Shutdown Reactivity Research Articles

Nuclear thermal coupling optimization of baffled NTP reactor

Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactors have received a lot of research due to simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than $4.59 and a minimum dry shutdown reactivity lower than -$7.94, where there was a high control accuracy of 6.7 pcm/mm. The reactor finally realized safety of the core temperature lower than 3000 K under the conditions of total exit temperature of 2650 K and thrust of 25000 lb.

EVALUATION OF EQUILIBRIUM CORE OPERATION OF THE RSG-GAS REACTOR

The Indonesian Multipurpose Reactor, RSG-GAS reactor will accomplish its first lifetime in December 2020. The reactor has been operated in safe and reliable manner for about 33 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. The paper is intended to evaluate its in-core fuel management that is the conformance between the theory and implementation of the equilibrium core. Evaluation of the reactor operation parameter was carried out for core numbers 91 – 100. The data show that excess reactivity, shutdown reactivity and control rod reactivity have no significant difference at each core. The result shows that the BATAN-FUEL accurately determine the equilibrium core and its fuel loading pattern.This in-core fuel management of the RSG-GAS reactor supports the safety of reactor operation.

Scalable modular dynamic molten salt reactor system model with decay heat

Nodal dynamic models allow rapid reduced-order simulations of complex systems. This paper presents a publicly available system dynamic model of a molten salt reactor system with a topology of the Molten Salt Reactor Experiment (MSRE), in MATLAB/Simulink, building upon a model previously published and validated with MSRE data. It adds scaling of nominal power, dynamic representation of decay heat generation, a generic decay heat removal system, improved documentation, modularity, and test cases. One hour of system transient runs in less than 10 min on a laptop. This model is intended for first order engineering calculations, developing insights into system transient behavior, equipment sizing, parameter sensitivity studies. The model contains an easy to use toolkit which can be customized and expanded to describe any molten salt reactor system. Scenarios involving off-normal transients are presented. Relation between shutdown reactivity insertions, decay heat removal, core recriticality, and safe shutdown conditions are discussed.

Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO2 fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.

Shutdown reactivity of platelets to adenosine diphosphate anr arachidoic acid and its prognostic value in patients with myocardial infraction with elevation of ST segment

The double antiplateled therapy (DAT) with an acetylsalicylic acid (ASA) and clopidogrel is a standard therapy of patients with myocardial infraction but its efficiency in patients is different. Cardiologists of all over the world chose the causes of aforesaid for a long time. Aim of research is to establish predictors of development of a high shutdown reactivity of platelets (HSTR) to ADP and AA (arachidoic acid) in patients with MFST (myocardial infraction with elevation of ST segment) and to evaluate its impact on the clinical course. Materials and methods : It was examined 105 patients with an acute coronary syndrome with elevation of ST segment who underwent thrombolytic therapy. Examination of SRP (shutdown reactivity of platelets) to ADP and AA was carried out on analyzer of aggregation ALAT-2 Results: Values of SRP to AA >1,33 relative units (RU) an odds ratio (OR) 4.6; 95% of confidence interval (CI) 3.3-15.4, р 1,78 RU (OR 5.4; 95%CI 4.7-13.3, р 1,33 and to ADP >1,78 ВО was reliably lower than in patients with level <1,33 RU (р=0,042) і <1,78 RU (р=0,034), respectively. Multifactorial regressive analysis revealed an independent predictors of an increase of SRP to AA and ADP, among them the levels of estradiol, progesterone and its ratio in postmenopause (PM), duration of an arterial hypertension (AH) and postmenopause, pancreatic diabetes (PD)of the 2 type and female sex (all p<0,05). Conclusions. In patients with MI an increase of SRP to AA and ADP is associated with a growth of risk of CEP development in 4,6 і 5,4 times

Self-Sustaining Thorium Boiling Water Reactors

A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants

The negative impact of a boron dilution accident on the safety of a current pressurized water reactor (PWR) initiated investigations with the aim of checking the feasibility of reduced boron concentration operation. In addition, reduction of the maximum boron concentration in a PWR is a practical and feasible means to substantially reduce the radiation dose to operators and to minimize corrosion damage. Four types of integral burnable absorbers have been considered: gadolinium, integral fuel burnable absorber (IFBA), erbia, and alumina boron carbide. Under consideration of four different kinds of fuel assemblies (FA), four core design candidates were developed by applying current PWR OPR-1000 technology and by keeping major engineering design constraints and the equivalent fuel enrichment level used in the reference core (REF) design. However, an optimal design was targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between the REF and the optimal core designs is presented here. One of the designs is suggested as the most promising and favorable low boron core (LBC) design in this framework. The proper combination of axial and radial enrichment zoning patterns plus a mixture of fresh FAs with depleted assemblies in an LBC design candidate with an IFBA-bearing FA at equilibrium cycle could bring a two times narrower axial offset variation than that of the REF design, maintain an acceptable power peaking factor ˜23% lower than the design limit, and achieve higher fuel burnup. It was observed that this optimal LBC design could comply with current OPR-1000 reactor acceptance criteria associated with smooth reactivity swing, more flattened power distribution, and desired limiting safety parameters despite an 18% loss of shutdown reactivity worth at beginning of cycle when compared to the REF design.

Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

In case of a failure of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only four out of five residual arms are capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is incapable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram.

Influence of fuel bundle loading errors on the subcritically during refueling campaigns for the present BWR cores of KRB-II

On the basis of real fuel assembly inventories as they are presently available in KRB-II, the influence of fuel bundle loading errors on the subcriticality during refueling campaigns was investigated with the calculational methods of the incore fuel management. To this, control rod cells which show the least shut-down reactivity were considered and less reactive fuel assemblies were successively exchanged with fuel assemblies of highest possible reactivity from distant core regions. The results show that the total shut-down reactivity is only reduced by a comparatively small amount. The stuck rod shut-down reactivity, on the other hand, is strongly diminished with increasing number of locally concentrated mislocated fuel assemblies of highest possible reactivity. Thus, unintentional criticality cannot be reached during refueling campaigns with all control rods inserted. In conjunction with the deliberate withdrawal of one control rod, two or three mislocated fuel assemblies can cause criticality, depending on the absolute value of the realized stuck rod shut-down reactivity.

The Concept of a Sub-Prompt-Critical Nuclear Reactor and Its Applications to the Solution of Safety-Related Neutronic Problems in Reactor Kinetics

The concept of a sub-prompt-critical nuclear reactor has been used to investigate a variety of neutron kinetics problems in reactor assemblies. Results of reactor response calculations (power, instantaneous period) and reactivity computations obtained using the present model are found to be in agreement with results obtained from the exact kinetics method to within 3 percent. Analytic expressions for the accident energy released in a controlled nuclear power excursion have been derived and their physical significances discussed. It is shown that if the energy released in the shut-down state is to be kept small, the first few dollars of reactivity should be removed as fast as possible and that it is adequate if the shut-down reactivity margin is kept within about 10 dollars. Application of the model to an earlier proposed pulsed fast reactor shows that its fictitious delayed neutron fraction lies between 35 and 45 pcm and its sensitivity between 4 and 6.

A Method of Performing Shutdown Reactivity Measurements in Spent Nuclear Fuel Storage Pools

The objective of this paper is to develop a device to measure the k/infinity/ of a spent fuel assembly used in light water reactors. A subcritical assembly having a cross configuration is designed to allow measurement of the k/sub //infinity/ of a spent fuel assembly by comparing the change in its multiplication with that of a fuel assembly of known k/infinity/. Calculations have been performed using nucleonic codes to develop polynomial equations that relate the k/infinity/ of the spent fuel assembly to measured data. The measurements involve taking count rates with the spent fuel assembly in the center position of the subcritical assembly, and the measured data are the count rate ratio of the spent fuel assembly over the count rate taken with a fuel assembly of known k/infinity/. The polynomial equations are easy to program on a microcomputer, which, together with the subcritical assembly, form the k/infinity/ meter. 9 refs.

Kinetic Parameters of the University of Maryland Reactor by the Interval-Distribution Method

The Rossi alpha of the University of Maryland Reactor was measured at criticality and at shutdown by the Babala interval-distribution method. At criticality, ..cap alpha../sub c/ = 145.8 +- 2.9 s/sup -1/. At shutdown, ..cap alpha../sub s/ = 1025.8 +- 4.1 s/sup -1/. The shutdown reactivity was found to be rho/sub s/..beta.. = -6.04 +- 0.14 dollars. An inverse kinetics analysis of a rod-drop experiment gave rho/sub s//..beta.. = -6.23 +- 0.13 dollars, which is considered with the interval-distribution result.

On integral-versions of the area-ratio method of pulsed neutron reactivity measurements

A new method for the measurement of the shut-down reactivity is suggested. The method involves area-ratio measurements at several points inside the core, and invokes auxiliary static calculations. While the method is completely general, in the present paper only the case of thermal-reactors is discussed in detail.

The Polarity Correlation of Reactor Noise in the Frequency Domain

A newly developed special two-detector cross-correlation experiment of zero-power reactor-noise analysis is applied to determining characteristic kinetic-reactor parameters. This method–called polarity correlation in the frequency domain–is based on the detailed analysis of the stochastic Coherence Function, in contrast to the analysis of auto and/or cross power spectral density functions. The coherence function is easily obtained by polarity correlation of two filtered random-neutron-noise signals simultaneously sampled in a reactor system using neutronsensitive detectors in the current mode of operation.The advantage and the progress in using this technique is both its simplicity and the automatic elimination of the equipment frequency response, which is of special importance when analyzing fast-reactor-noise spectra.A series of prompt-neutron-decay constant measurements was performed on several configurations of the STARK and SNEAK facilities. Moreover, special attention and effort is dedicated to reactivity shutdown measurements by the polarity correlation method. An on-line reactivity meter is described which is particularly suited to making reactivity measurements in large plutonium fueled reactors. Shutdown reactivity measurements were performed down to -8 dollars.

An On-Line Reactivity Meter Based on Reactor Noise Using Two-Detector Crosscorrelation

A prototype on-line reactivity meter using reactor noise analysis techniques is based on a two-detector cross correlation method, which offers as a special feature the absence of uncorrected noise contribution to the spectral density function of the reactor system. Reactivity shutdown measurements were performed on three different zero power reactors down to seven dollars. Special attention is given to the predicted and measured error margins of reactivity. The described on-line meter, when used with optimized current type neutron detectors for minimum gamma contamination, promises to be an encouraging way of making shutdown reactivity measurements, perhaps also in “dirty” power reactors where the high gamma-ray intensities can considerably destroy the “clean neutron signal-to-noise” ratio.

Shutdown reactivity measurements using noise techniques

Shutdown reactivity measurements using noise techniques

Control Rod Worth Studies on Seed and Blanket Reactors

>A technique is described for determining control rod shutdown reactivity for tall reactors (small relative axial leakage) which may also be applicable to other reactor geometries. This method assumes that the rod bank region is an exponential assembly with the critical portion of the reactor as the driving source for the exponential. Comparison is made with the results obtained from the rod drop technique. (auth)