 4S is that its reactivity is not controlled by neutron absorber rods but by neutron reflectors to cope with a long core lifetime and a negative coolant void reactivity. This study includes a design consideration of 4S. Design discussions are mainly focused on various core designs to meet above requirements. A tall core active height is adopted to gain long core lifetime. An averaged fuel burn-up is tried to be increased up to 100 GWd/ton from a point of economic view. The reference 4S designs are 10 MWe (30 years core lifetime) and 50 MWe (10 years core lifetime).

Modified Quasi-Steady-State Method to Evaluate the Mean Power Profiles of Nuclear Excursions in Fissile Solution

A modified quasi-steady-state method has been developed in order to evaluate the mean power during a nuclear excursion in fissile solution. The conventional method used the critical equation based on the one-group theory in order to calculate the reactivity. However, the one-group approximation reduces the calculation accuracy, and the geometrical buckling used in the critical equation is not applicable to complex geometries. Thus, we have modified the method to use the reactivity feedback coefficients, which are widely used in the calculation of one-point reactor kinetics. Although the modified method requires an external calculation to obtain the feedback coefficients, it is applicable to complex geometries and provides more accurate results than does the one-group approximation when the proper coefficients are given. Moreover, a new method to calculate the boiling power has been developed. In this method, the power corresponding to the void fraction that compensated for the inserted reactivity along with the temperature feedback was calculated using the relationship, which was derived using the French SILENE experimental data. Experimental analyses have been conducted to validate the new method for the French CRAC and Japanese TRACY experiments. The analytical results showed close agreement with the experimental results.

Modified Quasi-Steady-State Method to Evaluate the Mean Power Profiles of Nuclear Excursions in Fissile Solution.

A modified quasi-steady-state method has been developed in order to evaluate the mean power during a nuclear excursion in fissile solution. The conventional method used the critical equation based on the one-group theory in order to calculate the reactivity. However, the one-group approximation reduces the calculation accuracy, and the geometrical buckling used in the critical equation is not applicable to complex geometries. Thus, we have modified the method to use the reactivity feedback coefficients, which are widely used in the calculation of one-point reactor kinetics. Although the modified method requires an external calculation to obtain the feedback coefficients, it is applicable to complex geometries and provides more accurate results than does the one-group approximation when the proper coefficients are given. Moreover, a new method to calculate the boiling power has been developed. In this method, the power corresponding to the void fraction that compensated for the inserted reactivity along with the temperature feedback was calculated using the relationship, which was derived using the French SILENE experimental data. Experimental analyses have been conducted to validate the new method for the French CRAC and Japanese TRACY experiments. The analytical results showed close agreement with the experimental results.

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early design phase of any ADS systems in order to assure a benign transient response of the particular ADS design under investigation to typical plant transient initiators.

Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor

Detailed neutronic analysis of a typical swimming pool type research reactor, Pakistan Research Reactor-1 (PARR-1), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is presented in this article. The agreement is generally good.