This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be $\sim 13$ kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.

A 100 MWt reactor design has been conceived to support flux level of the order of 10¹⁵ n/cm²/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D₂O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium–aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of $8 \times 10^{14}$ n/cm² /s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.

In recent years, there has been an increasing worldwide interest in accelerator driven systems (ADS) due to their perceived superior safety characteristics and their potential for burning actinides and long-lived fission products. Indian interest in ADS has an additional dimension, which is related to our planned large-scale thorium utilization for future nuclear energy generation.

The physics of ADS is quite different from that of critical reactors. As such, physics studies on ADS reactors are necessary for gaining an understanding of these systems. Development of theoretical tools and experimental facilities for studying the physics of ADS reactors constitute important aspect of the ADS development program at BARC. This includes computer codes for burnup studies based on transport theory and Monte Carlo methods, codes for studying the kinetics of ADS and sub-critical facilities driven by 14 MeV neutron generators for ADS experiments and development of sub-criticality measurement methods. The paper discusses the physics issues specific to ADS reactors and presents the status of the reactor physics program and some of the ADS concepts under study.

A number of new wordings and statements regarding the targeted problem of destruction of long-lived wastes (transmutation) is considered. Some new criteria concerning the efficiency of a particular burner type are proposed. It is shown that the destruction efficiency of a specific burner is greatly influenced by the prospective time period of the whole destruction process.

Transmutation of long-lived actinides and fission products becomes an important issue of the overall nuclear fuel cycle assessment, both for existing and future reactor systems. Reliable nuclear data are required for analysis of associated neutronics. The present paper gives a review of the status of nuclear data analysis focusing on the waste transmutation problem.

Characteristics of the equilibrium fuel cycle for the core or a blanket of ADS having the structure of the core of a fast lead-cooled reactor of type BREST (Russian abbreviation for `Bystryy Reaktor so Svintsovym Teplonositelem') in a mode of americium transmutation are calculated. Americium loading was taken 5% of heavy atoms. Keeping the average multiplication factor the same as in a standard equilibrium cycle, reactivity swing over 1 year's microcycle is about 1%, that demands partial fuel reloading with a periodicity of about one month. For one year of operation, 61 kg of americium is destroyed, and due to increased ²³⁸Pu content, americium is mainly converted to fission products. Thus in a system of 1 GWt (thermal), 87 kg of americium can be transmuted yearly. The estimate of the reactivity void effect has shown that it increases to 0.6% almost linearly with the void fraction increasing up to 25% and reaches its maximum of 0.7% at a void fraction of about 50%. Application of similar strategy for ADS with a sub-criticality level $\approx 0.96–0.98$ can essentially relax safety problems related to positive void effects.

Target-blanket facility `Energy + Transmutation' was irradiated by proton beam extracted from the Nuclotron Accelerator in Laboratory of High Energies of Joint Institute for Nuclear Research in Dubna, Russia. Neutrons generated by the spallation reactions of 0.7, 1.0, 1.5 and 2 GeV protons and lead target interact with subcritical uranium blanket. In the neutron field outside the blanket, radioactive iodine, neptunium, plutonium and americium samples were irradiated and transmutation reaction yields (residual nuclei production yields) have been determined using 𝛾-spectroscopy. Neutron field's energy distribution has also been studied using a set of threshold detectors. Results of transmutation studies of ¹²⁹I, ²³⁷Np, ²³⁸Pu, ²³⁹Pu and ²⁴¹Am are presented.

Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in ($X, Y$) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.

In the proposed accelerator driven systems (ADS) the possible use of several milliamperes of protons of about 1 GeV incident on high mass targets like the molten lead–bismuth eutectic is anticipated to pose radiological problems that have so far not been encountered by the radiation protection community. Spallation reaction products like high energy gammas, neutrons, muons, pions and several radiotoxic nuclides including Po-210 complicate the situation. In the present paper, we discuss radiation safety measures like bulk shielding, containment of radiation leakage through ducts and penetration and induced activity in the structure to protect radiation workers as well as estimation of sky-shine, soil and ground water activation, release of toxic gases to the environment to protect public as per the stipulations of the regulatory authorities. We recommend the application of the probabilistic safety analysis technique by assessing the probability and criticality of different hazard-initiating events using HAZOP and FMECA.

Monte Carlo calculations have been performed using MCNP code to study the optimization of photo-neutron yield for different electron beam energies impinging on Pb, W and Ta cylindrical targets of varying thickness. It is noticed that photo-neutron yield can be increased for electron beam energies $\geq 100$ MeV for appropriate thickness of the target. It is also noticed that it can be maximized by further increasing the thickness of the target. Further, at higher electron beam energy heat gradient in the target decreases, which facilitates easier heat removal from the target. This can help in developing a photo-neutron source based on electron LINAC by choosing appropriate electron beam energy and target thickness to optimize the neutron flux for ADS, transmutation studies and as high energy neutron source etc. Photo-neutron yield for different targets, optimum target thickness and photo-neutron energy spectrum and heat deposition by electron beam for different incident energy is presented.

In this report, the purpose and status of the currently constructed ITEP experimental accelerator driven system (XADS) are discussed. This hybrid electro-nuclear facility of moderate power integrates the pulse proton linac (36 MeV, 0.5 mA) and heavy water sub-critical blanket assembly (heat power of 100 kW). Most parts of the equipment units are ordered for industrial manufacturing and some are under development. The facility is supposed to be used for investigations of a wide range of problems concerning both the target-blanket assembly and the accelerator-driver and at the same time explore the dynamical processes arising during their combined operation. Some other applications of the proton beam and neutron source are also discussed. It is possible in future to increase the current and energy of proton or heavy ion beam.

The development of accelerator driven sub-critical systems (ADSS) require significant amount of new nuclear data in extended energy regions as well as for a variety of new materials. This paper reviews these perspectives in the Indian context.

One of the outstanding new developments in the field of partitioning and transmutation (P&T) concerns accelerator-driven systems (ADS) which consist of a combination of a high-power, high-energy accelerator, a spallation target for neutron production and a sub-critical reactor core. The development of the commercial critical reactors of today motivated a large effort on nuclear data up to about 20 MeV, and presently several million data points can be found in various data libraries. At higher energies, data are scarce or even non-existent. With the development of nuclear techniques based on neutrons at higher energies, nowadays there is a need also for higher-energy nuclear data. To provide alternative to this lack of data, a wide program on neutron-induced data related to ADS for P&T is running at the 20–180 MeV neutron beam facility at `The Svedberg Laboratory' (TSL), Uppsala. The programme encompasses studies of elastic scattering, inelastic neutron production, i.e., ($n, xn^{'}$) reactions, light-ion production, fission and production of heavy residues. Recent results are presented and future program of development is outlined.

Experimental results of transmissions for the samples of natural molybdenum with thickness 0.0192 atoms/barn and for the four samples of natural tantalum with thickness 0.0222, 0.0111, 0.0055 and 0.0025 atoms/barn are presented in this work. Measurements were carried out at the Pohang Neutron Facility which consists of a 100 MeV Linac, water-cooled tantalum target, and 12 m flight path length. Effective total cross-sections were extracted from the transmission data, and resonance parameters were obtained by using the code SAMMY. The present measurements were compared with other measurements and with the evaluated nuclear data file ENDF/B-VI.8.

The work is aimed at experimental determination of the independent and cumulative yields of radioactive residual nuclei produced in intermediate-energy proton-irradiated thin targets made of highly isotopic enriched and natural lead ($^{206,207,208,nat}$Pb) and ²⁰⁹Bi. 5972 radioactive product nuclide yields have been measured in 55 thin targets induced by 0.04, 0.07, 0.10, 0.15, 0.25, 0.6, 0.8, 1.2, 1.4, 1.6 and 2.6 GeV protons extracted from the ITEP U-10 proton synchrotron. The measured data have been compared with data obtained at other laboratories as well as with theoretical simulations by seven codes. We found that the predictive power of the tested codes is different but is satisfactory for most of the nuclides in the spallation region, though none of the codes agree well with the data in the whole mass region of product nuclides and all should be improved further.

The set-up `energy plus transmutation', consisting of a thick lead target and a natural uranium blanket, was irradiated by relativistic proton beams with the energy from 0.7 GeV up to 2 GeV. Neutron field was measured in different places of this set-up using different activation detectors. The possibilities of using the obtained data for benchmark studies are analyzed in this paper. Uncertainties of experimental data are shown and discussed. The experimental data are compared with results of simulation with MCNPX code.

A beam of 1 GeV proton coming from Dubna Nuclotron colliding with a lead target surrounded by 6 cm paraffin produces spallation neutrons. A Th-foil was kept on lead target (neutron spallation source) in a direct stream of neutrons for activation and other samples of ¹⁹⁷Au, ²⁰⁹Bi, ⁵⁹Co, ¹¹⁵In and ¹⁸¹Ta were irradiated by moderated beam of neutrons passing through 6 cm paraffin moderator. The gamma spectra of irradiated samples were analyzed using gamma spectrometry and DEIMOS software to measure the neutron cross-section. For this purpose neutron fluence at the positions of samples is also estimated using PREPRO software. The results of cross-sections for reactions ²³²Th($n, \gamma$), ²³²Th($n, 2n$), ¹⁹⁷Au($n, \gamma$), ¹⁹⁷Au($n, \alpha$), ¹⁹⁷Au($n, xn$), ⁵⁹Co($n, \alpha$), ⁵⁹Co($n, xn$), ¹⁸¹Ta($n, \gamma$) and ¹⁸¹Ta($n, xn$) are given in this paper. Neutronics validation of the Dubna Cascade Code is also done using cross-section data by other experiments.

Dubna Cascade Code (version-2004) has been used for the Monte Carlo simulation of the 1500 MW$_{t}$ accelerator driven sub-critical system (ADS) with ²³³U + ²³²Th fuel using the IAEA benchmark. Neutron spectrum, cross-section of $(n, xn)$ reactions, isotopic yield, heat spectra etc. are simulated. Many of these results that help in understanding the IAEA benchmark are presented. It is revealed that the code predicts the proton beam current required for the 1500 MW$_{t}$ ADS for $K_{\text{eff}} = 0.98$ to be 11.6 mA. Radial distribution of heat is fairly in agreement with other codes like the EA-MC and it needs nearly 1% less enrichment than given by other codes. This may be because the code takes care of the role of larger order of the $(n, xn)$ reactions. It is emphasized that there is a strong need to study $(n, xn)$ reactions both theoretically and experimentally for better design.

The paper discusses the basic design of the critical facility, whose main purpose is the physics validation of AHWR. Apart from moderator level control, the facility will have shutdown systems based on shutoff rods and multiple ranges of neutron detection systems. In addition, it will have a flux mapping system based on 25 fission chambers, distributed in the core. We are planning to use this reactor for experiments with a suitable source to simulate an ADS system. Any desired sub-criticality can be achieved by adjusting the moderator level. Apart from perfecting our experimental techniques, in simple configurations, we intend to study the one-way coupled core in this facility. Preliminary calculations, employing a Monte Carlo code TRIPOLI, are presented.

At BARC, development of a Low Energy High Intensity Proton Accelerator (LEHIPA), as front-end injector of the 1 GeV accelerator for the ADS programme, has been initiated. The major components of LEHIPA (20 MeV, 30 mA) are a 50 keV ECR ion source, a 3 MeV Radio Frequency Quadrupole (RFQ) and a 20 MeV drift tube linac (DTL). The Low Energy Beam Transport (LEBT) and Medium Energy Beam Transport (MEBT) lines match the beam from the ion source to RFQ and from RFQ to DTL respectively. Design of these systems has been completed and fabrication of their prototypes has started. Physics studies of the 20{1000 MeV part of the Linac are also in progress. In this paper, the present status of this project is presented.

Department of Atomic Energy, India has taken up the development of ADS in view of many attractive features like inherent safety, capability to transmute large quantities of nuclear waste, better utilization of thorium etc. A roadmap has been finalized for the development of ADS. One of the key components of the ADS is the spallation target. Considering the neutron yield, thermal-hydraulics and radiation damage issues, we are proposing to develop spallation target based on heavy density liquid metals like lead and lead-bismuth-eutectic (LBE). Both window and windowless target configurations are presently being studied. In view of the various advantages we are also studying liquid metal flow circulation based on gas lift mechanism. An R&D programme has been initiated to address various physics and technology issues of ADS target. Under this programme, mercury and LBE experimental facilities are presently being set up. Along with these facilities, computational tools related to spallation physics (FLUKA) and CFD are being developed, and the existing ones are utilized to design the entire target loop as well as sub-systems. In this presentation the details of these activities are presented.

BARC has recently proposed a one-way coupled ADS reactor. This reactor requires typically $\sim 1$ GeV proton beam with 2 mA of current. Approximately 8 kW of heat is deposited in the window of the target. Circulating liquid metal target (lead/lead-bismuth-eutectic) has to extract this heat and this is a critical R&D problem to be solved. At present there are very few accelerators, which can give few mA and high-energy proton beam. However, accelerators with low energy and hundreds of micro-ampere current are commercially available. In view of this, it is proposed in this paper to simulate beam window heating of $\sim 8$ kW in the target with low-energy proton beam. Detailed thermal analysis in the spallation and window region has been carried out to study the capability of heat extraction by circulating LBE for a typical target loop with a proton beam of 30 MeV energy and current of 0.267 mA. The heat deposition study is carried out using FLUKA code and flow analysis by CFD code. The detailed analysis of this work is presented in this paper.

The mechanical design of the target module of an accelerator driven sub-critical nuclear reactor system (ADSS) calls for an analysis of the related thermal-hydraulic issues because of large amount of heat deposition in the spallation region during the course of nuclear interactions with the molten lead bismuth eutectic (LBE) target. The LBE also should carry the entire heat generated as a consequence of the spallation reaction. The problem of heat removal by the LBE is a challenging thermal-hydraulic issue. For this, one has to examine the flows of low Prandtl number fluids (LBE) in a complex ADSS geometry. In this study, the equations governing the laminar flow and thermal energy are solved numerically using the streamline upwind Petrov-Galerkin (SUPG) finite element (FE) method. The target systems with a straight and a nozzle guide have been considered. The principal purpose of the analysis is to trace the flow and temperature distribution and thereby to check the suitability of the flow guide in avoiding the recirculation or stagnation zones in the flow space that may lead to hot spots.