Build Up and Characterization of Ultraslow Nuclear Burn-Up Wave in Epithermal Neutron Multiplying Medium

A study is carried out on the build-up and characterization of ultraslow nuclear burn-up wave in epithermal neutron multiplying medium for slab geometry. Uranium-plutonium fissile medium is considered for the calculation. Transient part of the wave is characterized by transient time (TT), transient length (TL), and TT and TL are defined as the time and distance required to develop asymptotic neutron flux propagating through the media. Steady-state part of the wave is characterized with wave velocity and reaction zone width (full width half maximum (FWHM) and full width 10% of maximum (FW10M)). Parametric studies are carried out for different enrichment of 235U and different values of external source of neutron. It is observed that TT, TL, FWHM, and FW10M decrease with the increase in enrichment. The velocity of the wave increases with the enrichment of 235U. This study is beneficial for understanding the characteristics of nuclear burn-up wave in epithermal region as it will help in further researches in this area.

Nonhyperbolicity of Conservation Equations of RELAP5 Two-Fluid Model in Nuclear Reactor Safety Results (Investigation and Eigenvalue Analysis)

The RELAP5 code simulates the thermal-hydraulic characteristics of nuclear reactors by the use of a two-fluid one-dimensional, nonequilibrium, nonhomogeneous two-phase flow model. This model consists of six governing equations to describe the mass, energy, and momentum of the two fluids. The scope of this work comprises the study of the mathematical nature of the code model and to predict the accuracy of the model in the nuclear reactor safety analysis. The method of characteristics (MOC) is applied to check the nonhyperbolic nature of conservation equations for all normal and accident conditions of light water reactors (LWRs). The analysis also gives information about the soundness of the model and to identify the regions where the solutions obtained from it will be numerically convergent. The characteristics of equations of nonhyperbolic nature are complex. It implies that results thus obtained (by finite difference method) have to be interpreted very carefully in view of the sensitive nature of reactor safety analysis. The present analysis shows that governing equations of the code exhibit complex characteristics for some operating conditions thus implying nonhyperbolicity under those conditions. Results are less accurate under such conditions, so sensitivity analysis plays an important role. The sensitivity of closure relationship on the conservation equation's stability is also checked. The analysis is performed in matlab environment for three different systems, (a) pressurized water reactor (PWR), boiling water reactor (BWR), and (c) natural circulation reactor or advance heavy water reactor (AHWR). These results can also be extended to other thermal-hydraulic systems. The different values of the coefficient of closure relationship are taken for different flow regimes. It is observed that the coefficient of virtual mass (for momentum equation) has a significant effect on the hyperbolicity of the system. It is recommended that further development of the RELAP5 model be performed to identify changes that would reduce the region of complex characteristics. The importance of MOC (in nuclear reactor thermal-hydraulic safety analysis) is evident here. In addition, a detailed analysis for operating pressures range of 0.1–22.5 MPa is also performed to find out the nonhyperbolic regions of code model and realistic data of the different type of reactors is used as input of the code. It is also observed here that RELAP5 results are less accurate when system pressure exceeds 19.5 MPa.

Analysis of Loss of Heat Sink for ITER Divertor Cooling System Using Modified RELAP/SCDAPSIM/MOD 4.0

The present work includes thermal hydraulic modeling and analysis of loss of heat sink (LOHS) accident for the ITER divertor cooling system. The analysis is done for the new design of full tungsten divertor. The new design is also analyzed for different local heat loads ranging from 10 MW/m2 to 20 MW/m2 (while maintaining the total heat load 200 MW) under the steady-state fluid conditions. The LOHS event is selected since divertor is the most sensitive component to loss or reduction in coolability of divertor primary heat transport system (DV-PHTS) loop as it receives large heat flux from plasma. The main objective of this analysis is to find margins to unwanted conditions like overstress temperatures of structure and elevated water level in the pressurizer. The analysis is done by modified thermal hydraulic code RELAP/SCDAPSIM/MOD 4.0. The results obtained are compared with the results of old divertor design which uses carbon fiber composite (CFC) layer to show that how the new design of divertor behaves compared to the older design under the accident scenario. A detailed model of DV-PHTS loop and its ancillary system is presented. The model includes promotional integral differential (PID) controller for controlling the pressurizer heater and spray system. A detailed pump model is also included in the present analysis which was previously used as a time-dependent junction. The analysis shows that under the accident scenario, (a) the divertor structure temperature at the critical sites (inner vertical target (IVT) and outer vertical target (OVT)) is always within the design limit and does not affect the structural integrity of the divertor. (b) The water level in the pressurizer increases moderately and finely controlled by the PID controller, and pressurizer safety valve does not open.

Thermal Hydraulic Safety Assessment of LLCB Test Blanket System in ITER Using Modified relap/scdapsim/mod4.0 Code

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code relap/scdapsim/mod4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Thermal Hydraulic and Safety Assessment of First Wall Helium Cooling System of a Generalized Test Blanket System in ITER Using RELAP/SCDAPSIM/MOD4.0 Code

The key objective of the test blanket module (TBM) program is to develop the design technology for DEMO and future power-producing fusion reactors. The proposed first wall of the test blanket system (TBS) is a generalized concept for testing in ITER, an experimental fusion reactor under construction in France presently. The first wall of TBM (TBM FW) directly faces the plasma and is cooled by the first wall helium cooling system (FWHCS), which is considered as a critical component from an ITER safety point of view. The scope of this work comprises thermal hydraulic analysis of the FWHCS of a generalized TBS and the assessment of postulated initiating events (PIEs) on the ITER safety with the help of thermal-hydraulic code RELAP/SCDAPSIM/MOD4.0. The three reference accidents: in-vacuum vessel (VV) loss of coolant accident (in-vessel LOCA), ex-vessel LOCA, and loss of flow accident (LOFA) in FWHCS are selected for the safety assessment. This safety assessment addresses safety concerns resulting from FWHCS component failure, such as VV pressurization, TBM FW temperature profile, pressurization of port cell (PC) and Tokomak cooling water system vault annex (TCWS-VA), and passive decay heat removal capability. The analysis shows that the critical parameters are under the design limit and have large safety margins, in the investigated accident scenarios. A comparative analysis is also carried out with the previous results to validate the results.