The current goal of many countries is to achieve carbon neutrality, and nuclear energy will form a major share in this objective. To this end, several new reactor systems are being designed with emphasis on enhanced safety, better fuel utilization, and more economic power generation. Safety in today’s nuclear reactors covers both process, plant, and radiological safety, and these are being improved by several orders of magnitude by design. Fuel utilization is being increased by using advanced fuels with significantly higher burnup and with better fuel and clad integrity to withstand longer residence times. From a neutronic standpoint, the feedback reactivity coefficients are being designed to be negative in all operating phases. These two aims are sometimes opposing, as changes in fuel compositions may lead to lower reactivity feedbacks. Radiological safety becomes extremely important as new energy systems are being considered close to human habitation. This paper will cover new design and safety features of current PHWRs, SMRs, and thorium-based reactors.

The knowledge of time dependent behaviour of neutron population is extremely important for any nuclear reactor operation from safety aspect. The reactor transient can be accurately predicted by solving time dependent neutron diffusion equations. A solution scheme based on implicit method has been implemented in NDIFF3D. The loss of regulation (LOR) transient in initial, intermediate and equilibrium core of AHWR has been simulated. The results have shown good agreement with the data calculated with 3DFAST.

The multigroup cross sections used in the reactor physics calculations have associated uncertainties. The cross section uncertainties, present in the evaluated nuclear data libraries in the form of covariance matrices, are considered as the most significant uncertainty source in the reactor physics analysis. The quantification of these uncertainties is necessary to have confidence in simulation models and ascertain safety margins in operating and future nuclear facilities. A perturbation theory based model has been developed for sensitivity & uncertainty (S&U) analysis capability to the indigenous lattice physics analysis code VISWAM. The sensitivity profiles of k∞ for different reactions such as capture (n,γ), fission (n, f) etc. of significant isotopes such as 235U, 238U etc. are estimated using adjoint flux weighting method based on perturbation theory. The uncertainty quantification in k∞ for these reactions is estimated using the sandwich rule. In the sandwich rule, the relative covariance matrices for a reaction are weighted with the corresponding sensitivity profile. The multigroup relative covariance matrix library required for the uncertainty calculation is generated using NJOY 2016 code system. The accuracy of S&U model developed in VISWAM code has been verified by analyzing Exercise I-1 of the OECD/NEA Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation, and Safety Analysis of LWRs. The Exercise I-1 of the benchmark is dedicated to evaluate uncertainties for a lattice cell due to different microscopic cross sections. The results obtained using VISWAM are compared with other published results for the cell physics benchmark in the present paper.

An experiment was carried out to study the feasibility of 99Mo production using an electron accelerator based epi-cadmium neutron source. The neutrons were produced by the interaction of bremsstrahlung end-point energy of 10 MeV with BeO blocks and the neutrons thus produced were moderated using High Density Polyethylene blocks. Effective cross-section of 98Mo(n,γ)99Mo reaction was measured using foil activation method. An enhancement in the effective cross-section due to resonances in the epi-cadmium region was observed. The photons and neutron production/transport in the experimental set-up were simulated using the FLUKA Monte Carlo code. The simulations were found to be in good agreement with the experimental findings.

Indian Pressurized Water Reactor (IPWR) is being designed with the aim of obtaining better fuel utilisation with enriched uranium as fuel. Stability Analysis is one of the main important aspects of safety studies. A preliminary linear stability analysis has been performed for IPWR equilibrium core using point reactor and lumped fuel and coolant temperature feedback model. The stability region for IPWR with respect to fuel and coolant temperature feedback has also been found.

India has the largest thorium reserves in the world which has been recently estimated to be about 13.3

AHWR - Critical Facility (AHWR - CF) is a "zero power" reactor designed to carry out various reactor physics experiments for validation of AHWR physics design. AHWR-CF is a vertical tank type reactor. This paper describes the experiments carried out for neutron flux measurements on the reactor tank surface of AHWR-CF. A number of bare Au and cadmium covered Ni foils were used to measure thermal and fast neutron flux at the outer surface of the reactor tank. One cadmium covered Au foil was also irradiated at one of the locations on the tank surface, to estimate the epithermal component of neutron flux. Westcott thermal flux on the tank surface (at an elevation of 120 cm from tank bottom) was found to be (5.48 +/- 0.57)E+6 n cm-2 s- 1 on one of the sides of the reactor tank. Cadmium ratio for Au for this location was 40.28, indicating a highly thermalized neutron spectrum. No statistically significant activity was found in the irradiated Ni foils, indicating the absence of fast neutron flux above the measurable threshold. Low value of flux on the reactor tank surface implies that any irradiation damage in the AHWR-CF reactor tank material will be insignificant over its period of operation.

Criticality Accident Alarm System (CAAS) is a mandatory system for prompt alarm and immediate evacuation of personnel during the unusual event of a criticality accident in all nuclear facilities using fissile materials. False criticality alarms are not desirable as they create a panic environment for radiation workers of the plant and the public. We have studied the use of different materials for such alarms and effective monitoring during such events. The present paper describes the design and development of a neutron detection based novel CAAS using the unique method of liquid activation technique of vanadyl sulphate solution, which can prevent false alarms. The neutronic properties of vanadium has been utilised for fast response. The performance evaluation of CAAS such as: testing, efficiency evaluation, threshold setting and various response studies using laboratory neutron sources including 252Cf fission neutron source, standard thermal neutron facility and reactor neutrons has been carried out by following the international recommendations and ensured the compliance. The results are also validated using Monte -Carlo FLUKA simulation. The concentration of vanadyl sulphate solution for liquid activation is optimized as 0.0082 g. cm -3 and the alarm threshold setting is evaluated as: 3881 CPS and 3.61 MBq of induced activity based on the current study. It is also worth mentioning that this liquid based system can be deployed in any reactor and fuel re -processing plants easily.

The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.

Detailed fine mesh reactor physics simulations are warranted for reactor cores in order to estimate the important safety parameters such as detailed pin power distribution in the core and local pin peaking factors. A pin-by-pin reactor analysis code HEXPIN has been developed to perform 3D reactor core analysis with a pincell size mesh. HEXPIN uses the conventional two step calculation methodology for performing reactor calculations. India is developing a 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR) for developing indigenous capability for commercial light water technology. It uses enriched uranium oxide fuel. The batch refueling scheme is adopted for fuel cycle management. A fine mesh pin-by-pin analysis of equilibrium core of IPWR has been performed using HEXPIN code system. The pin homogenised parametric cross section library, required for HEXPIN calculation, has been generated using lattice burn up code EXCEL. The equilibrium core calculation has been performed by a set of repeated calculations till a converged core configuration is achieved. The present paper gives the methodology and brief results of the analysis.

Molten Salt Reactors (MSRs) are characterized by power generation in circulating liquid fuel salt rather than in static fuel pins of traditional nuclear reactors like in water cooled reactors. The fuel flow in MSRs drift the delayed neutron precursors (DNP) and results in lower delayed neutron fraction. Thus, MSRs show added complexity in dynamic analysis due to inherently coupled neutronics and thermal hydraulics. With regard to the reactor safety analysis, accurate predictions of DNPs distribution and its effect on reactor transients turn out to be an interesting subject of research. Due to these unusual features of MSRs, the traditional safety analysis codes are not applicable in dynamic analysis of reactors with circulating fuel. Challenges in accurate modelling and analysis of dynamic responses of MSRs have been described in present work using new capabilities in 3D space-time kinetics code ARCH. The numerical results of the implemented methodology presented in this study are compared with data available from Molten Salt Reactor Experiment (MSRE), a reactor operated at ORNL.

The characteristics of a Pressurised Water Reactor (PWR), considered representative Light Water reactor (LWR), operating on thorium-based fuel has been analysed. An equilibrium cycle of PWR consisting of 151 fuel assemblies and having three-batch fuel management has been considered for the analysis. The two types of thorium-based fuels resulting from mixing LEU and plutonium with ThO2 and representative UO2 have been taken for the core loading and fuel cycle study. The reactivity coefficients, kinetic parameters over the equilibrium cycle have similar trend for three fuel types considered. The control inventory with the natural boron in B4C and in dissolved boric acid is sufficient for the safe operational reactivity control and shutdown in all three alternatives. We find that enriched UO2 based fuel in conventional PWRs can be successfully replaced with ThO2 with Plutonium or LEU as fissile feed. Such approach is useful in initiating thorium as nuclear fuel which will support other diverse approaches requiring design and dimension changes for the utilization of thorium.

Innovative fuel management schemes based on evolutionary algorithms (EAs) now form an integral part of design and operation for nuclear reactors. In the present study, Estimation of Distribution Algorithm (EDA) has been developed for light water reactors (LWRs) and as a case study EDA has been applied to optimize the core loading pattern of VVER X2 benchmark's initial core. EDA belongs to class of EAs where the optimization solution evolves through progress of generations and in each generation sampling of best candidates of previous generation is done. The main objectives of this study is firstly to develop and test EDA to optimize core loading pattern (LP) for VVER type cores and secondly to generate new optimized LPs having same safety features but with improved fuel utilization with respect to reference benchmark case. The most suitable values for internal parameters for EDA like population size (N), percentage best candidates (M) and weighting factor '& alpha;' were evaluated for reaching a reasonable optimized solution in computationally efficient way. During this study, a number of optimized initial core loading patterns were generated, where the target parameters and safety limits are met. A comprehensive analysis for selected LPs has been carried out for full cycle and a comparison has been done for the reported safety and operational characteristics. It is observed that a few of the optimized LPs have better fuel economy than the benchmark LP.

Today's energy systems have diverse objectives catering to enhanced safety, ease of operation and better fuel utilization both for power generation and several other applications of nuclear reactions. Bhabha Atomic Research Centre (BARC) has been engaged in the design of new energy systems such as Advanced Heavy Water Reactor (AHWR), Indian Pressurised Water Reactor (IPWR), Molten Salt Reactor (MSR) and High Temperature Reactors (HTRs). The design optimisation of these reactors has varied challenges as they involve new fuel, newer materials, complex geometries, and inherent safety. AHWR, MSR and HTRs based on thorium fuel cycle while IPWR is optimized for longer cycle length with enriched uranium. HTRs are being developed as nuclear power packs for remote areas and they have interesting physics features of long-life cores and hence more challenging to be modelled.

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

The 100kWth Compact High-Temperature Reactor (CHTR) is envisaged as a technology demonstrator for the Indian high-temperature reactor programme enabling process heat applications of nuclear energy such as thermochemical splitting of water for hydrogen. Coupled neutronics–thermal-hydraulics analyses of safety transients in prototype reactors like CHTR are essential for safe design and operation under the high-temperature regime. In this regard, the unprotected loss of flow accident (ULOFA) of CHTR core in full-power configuration has been investigated with the indigenously developed and validated code system. The ULOFA of CHTR has also been analysed for severe transient case considering the simultaneous accidental withdrawal of one of the rods for power regulation. The simulations of these transients have been carried out with indigenously developed point kinetics code PATH. The code is integrated with the thermal-hydraulics module to predict the temperature distribution in the fuel assembly for feedback. The variations of nuclear power, core reactivity and radial and axial temperature profile in an average-powered and peak-powered channel of CHTR are predicted for the 90 min of transients after the trigger. The simulation methodology and results of the safety transients of CHTR are reported in this article.

Development of new reactor concepts to suit future requirements and safety standards, needs very accurate and efficient reactor simulations and analysis. This necessitates development of high fidelity reactor physics computer codes. The present paper outlines various physics design tools being developed for the core design of advanced reactor concepts such as High Temperature Reactor (HTR) and Molten Salt Reactor (MSR) etc. Modelling advancements to analyze complex features unique to new reactor designs are also discussed in brief.

Benchmarking and validation of the neutronic codes is an important aspect of the reactor design. This paper presents VVER-1000 X2 Benchmark steady state core physics analysis carried out using the lattice level burn-up code, EXCEL and core physics analysis using multi-group diffusion code, TRIHEXFA developed at BARC. These codes have been augmented with new features for modeling of LWR Cores. The VVER-1000 X2 Benchmark with modern fuel assembly (FA) design is a comprehensive benchmark proposed for validating and verifying package of codes and data libraries for reactor physics calculations of LWRs, which includes fuel assembly modeling, generation of homogenized cross section data base for all envisaged core conditions, and reactor core modeling. The trends of k infinity (k-inf) values as a function of various parametric variations and as a function of burn-up for all the fuel assemblies considered in benchmark are analyzed in detail and the deviations have been estimated.The VVER-1000 core follow-up simulations are done by using core diffusion code TRIHEXFA using the few group data generated by EXCEL code. Important core physical characteristics like critical soluble boron, 2-D radial power distribution, axial power distribution, radial peaking factors and volumetric power peaking factors, FA burn-up values obtained from the TRIHEXFA simulations are compared with the experimental core follow up data. All these parameters are found to be reasonably in good agreement with the values provided in benchmark.

From the Indian perspective, molten salt reactor technology is a promising option for thorium utilization in the third stage of the nation's nuclear programme. Recently, the physics design of a fluoride salt fuelled small molten salt reactor has been initiated as an experimental core to demonstrate the circulating fuel reactor technology. However, the literature suggests that chloride fuel salts have better breeding characteristics, particularly in fast-spectrum MSRs. The thermal-hydraulic and neutronics properties of the chloride fuel salts could considerably differ from that of fluoride. Therefore, the applicability of chloride fuel salts in a small MSR core has been investigated for the neutronics feasibility of the fuel type. In the study, criticality and safety parameters such as neutron multiplication factors and fuel temperature coefficient of reactivity are assessed for three options of chloride-based molten fuel salts. The predicted results for the experimental MSR thermal core fuelled with chloride-based fuel salts are reported in the paper. A brief review on the comparison of chloride fuel salts with the fluoride salts in MSR designs is also presented.