Dynamic PRA Prospects for the Nuclear Industry

This review paper highlights approaches and tools available to the nuclear industry for dynamic probabilistic risk assessment (DPRA) using dynamic event trees. DPRA is an emerging methodology that has advantages as compared to traditional, static PRA predominantly owing to the addition of time dependent modeling. Traditional PRAs predefine events and outcomes into Event Trees (ET) and Fault Trees (FT), that are coupled with various combinations of Initiating Events (IE), Top Events (TE), branches, end states and sequences. A more complete depiction of the system and accident progression behavior can be quantified using DPRA to account for dynamic events such as those involving human actions. This paper discusses the strengths and needs of existing DPRA tools to align with the risk informed methodology currently used in the nuclear industry. DPRA is evolving during an exciting time in the nuclear industry with emerging advanced reactor designs also coming on the scene. Advanced nuclear (Gen IV) designs often incorporate passively safe systems that have less readily available data for traditional PRA due to their limited operating history. DPRA is a promising methodology that can address this challenge and demonstrate to the regulatory bodies and public that advanced designs operate within safety margins. In this light, the paper considers the historical role of PRA in the nuclear industry and motivation for considering dynamic PRA models. An introduction to the differences inherent in DPRA and how it complements and enhances existing PRA approaches is discussed. Additionally, a review of research from U.S national laboratories and universities features recent DPRA tool advancements that could be applied in the nuclear industry. These DPRA approaches and tools are summarized and examined to thoughtfully provide a path forward to best leverage existing research and integrate DPRA into advanced reactor design and analysis.

Nuclear Micro Power Reactor: The New Generation of Innovative Small Reactors

This short communication gives an overall account of small modular reactors and then walks through the nuclear micro reactors as the next generation of small modular reactors, which is the next wave of innovation for these SMRs. These next wave ride on the fact that future nuclear reactors are getting smaller and modular as well transportable. In this paper we are covering a summary and overall aspect of Generation IV (GEN-IV) or they are also known as Small Modular Reactors (SMRs) as well. In this book, we also, cover Nuclear Micro Reactor and its need and implementation within Department of Defense (DOD) military organizations.

Multi-Physics Model Development for Polonium Transport Behavior in a Lead-Cooled Fast Reactor

210Po, a highly toxic element with strong volatility, is one of the main source terms of a Gen-IV lead-cooled fast reactor (LFR). Therefore, the radioactive safety caused by 210Po has become an important topic in LFR-related research. In order to simulate the behavior of 210Po in an LFR, this work developed a multi-physics model of an LFR from the perspective of radioactive transport. Considering the effects of nuclide decay, cover gas leakage, containment ventilation, and Po aerosol deposition, a comprehensive simulation was carried out to evaluate the sensitivity of those effects on the 210Po distribution in detail. Preliminary results indicate that during normal operation, most of the 210Po in the LBE exist in the form of PbPo, and around 10–9 of 210Po could evaporate from the LBE into the cover gas, and then further leak into the containment. In addition, even if the leakage rate of 210Po in the cover gas into the containment is maintained at 5‰ per day, due to the deposition of Po aerosol, the 210Po contamination on the inner surface of the containment is still below the radioactivity concentration limits.

New Reactor Safety Measures for the European Sodium Fast Reactor. Part II: Preliminary Assessment

The European project ESFR SMART offers innovative options of a sodium fast reactor to improve its safety. This paper explains the results of preliminary calculations made of the main options to verify the big lines of their feasibility. Design propositions and calculations are here provided of following innovative options: removal of the safety vessel, innovative decay heat removal systems, core catcher, thermal pumps and secondary loops. In conclusion, all these options seem able to fulfil the big lines of new safety rules for GEN-IV reactors. A status of the R&D necessary to validate these new options is also proposed.

A Heuristic for Designing Hybrid Compact Heat Exchangers for Nuclear Applications

Hybrid compact heat exchangers (HCHEs) are a potential source of innovation for intermediate heat exchangers in nuclear industry, with HCHEs being designed for Gen-IV nuclear power applications. Compact heat exchangers are commonly fabricated using diffusion bonding, which can provide challenges for HCHEs due to resultant non-uniform stress distributions across hybrid structures during bonding, leading to variations in joint properties that can compromise performance and safety. In this paper, we introduce and evaluate a heuristic for determining whether a feasible set of diffusion bonding conditions exist for producing HCHE designs capable of meeting regulatory requirements under nuclear boiler and pressure vessel codes. A diffusion bonding model for predicting pore elimination and structural analyses are used to inform the heuristic and a heat exchanger design for 316 stainless steel is used to evaluate the efficacy of the heuristic to develop acceptable diffusion bonding parameters. A set of diffusion bonding conditions were identified and validated experimentally by producing various test coupons for evaluating bond strength, ductility, porosity, grain size, creep rupture, creep fatigue and channel deviation. A five-layer hybrid compact heat exchanger structure was fabricated and tensile tested demonstrating that the bonding parameters satisfy all criteria in this paper for diffusion bonding HCHEs with application to the nuclear industry.

A Review of the Surface Modifications for Corrosion Mitigation of Steels in Lead and LBE

The review paper starts with the applications of liquid metals and then concentrates on lead and lead–bismuth eutectic used in Gen IV nuclear reactors and accelerator-driven systems. Key points of degradation modes of austenitic stainless steels and ferritic-martensitic steels, candidates for the structural components, are briefly summarized. Corrosion and liquid metal embrittlement are critical issues that must be overcome. Next, the paper focuses on the strong efforts paid to the mitigation of corrosion and reviews the different solutions proposed for the protection of steels in lead and lead–bismuth eutectic. There exist promising solutions based on protection by deposition of protective coatings or protection by “natural” oxidation resulting from optimized chemical composition of the steels. However, the solutions have to be confirmed especially by longer-term experiments and by additional mechanical testing.

PBMR-400 BENCHMARK SOLUTION OF EXERCISE 1 AND 2 USING THE MOOSE BASED APPLICATIONS: MAMMOTH, PRONGHORN

High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR’s safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants’ solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs.