Pakistan and the Nuclear Non-Proliferation Regime

The arrival of nuclear weapons was detrimental in changing the nature of warfare; duringWWII, we used nuclear weapons on two Japanese cities. The catastrophic effect of nuclear weapons made the pioneers apprehensive about the spread of nuclear technology across the globe. The Nuclear Non-Proliferation Regime (NNPR) is a set of international rules, norms, initiatives, agreements, arrangements, bilateral and multilateral treaties to curb the spread of nuclear weapons and technology. The Nuclear Non-Proliferation Treaty (NPT) is the backbone of NNPR. This paper discusses the significance of NNPR and will focus on how Pakistan fits into the bigger picture of the NNPR as a nuclear state. This research study will further analyze the prospects and challenges for Pakistan viz a viz the NNPR. It will specifically focus on Pakistan's official position on joining the Nuclear Non-Proliferation Treaty (NPT), Comprehensive Test Ban Treaty (CTBT), Fissile Material Cut-off Treaty (FMCT), and the two relevant Export Control Regimes (Nuclear Suppliers Group and Missile Technology Control Regime). Moreover, the challenges Pakistan faces viz-a-viz these treaties and arrangements will be highlighted. Recommendations will be provided based on the prospects of how Pakistan can overcome these challenges.

Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC

The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.

30 years of IRMM-1027 reference materials for fissile material accountancy and control: development, production and characterisation

The IRMM-1027 Large-sized dried (LSD) spikes are certified reference materials applied to measure the uranium and plutonium content of dissolved fuel solutions using isotope dilution mass spectrometry. High quality starting metals of uranium and plutonium are dissolved to produce a stock solution, which is dispensed into individual vials and dried down. The present spikes are mixtures of typically 50 mg 20% enriched U and 2 mg enriched 239Pu as dried nitrates, conditioned in an organic substance for stability. Each vial of an IRMM-1027 LSD spike comes with certified masses of uranium (235U and 238U) and 239Pu and isotopic composition with associated uncertainty. This article will discuss major developments since the production of the first batch of LSD spikes and will reflect on the current preparation and certification approaches.

Neutronic Evaluation of MSBR System Using MCNP Code

The concept of Molten Salt Reactor use Th to breed fissile 233U, where an initial source of fissile material needs to be provided. However, there is no available 233U and so; the fissile fuel supply is one of the unresolved problems. Thus, it is necessary to use existing fissile materials such as 235U or Pu to produce 233U. Current studies analyze the fuel transition from 235U/Th or Pu/Th to 233U/Th and, in this context, the present work evaluates the criticality and the neutron flux of MSBR (Molten Salt Breeder Reactor) considering the fuel: (i) mix of Th and enriched U; (ii) the combination of Th and reprocessed Pu; and (iii) matrix of reprocessed Pu/minor actinides (MAs) and Th. The goal is to verify which of these fuels can be used as initial fissile supply. The MSBR core was simulated by MCNPX 2.6.0 code and the criticality model presents similar behavior of previous studies. The results show that reprocessed fuels could have a potential to be used as initial fissile supply, but these fuels present a neutron flux profile less flattens than traditional 233U/Th. It is possible that a new distribution of fuel elements may improve this profile and future simulations will be performed to evaluate this behavior. The uranium, must has high enrichment value to be used as initial seed. Other studies need be performed to evaluates the uranium enrichment and the U/Th ratio that produces similar core criticality to traditional fuel.

Candidate Core Designs for the Transformational Challenge Reactor

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YHx) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YHx moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems are assessed and summarized herein.