Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

Approach to equilibrium fuelling scheme of 500 MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on fuel and blanket pins it is found that the nominal core achieves full power only in mid-cycle. A novel interpolation scheme is used to find the peak LHR in any axial section of a fuel/blanket sub-assembly. Breeding ratio is adequate for self-sufficient Pu generation in a closed fuel cycle with Pu from axial blankets and two rings of radial blanket sub-assemblies.     

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500 MWe PFBR

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermal-hydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.     

Thermohydraulic transient studies of the Chinese 200 MWe HTR-PM for loss of forced cooling accidents

Pressurized and Depressurized Loss Of Forced Cooling (PLOFC and DLOFC) are two important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behaviors during LOFC can provide useful reference to the physics, thermohydraulic and structure designs of the reactor core, and can also verify the design of the Residual Heat Removal System (RHRS). The 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China, is characterized by its inherent safety features, such as shutdown ability via negative temperature coefficients of reactivity, passive mechanism of decay heat removal and so on.     

Integration of functional reliability analysis with hardware reliability: An application to safety grade decay heat removal system of Indian 500 MWe PFBR

A passive system can fail either due to classical mechanical failure of components, referred to as hardware failure, or due to the failure of physical phenomena to fulfill the intended function, referred to as functional failure. In this paper a methodology is discussed for the integration of these two kinds of unreliability and applied to evaluate the integrated failure probability of the passive decay heat removal system of Indian 500 MWe prototype fast breeder reactor (PFBR). The probability of occurrence of various system hardware configurations is evaluated using the fault tree method and functional failure probabilities on the corresponding configurations are determined based on the overall approach reported in the reliability methods for passive system (RMPS) project. The variation of functional reliability with time, which is coupled to the probability of occurrence of various hardware system configurations is studied and incorporated in the integrated reliability analysis. It is observed that this consideration of the dependence of functional reliability on time will give significant advantages on system reliability. The integrated reliability analysis is also explained using an event tree. The impact of the provision for forced circulation in the primary circuit on functional reliability is also studied with this procedure and it is found that the forced circulation capability helps to bring down the total decay heat removal failure probability by lowering the peak temperatures after the reactor shut down.     

Design concepts for reactor assembly components of 500 MWe future SFRs

500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction in India. Beyond PFBR, it is planned to construct 3 twin units; each one is 2 MWe × 500 MWe capacity reactors with improved economy and enhanced safety. Significant capital cost reduction is targeted for the reactor assembly, by way of introducing new concepts for the grid plate, primary pipes, top shield and fuel handling system and optimizing the main vessel diameter and bottom dished head shape. The capital cost reduction of the reactor assembly components that could be achieved through these improved concepts is estimated to be about 25%. To validate these concepts, preliminary analysis has been completed, R&D areas have been identified and strategy to execute the R&D has been defined clearly. The basis of each concept is highlighted to depict the Indian approach and strategy to make the fast reactor economically competitive.     

Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.     

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

Measurements of the flow characteristics of the lateral flow in the 6 × 6 rod bundles with Tandem Arrangement Vanes

It is very important to increase the heat transfer efficiency in rod bundles in order to prevent the hot spot on the surface of fuel rods in view of the thermal hydraulic safety of nuclear power plants. It is representative to mount vanes in the support grid, which generate swirling flow. It is necessary to measure the flow pattern for investigating the thermal hydraulic flow characteristics in subchannels. In this study, it is performed to measure experimentally the flow field in cross-sections of the 6 × 6 rod bundles with new type vanes - Tandem Arrangement Vanes (TAV) by using Laser Doppler Anemometry. Through measurements, data are acquired at a nominal Reynolds number of 50,000 and for three streamwise locations at 3, 10, and 20 hydraulic diameters. Many previous experimental studies by the existing split mixing vanes show small turbulent length scales and short retention time till 10D_h after spacer grid. On the other hand, the TAVs proposed in the present study generate the big enforced swirl flow more than 20D_h after spacer grid and heat transfer effect are maintained through this distance.     

Corrosion experiments of steels in flowing Pb at 500 °C and in flowing LBE at 450 °C

In accelerator driven systems (ADS), as well as in the next Generation IV reactors, one of the concerned issues is the material compatibility and corrosion in liquid Pb, which is considered a candidate coolant. Liquid metal corrosion of the structural materials can proceed via different processes: species dissolution and penetration of liquid metal along grain boundaries and metal. The occurrence of these corrosion phenomenon depend on the experimental parameters, such as temperature, thermal gradients, solid and liquid metal compositions, velocity of the liquid metal and oxygen activity in Pb. One possible technique to prevent any corrosive attack by the liquid metals is the in situ passivation of the containment steels. This technique is achieved through an active control and monitoring of the dissolved oxygen concentration. This paper summarizes the data gathered from the CHEOPE III loop, where passivation of T91 and AISI 316L steels is tested in pure Pb at 500 °C were carried out, comparing them with preliminary corrosion data, in LBE, gathered from the LECOR loop.     

A Monte Carlo computer code for evaluating energy loss of 10 keV to 10 MeV ions in amorphous silicon materials

The basic concepts of a computer simulation code for determining the energy loss of ions in the 10 keV to 10 MeV energy range in amorphous silicon materials were presented and discussed. Data obtained were found in good agreement with those obtained by using a SRIM programme. Electronic and nuclear energy losses were evaluated. Variation of the energy loss as a function of the incident ion energy were studied. This new computer code is a good tool for evaluating stopping powers of various materials for light and heavy ions.     

Creep rupture of a 9Cr1MoNbV steel at 500 °C: Base metal and welded joint

ASME Grade 91 steel base metal and a similar weld were tested under creep at 500 °C for rupture time up to 18,000 h. Creep failure of cross-weld specimens occurs in the weld metal at this temperature. No significant microstructural changes were observed after creep. Analysis of creep deformation of smooth creep bars, welded joints and slightly notched bars indicated an apparent creep stress exponent of 19. For the creep conditions considered, failure of the material can be explained by the viscoplastic instability of the specimens without significant damage development. This allowed to develop a simple analysis for time to failure prediction.     

Contribution of near-edge processes to attenuation of the characteristic X-rays in elements with 48 ≤ Z ≤ 83

Attenuation of the characteristic K X-rays in the ₄₈Cd, ₅₀Sn, ₅₂Te, ₆₄Gd, ₆₅Tb, ₆₆Dy, ₆₈Er, ₇₄Ta, ₇₅Re, ₇₉Au, ₈₂Pb and ₈₃Bi elements have been measured with especial emphasis for the X-ray energies (E_in) in the region of respective K-shell/L_i subshell (i = 1, 2, 3) ionization threshold (B_K/B_Li). The characteristic X-rays were obtained from different fluorescent target elements excited by the X-rays and γ-rays emitted from the ⁵⁵Fe and ²⁴¹Am radioisotopes, respectively. The measurements were performed using an energy-dispersive detection set up involving a low-energy Ge detector. The measured attenuation coefficients for the X-rays with energies away from ionization thresholds of the attenuator element are found to be in good agreement with the available theoretical coefficients, which incorporate contributions of the photoionization, and the Rayleigh and Compton scattering processes. However, the measured attenuation coefficients are found to deviate significantly from the theoretical values for the X-rays with energies in vicinity of B_K/Li. The observed alteration is attributed to the X-ray Absorption Fine Structure (XAFS) for negative B_K/Li − E_in values, and the K-shell/L_i subshell resonant Raman scattering (RRS) process for positive B_K/Li − E_in values. Systematic of the K-shell/L_i subshell RRS contribution to attenuation of the X-rays are discussed in terms of the respective oscillator density and fraction of electrons available in the K-shell/L_i subshell Lorentzian profile of the attenuation element below E_in.     

Non-intrusive experimental investigation of flow behavior inside a 5 × 5 rod bundle with spacer grids using PIV and MIR

The validity of the simulation results from computational fluid dynamics (CFD) is still under scrutiny. Some existing CFD closure models for complex flow produce results that are generally recognized as being inaccurate. Development of improved models for complex flow simulation requires an improved understanding of the detailed flow structure evolution with dynamic interaction of the flow multi-scales. Thus, the goal of this work is to contribute to a better understanding of presupposed and existent events that could affect the safety of nuclear power plants. The fundamental phenomena of fluid flow in rod bundles with spacer grids can be elucidated by using state-of-the-art measurement techniques. This study aims to develop an experimental data base with high spatial and temporal resolution of fluid flow velocity inside a 5 × 5 rod bundles with spacer grids. The full-field detailed data base is intended to validate CFD codes at various temporal-spatial scales. Measurements are carried out using dynamic particle image velocimetry (DPIV) technique inside an optically transparent rod bundle utilizing the matching index of refraction (MIR) approach. This work presents full field velocity vectors and turbulence statistics for a rod bundle under single phase flow conditions.     

Angular spectra of rainbow scattering at glancing keV He bombardment of NiAl(1 0 0) surface with transverse energies in the range 1-10 eV

Previous simulations of glancing incidence ion-surface interaction have demonstrated that classical dynamics using the row-model have successfully reproduced multimodal azimuthal and polar spectra. These studies have also shown considerable sensitivity to the form of the interatomic potential thus making it a strong test of the validity of such potentials and even allow deduction of the ion-surface potentials. In these simulations the individual pairwise interactions between the projectile and the target atoms have been replaced by cylindrical potentials.Comparison to numerous experimental studies have confirmed the existence of rainbow scattering phenomena and successfully tested the validity of the cylindrical potential used in these simulations. The use of cylindrical potentials avoids stochastic effects due to thermal displacements and allows faster computer simulations leading to reliable angular distributions.In the present work we extend the row-model to consider scattering from binary alloys. Using He⁺ scattered at glancing incidence from NiAl surfaces, Al or Ni terminated, a faster method has been developed to easily and accurately quantize not only the maximum deflection azimuthal angle but all the singular points in the angular distribution. It has been shown that the influence of the surface termination on the rainbow angle and the inelastic losses is small.     

Effects of Fe and Cr on corrosion behavior of ZrFeCr alloys in 500 °C steam

A study of the corrosion behaviors of ZrFeCr alloy and the influence of microstructure on corrosion resistance are described by X-ray diffraction and scanning electron microscope in this paper. The results show that several ZrFeCr alloys exhibit protective behavior throughout the test and oxide growth is stable and protective. The best alloy has the composition Zr1.0Fe0.6Cr. Fitting of the weight gain curves for the protective oxide alloys in the region of protective behavior, it showed nearly cubic behavior for the most protective alloys. The Zr1.0Fe0.6Cr has the more laves Zr(Fe,Cr)₂ precipitate in matrix and it has the better corrosion resistance. The Zr0.2Fe0.1Cr has little precipitate, the biggest hydrogen absorption and the worst corrosion resistance. The number of precipitates and the amount of hydrogen absorption in Zr alloy plays an important role on corrosion resistance behaviors in 500 °C/10.3 MPa steam.     

Comparison of some lead and non-lead based glass systems, standard shielding concretes and commercial window glasses in terms of shielding parameters in the energy region of 1 keV-100 GeV: A comparative study

The effective atomic numbers, Z_eff of some glass systems with and without Pb have been calculated in the energy region of 1 keV-100 GeV including the K absorption edges of high Z elements present in the glass. Also, these glass systems have been compared with some standard shielding concretes and commercial window glasses in terms of mean free paths and total mass attenuation coefficients in the continuous energy range. Comparisons with experiments were also provided wherever possible for glasses. It has been observed that the glass systems without Pb have higher values of Z_eff than that of Pb based glasses at some high energy regions even if they have lower mean atomic numbers than Pb based glasses. When compared with some standard shielding concretes and commercial window glasses, generally it has been shown that the given glass systems have superior properties than concretes and window glasses with respect to the radiation-shielding properties, thus confirming the availability of using these glasses as substitutes for some shielding concretes and commercial window glasses to improve radiation-shielding properties in the continuous energy region.     

Double excitation of helium in collisions with proton and antiproton impact in the energy range 50-500 keV

Double-electron excitation processes of helium atoms by proton and antiproton impact have been theoretically investigated using the four-body formalism of boundary corrected continuum intermediate state (BCCIS-4B) approximation in the energy range of 50-500 keV. In this formalism, the presence of the projectile in the exit channels is described by distorting the final bound state wave functions with coulomb waves (associated with the projectile-electron interactions). The results are in good agreement with the other theoretical and experimental results. Reasonably better agreements have been found in the intermediate and high energy regions. Contributions to the cross section of the different magnetic sub-shells are also analysed.     

Corrosion rate of parent and weld materials of F82H and JPCA steels under LBE flow with active oxygen control at 450 and 500 °C

Corrosion behavior of parent and weld materials of F82H and JPCA was studied in the circulating LBE loop under impinging flow. These are candidate materials for Japanese Accelerator Driven System (ADS) beam windows. Maximum temperatures were kept to 450 and 500 °C with 100 °C constant temperature difference. Main flow velocity was 0.4-0.6 m/s in every case. Oxygen concentration was controlled to 2-4 × 10⁻⁵ mass% although there was one exception. Testing time durations were 500-3000 h. Round bar type specimens were put in the circular tube of the loop. An electron beam weld in the middle of specimens was also studied. Optical microscopy, electron microscopy, X-ray element analyses and X-ray diffraction were used to investigate corrosion in these materials. Consequently corrosion depth and stability of those oxide layers were characterized based on the analyses. For a long-term behavior a linear law is recommended to predict corrosion in the ADS target design.