Nuclear reactor safety system

Summary

The three primary objectives of nuclear reactor safety systems as defined by the U.S. Nuclear Regulatory Commission are to shut down the reactor, maintain it in a shutdown condition and prevent the release of radioactive material. A reactor protection system is designed to immediately terminate the nuclear reaction. By breaking the nuclear chain reaction, the source of heat is eliminated. Other systems can then be used to remove decay heat from the core. All nuclear plants have some form of reactor protection system. Control rods are a series of rods that can be quickly inserted into the reactor core to absorb neutrons and rapidly terminate the nuclear reaction. They are typically composed of actinides, lanthanides, transition metals, and boron, in various alloys with structural backing such as steel. In addition to being neutron absorbent, the alloys used also are required to have at least a low coefficient of thermal expansion so that they do not jam under high temperatures, and they have to be self-lubricating metal on metal, because at the temperatures experienced by nuclear reactor cores oil lubrication would foul too quickly. Boiling water reactors are able to SCRAM the reactor completely with the help of their control rods. In the case of a loss of coolant accident (LOCA), the water-loss of the primary cooling system can be compensated with normal water pumped into the cooling circuit. On the other hand, the standby liquid control (SLC) system (SLCS) consists of a solution containing boric acid, which acts as a neutron poison and rapidly floods the core in case of problems with the stopping of the chain reaction. Pressurized water reactors also can SCRAM the reactor completely with the help of their control rods. PWRs also use boric acid to make fine adjustments to reactor power level, or reactivity, using their Chemical and Volume Control System (CVCS). In the case of LOCA, PWRs have three sources of backup cooling water, high pressure injection (HPI), low pressure injection (LPI), and core flood tanks (CFTs).

Official source

https://en.wikipedia.org/wiki/Nuclear_reactor_safety_system

About this result

This page is automatically generated and may contain information that is not correct, complete, up-to-date, or relevant to your search query. The same applies to every other page on this website. Please make sure to verify the information with EPFL's official sources.

Related publications (3)

Related concepts (19)

Related courses (5)

Related lectures (50)

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021

Sodium-cooled fast reactor (SFR) technologies have the potential to guarantee energy supply and to reduce the burden of nuclear waste for future generations. For an adequate simulation of these reactor systems, well-established tools that have so far been applied mainly to light water reactor (LWR) concepts need to be validated and enhanced. For licensing purposes, there is an increasing interest in replacing conservative calculations by best-estimate calculations supplemented by uncertainty analyses. Nuclear data are a major source of uncertainties in reactor physics calculations. The propagation of nuclear data uncertainties to important system responses is important for determining appropriate safety margins in safety analyses. A systematic approach for quantifying nuclear data--induced uncertainties for all stages of modeling is needed to assess the performance of traditional methods for uncertainty and sensitivity analysis and to unveil the major drivers of observed uncertainties in SFRs. This thesis presents a basis for such a systematic approach through the use of sub-exercises that address different levels of modeling as addition to the OECD/NEA Benchmark for Uncertainty Analysis in Modelling of SFRs. The major analysis method applied within this thesis was the random sampling--based XSUSA method in which nuclear data is varied based on the corresponding covariance data. As a basis for analyses using several multigroup neutron transport codes from the SCALE code system, new multigroup cross section and covariance libraries were developed and optimized for the analysis of SFR systems. In order to use the time-efficient XSUSA method in combination with the SCALE 6.2 release, SCALE's random sampling sequence Sampler was extended to allow the perturbation of cross sections after the self-shielding calculation, including an optional approximation for consideration of implicit effects. XSUSA allowed for the determination of one correlation-based sensitivity index to identify the main contributors to observed uncertainties. This sensitivity analysis was extended by a second correlation-based sensitivity index, as well as variance-based Sobol' sensitivity indices. Furthermore, corresponding indices that use sensitivity coefficients from perturbation theory were developed to allow for comparisons between the various approaches. Finally, systematic uncertainty and sensitivity analyses with respect to nuclear data were performed based on the developed specifications and the described developments. It was found that the analysis of simple models is sufficient for initial assessments of the impact of nuclear data uncertainties on larger scale models as well as the corresponding identification of the uncertainties' major drivers. In general, significantly larger uncertainties for eigenvalues and reactivity coefficients were observed than in corresponding LWR calculations. The main contributor to the uncertainty for most output quantities was identified as inelastic scattering of U-238. Other relevant contributors are the scattering reactions of the coolant and the structural material. By comparing results based on various methods and models, the studies presented in this thesis contribute to the development and assessment of calculation methods and models for uncertainty analysis accompanying best-estimate reactor simulations of SFR.

EPFL2020

Recent LOCA tests with high burnup fuel at the OECD Halden Reactor Project and at Studsvik demonstrated the susceptibility of the fuel to fragment to small pieces, to relocate and possibly cause a hot-spot effect and to be dispersed in the event of cladding rupture. However, the LOCA safety criteria defined by the US NRC are still based on fuel tests with fresh and low burnup fuel and therefore require revision for high burnup fuel. In this context a PhD project with the goal of developing new models for high burnup fuel fragmentation, relocation and dispersal during Loss of Coolant Accident in Light Water Reactors was launched at Paul Scherrer Institute in June 2013 with a financial support from swissnuclear. The work continued with base irradiation simulation with a closer look at the fission gas release measurements of high burnup BWR fuel rods.The data showed significant scatter for fuel rods at the same average burnup, originating from the same fuel assembly and having almost the same enrichment. The scatter was explained with a BWR-specific mechanism for fission gas trapping. It was motivated with Scanning Electron Microscopy images at the pellet-cladding interface that showed very strong fuel-cladding bonding layer on one side of the fuel rod. Base irradiation with the EPRI's FALCON code coupled with an in-house advanced fission gas release and gaseous swelling model GRSW-A was done for selected high burnup BWR fuel rods. The calculated fission gas release was overestimated in all cases. This was expected, because at the time the modelling did not account for fission gas trapping. From the available modelling and experimental data, a model for fission gas trapping was proposed. After calibration, the agreement between calculated and measured fission gas release was significantly improved. Fuel fragmentation modelling was focused on fuel pulverization. This is a high burnup fuel-specific phenomenon, in which the fuel pellet periphery may fragment to sizes less than 100 ÎŒm. Such fragments are very mobile and can be easily relocated and dispersed in the event of cladding rupture regardless of the rupture opening size. Therefore, the fuel fragmentation model addresses the potentially most important mode of fragmentation - fuel pulverization. Fuel relocation inside fuel rod is simulated during cladding ballooning and until cladding failure. The model takes as input the time-dependent cladding deformation supplied by FALCON and the fragment size distribution either provided directly by experimental data or by the fuel fragmentation model. Fuel relocation was modelled under the specific assumption that outermost fragments (e.g. pulverized fuel) relocate first, resulting in a large packing factor and local cladding temperature increase at the balloon (i.e. hot-spot effect) and, as a consequence, in enhanced cladding oxidation at the balloon. Fuel dispersal is modelled by solving the mass, energy and momentum conservation equations for two phases - the gas inside the fuel rod and the fraction of fuel which is considered "moveable". Although the model uses simplified geometrical representation of the fuel rod and some other simplifying assumptions, the underlying reason for fuel dispersal (namely the interfacial friction between the gas outflow and the solid) is explicitly simulated. The model for fuel dispersal is calibrated using Halden and Studsvik LOCA tests.

EPFL2017

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

EPFL2021

EPFL2020

EPFL2017