The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit (PCU) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger is very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical coil heat exchanger, and shell/tube heat exchanger.

In this article an advanced fuel cycle for pebble bed reactors is introduced that can safely and efficiently incinerate pure reactor-grade Pu [Pu(LWR)], thereby fulfilling the bulk of the GNEP waste incineration requirements. It is shown below that the very high fissile content of the Pu(LWR)-fuel enables it to convert practically all of the 240 Pu to 241 Pu and incinerate it. Since the fuel contains no 238 U, no fresh 239 Pu is produced. The 239 Pu is reduced in-situ by 99.5% and the 240 Pu by 97.6%. The only significant fissile isotope remaining is 241 Pu, however, it will decay with a half life of 14.4 years to the fertile 241 Am by β -decay.

The Very High Temperature Reactor (VHTR) has been selected by the U.S. as the Generation IV technology for the Next Generation Nuclear Plant (NGNP), and also by the Republic of Korea for the Nuclear Hydrogen Development and Demonstration (NHDD) project. One of the key long-lead items for the VHTR is the Reactor Pressure Vessel (RPV). In the absence of active vessel cooling, the RPV temperature during normal operation is determined by the design point selected for the primary coolant inlet temperature and the design of the reactor internal components, including the physical location of riser channels that route the coolant flow to the plenum above the reactor core. For the VHTR, the primary coolant (helium) inlet temperature is expected to be in the range 490°C to 590°C and the outlet temperature is expected to be in the range 850°C to 950°C. For the RPV, both SA-508/533 steel and higher alloy steels with higher temperature capability (e.g., 9Cr-1Mo-V steel) are being considered. Because of its extensive experience base as an ASME Section III code-approved material for Light Water Reactor (LWR) pressure vessels, SA-508/533 steel is emerging as a strong candidate for the VHTR RPV. However, in order to use this material, the RPV temperature must be maintained below ASME code limits, which are 371°C during normal operation and 538°C for up to 1000 h during accident conditions.

Over its 1968–1988 life, PSCo relicensed the Fort St. Vrain (FSV) High-temperature Gas Reactor (HTGR) for light water reactor (LWR) technology requirements. Estimates of the financial losses associated with the plant range from $500 million to $2 billion in 1980 dollars. Colorado ratepayers, the shareholders of Gulf General Atomics and its corporate successors — General Atomics, GA Technologies or just GA and Public Service Company of Colorado (PSCo) bore these losses. Two critical plant issues required solution for the plant’s economic success — (1) the high-cost of 93% enriched uranium fuel and (2) low unit availability. While fuel costs were beyond utility control, low availability was controllable, yet remained unresolved. Commercially isolated for twenty years, PSCo shut the plant down in 1988. Economic success of future HTGRs depends upon avoiding similar complications. This paper examines the issues that made FSV uneconomic, including those fundamental to HTGR technology and others attributable to the utility operator and its culture. Knowing the history of FSV and HTGR design, designers should anticipate reasonable challenges. Preparations will help manage future HTGR risks, costs, and assure operating success. Regulators and industry can assure more effective, economic operations in the next round of HTGR designs.

To measure the value of a technology investment under uncertainty with standard techniques like net present value (NPV) or return on investment (ROI) will often uncover the difficulty to present convincing business case. Projected cash flows are inefficient or the discount rate chosen to compensate for the risk is so high, that it is disagreeable to the investor’s requirements. Decision making and feasibility studies have to look beyond traditional analysis to reveal the strategic value of a technology investment. Here, a Real Option Analysis (ROA) offers a powerful alternative to standard discounted cash-flow (DCF) methodology by risk-adjusting the cash flow along the decision path rather than risk adjusting the discount rate. Within the GEN IV initiative attention is brought not only towards better sustainability, but also to broader industrial application and improved financing. Especially the HTR design is full of strategic optionalities: The high temperature output facilitates penetration into other non-electricity energy markets like industrial process heat applications and the hydrogen market. The flexibility to switch output in markets with multi-source uncertainties reduces downside risk and creates an additional value of over 50% with regard to the Net Present Value without flexibility. The supplement value of deploying a modular (V)HTR design adds over 100% to the project value using real option evaluation tools. Focus of this paper was to quantify the strategic value that comes along a) with the modular design; a design that offers managerial flexibility adapting a step-by-step investment strategy to the actual market demand and b) with the option to switch between two modes of operation, namely electricity and hydrogen production. We will demonstrate that the effect of uncertain electricity prices can be dampened down with a modular HTR design. By using a real option approach, we view the project as a series of compound options — each option depending on the exercise of those that preceded it. At each end of the design phase, the viability will be reviewed conditional on the operating spread at each time step. We quantify the value of being able to wait with the investment into a next block until market conditions are favourable and to be able to abandon one block if market conditions are disapproving. To derive the intrinsic value of this multi block HTR design, it will be compared with a reference investment of a full commitment light water reactor without any managerial flexibility. In another case, we raise the question to what extent product output diversification is a suitable strategy to cope with long term market uncertainty in electricity price. What is the value of a multi-potent technology that is able to produce output for energy markets others than the electricity market? To investigate this, we concentrate on The Netherlands, a country with an intense industrial demand in electricity and hydrogen.

This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic non-destructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed inservice inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper.

The ASME Committee on Nuclear Risk Management (CNRM) has established a working group to pursue the development of a PRA standard that can be used for advanced non-LWR plants. The applications of such PRAs include the performance of PRAs to support licensing and design decisions, and to meet NRC requirements for Design Certifications and Construction and Operating Licenses. The purpose of this paper is to summarize the significant progress that has been made to date in developing a new PRA standard for non-LWRs from the personal point of view of the working group chairman.

The Very High Temperature Reactor (VHTR) is the leading candidate for the reactor component of the Next Generation Nuclear Plant (NGNP). This is because the VHTR demonstrates great potential in improving safety characteristics, being economically competitive, providing a high degree of proliferation resistance, and producing high outlet temperatures for efficient electricity generation and/or other high temperature applications, most notably hydrogen production. In addition, different fuel types can be utilized by VHTRs, depending on operational goals. In this case, the recovery and utilization of the valuable energy left in LWR fuel in order to create ultra long life single batch cores by taking advantage of the properties of TRU fuels. This paper documents the initial process in the study of TRU fueled VHTRs, which concentrates on the verification and validation of the developed whole-core 3D VHTR models. Many of the codes used for VHTR analysis were developed without a full appreciation of the importance of randomness in particle distribution. With this in mind, the SCALE code system was chosen as the computational tool for the study. It provides the opportunity of utilizing SCALE versions 5.0 and 5.1, making it possible to compare and analyze different techniques accounting for the double heterogeneity effects associated with VHTRs. Startup physics results for Japan’s High Temperature Test Reactor (HTTR) were used for experiment-to-code benchmarking. MCNP calculations were employed for code-to-code benchmarking. Results and analysis are included in this paper.

This paper addresses that major changes in the safety approach, for instance the increased use of Probabilistic Risk Assessment (PRA), have been made. All commercial reactors in operation today belong to the Generations II and III. Generation IV International Forum (GIF) has launched several programs aimed at developing the next generation of nuclear energy systems. Part of the research effort is focused on new reactor concepts, such as the Very High Temperature Reactor (VHTR), currently developing in Korea. In parallel to the design process of VHTR currently underway, regulatory approach is moving forward to define new licensing rules. So, Korea Institute of Nuclear Safety (KINS) is defining, as a goal to risk-inform, the regulation and developing the regulatory framework and licensing process more efficient, predictable, and stable. However, the licensing of NPPs has focused until now on Light Water Reactors (LWRs) and has not incorporated systematically insights and benefits from PRA. In the meantime, USNRC and IAEA have recently drafted a risk-informed regulation and technology-neutral framework (TNF) for new plant licensing along with the innovative Gen-IV system design. KINS also expects that advanced NPPs will show enhanced margins of safety, and that advanced reactor designs will comply with the national safety goal policy statement. In order to meet these expectations, PRA tools are currently being considered by KINS; among them are frequency-consequence (F-C) curves, which plot the frequency of having Consequence. This paper discusses the role and the usefulness of such curves in risk-informing the licensing process in Korea, and shows that the use of F-C curve allows the implementation of both structural and rational Defence-In-Depth (DID). This paper focuses on F-C curves as means to assess the licensing basis events (LBEs) from the regulatory viewpoint on the innovative small and medium reactor (SMR) sized VHTR deployment in Korea. The principle underlying the F-C curve is that event frequency and dose are inversely related, i.e., the higher the dose consequences, the lower is the allowed event frequency.