Conceptual Design of Sulfur-Iodine Hydrogen Production Cycle of Korea Institute of Energy Research

The Sulfur-Iodine thermochemical cycle offers a promising approach to the high efficiency production of hydrogen from nuclear power. Several SI cycles have been proposed by several research group. General Atomic (GA) studied I2 separation by extractive distillation using H3PO4. RWTH introduced the concept of reactive distillation. In this process, HIx stream coming from the Bunsen reaction is fed to the column. And HIx is distillated and decomposed at the same time to obtain hydrogen. Korea Institute of Energy Research (KIER) and Japan Atomic Energy Agency (JAEA) concentrate HIx using electro-dialysis cell and concentrated HIx is fed to the column to produce HI vapor, which is decomposed to produce hydrogen. HI was separated from HIx solution by an extractive distillation using H3PO4. However, a large amount of electric energy was required to recycle H3PO4. Most of SI processes have difficulties producing hydrogen because it has excess iodine in HI decomposition Section. SI cycle with electrodialysis cell uses membrane reactor to separate H2 and HIx. The current state of the membrane technology is not compatible with the process needs. This study examined several cases of flowsheets to overcome the problems mentioned above. The flowsheets were revised by adding the iodine separator and excluding membrane reactor. The thermal efficiency of SI process was analyzed using the revised flowsheet.

HTGR Component Test Facility (CTF) Feasibility Study for NGNP

The Next Generation Nuclear Plant (NGNP) Project is a US-based initiative led by Idaho National Laboratories to demonstrate the viability of using High Temperature Gas-Cooled Reactor (HTGR) technology for the production of high temperature steam and/or heat for applications such as heavy oil recovery, process steam/cogeneration and hydrogen production. A key part of the NGNP Project is the development of a Component Test Facility (CTF) that will support the development of high temperature gas thermal-hydraulic technologies as applied in heat transport and heat transfer applications in HTGRs. These applications include, but are not limited to, primary and secondary coolants, direct cycle power conversion, co-generation, intermediate, secondary and tertiary heat transfer, demonstration of processes requiring high temperatures as well as testing of NGNP specific control, maintenance and inspection philosophies and techniques. The feasibility of the envisioned CTF as a development and testing platform for components and systems in support of the NGNP was evaluated. For components and systems to be integrated into the NGNP full scale or at least representative size tests need to be conducted at NGNP representative conditions, with regards to pressure, flow rate and temperature. Typical components to be tested in the CTF include heat exchangers, steam generators, circulators, valves and gas piping. The Design Data Needs (DDNs), Technology Readiness Levels (TRLs) as well as Design Readiness Levels (DRLs) prepared in the pre-conceptual design of the NGNP Project and the NGNP lifecycle requirements were used as inputs to establish the CTF Functional and Operating Requirements (F&ORs). The existing South African PBMR test facilities were evaluated to determine their current applicability or possible modifications to meet the F&ORs of the CTF. Three concepts were proposed and initial energy balances and layouts were developed. This paper will present the results of this CTF study and the ongoing efforts to establish the CTF.

A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

Development of an ASME PRA Standard for Non-LWRs

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.

Evaluation of Design, Leak Monitoring, and NDE Strategies to Assure PBMR Helium Pressure Boundary Reliability

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.

Corrosion of the Materials in Sulfuric Acid

The program for hydrogen production with high temperature nuclear heat has been launched in Korea since 2004. Iodine sulfur (IS) process is one of the promising processes for a hydrogen production because it does not generate a carbon dioxide and massive hydrogen production may be possible. However, the highly corrosive environment of the process is barrier to the application in the industry. Therefore, corrosion behaviors of various materials were evaluated in sulfuric acid to select appropriate materials compatible with the IS process. The materials used in this work were Ni base alloys, Fe-Si alloys, Ta, Au, Pt, Zr, SiC and so on. The test environments were boiling 50wt% sulfuric acid without/with HI as an impurity and 98wt% sulfuric acid. The surface morphologies and cross sectional areas of the corroded materials were examined by using SEM equipped with EDS. From the results of the weight loss and potentiodynamic experiments, it was found that a Si enriched oxide is attributable to a corrosion resistance for materials including Si in boiling 98wt% sulfuric acid. Moreover the passive Si enriched film thickness increased with the immersion time leading to an enhancement of the corrosion resistance. Corrosion behaviours of the material tested are discussed in terms of the chemical composition of the materials, a corrosion morphology and the surface layer’s composition.

Corrosion Issues of HTR Structural Metallic Materials

Within the framework of the ANTARES program, AREVA NP, EDF and the CEA have launched a joint R&D program on metallic materials for VHTR. Reference alloys for circuit and Intermediate Heat eXchanger (IHX) are nickel-based with about 22%wt. of chromium. Compatibility with the HTR primary helium appears to be a determining property for the material selection and qualification. The coolant is actually polluted by a low level of impurities that can interact with metals at high temperature. Oxidation, carburization and/or decarburization occur, in relation to atmosphere characteristics, temperature and alloy chemical composition. As these corrosion effects can notably influence the mechanical properties, they often are determining to the component service life. Since the corrosion behavior is highly sensitive to environmental conditions, material studies require dedicated facilities that shall allow for a strict control of the environment throughout the entire specimen exposure. AREVA NP, CEA, and EDF have developed experimental loops respectively under the names the Chemistry Loop, CORINTH and CORALLINE, ESTEREL; these high temperature helium flow systems are equipped with high accuracy hygrometers and gas analyzers. A benchmark was defined to cross-validate the lab devices and procedures. It is composed of two tests. The joint protocol sets the operating parameters in terms of material, specimen preparation, temperature and heating program, gas pressure and flow rate, time, gas composition. Corrosion is assessed by mass change associated to observations and analyses of the corroded coupons considering the surface scales (nature, morphology and thickness), the internal oxidation (nature, distribution and depth) and the possible carburization/decarburization (type and depth). For benchmark test 1, AREVA NP, CEA, and EDF produced similar results in terms of operation of the tests as well as about the corrosion criteria. On the other hand, benchmark test 2 showed a difference in the residual water vapor level between loops that was shown to strongly influence the specimen behavior. Discrepancies in the alloy corrosion were studied regarding gas flow rates and effective oxygen potential in helium. As a consequence, the experimental tools and procedures have been upgraded. French laboratories have now efficient corrosion facilities and methods at their disposal to study and qualify the corrosion behavior of structural materials in HTR environment.

Fabrication, Properties and Irradiation Performance of Molded Block Fuel Elements for HTGRs

The molded block fuel element (FE) also called monolith is a molded body, consisting of a substantially isotropic highly crystalline graphite matrix, fuel regions within the same matrix and cooling channels. The fuel regions contain the fuel in the form of coated particles which are well bonded to the remaining graphite matrix, so that both parts of the block form a monolithic structure. The monolith meets the requirements for the very high temperature reactors attaining helium outlet temperatures above 1000°C. To fabricate the molded blocks FE demonstration plant was erected and put into operation. The equipment worked without malfunction. The produced block FEs meet the specifications of GA machined block FEs. All specimens and block segments irradiated at temperature up to 1600°C and max. fast fluence E > 0, 1 MeV of 11×1021 n/cm2 show perfect behaviour without any damage.

The Feasibility Study of AGR 7-Position Fuel Testing Assembly in NEFT Position

In order to support the Next Generation Nuclear Plant (NGNP) Program 2018 deployment schedule, the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program must reduce the AGR fuel irradiation testing time in the Advanced Test Reactor (ATR) from approximately 2 1/2 calendar years to 1 1/2 calendar years. The AGR fuel irradiation testing requirements are: (a) burn-up of at least 14% FIMA; (b) Fast neutron fluence (E > 0.18 MeV) – maximum < 5.1 × 1025 n/m2; (c) limit of fission power density is 350 W/cc; and (d) irradiation time < 1 1/2 calendar years. The accelerated testing could be accomplished by utilizing the ATR North East flux trap (NEFT) position, which can provide more control of the thermal neutron flux rate than the ATR B-10 position currently being used for the AGR-1 fuel testing, which is regulated to achieve the fuel temperature and burn-up rate requirements. In addition, the Fast (E > 1.0 MeV) to Thermal (E < 0.625 eV) neutron flux ratio (F/T) for the NEFT is much harder (higher) than the F/T ratio for the B-10 position. Thus, an appropriate configuration of Beryllium (Be) and water will need to be determined in order to soften (lower) the F/T ratio to the desired value. The proposed AGR 7-position fuel test configuration in the NEFT will utilize a graphite holder consisting of six fuel specimen positions arranged around the perimeter of the graphite holder with a seventh fuel specimen position in the center of the holder. To soften the neutron spectrum in the fuel compacts, the water volume in the outer water annulus can be increased. To reduce the compact power density, a hafnium filter could be incorporated around the graphite holder. After several trials, a hafnium filter with a thickness of 0.008 inches appeared to adequately reduce the power density to achieve the fuel testing requirements. It was also determined that the chosen beryllium-tube and water annulus configuration would adequately soften the neutron spectrum to achieve the fuel testing requirement. This neutronics study is based upon typical ATR cycle operation of 50 effective full power days (EFPD) per cycle for seven proposed irradiation cycles, and a NE lobe power of the 14 MW. The MCWO-calculated fuel compact power density, burnup (% FIMA), and fast neutron fluence (E > 0.18 MeV) results indicate that the average fuel compact burnup and fast neutron fluence reach 14.79% FIMA and 4.16 × 1025 n/m2, respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 × 1025 n/m2, which satisfied the fuel testing requirements. It is therefore concluded that accelerating the AGR fuel testing using the proposed AGR 7-position fuel test configuration in the NEFT is very feasible.

Regulatory Viewpoint on Innovative VHTR Development in Korea

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.