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DOE OSTI2011

AFIP-1 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-1 was designed to demonstrate the performance of second-generation dispersion fuels at a prototypic scale with a length of 21.5 inches (54.6 cm), width of 2.25 inches (5.75 cm) and a thickness of 0.050 inch (0.13 cm). The experiment was fabricated using commercially standard practices at BWX Technology, Inc. (BWXT). The U-7Mo fuel particles were supplied by the Korean Atomic Energy Research Institute (KAERI) using equipment intended for commercial supply. Two fuel plates were tested that incorporated two different matrix compositions, Al-2Si and Al-4043.1 The following report summarizes the life of the AFIP-1 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

I-NERI-2007-004-K, Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Sodium Fast Reactor Safety and Licensing Research Plan – Volume II

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Neutron/gamma coupled library generation and gamma transport calculation with KARMA 1.2

KAERI has developed a lattice transport calculation code KARMA and its multi-group cross section library generation system. Recently, the multi-group cross section library generation system has included a gamma cross section generation capability and KARMA also has been improved to include a gamma transport calculation module. This paper addresses the multi-group gamma cross section generation capability for the KARMA 1.2 code and the preliminary test results of the KARMA 1.2 gamma transport calculations. The gamma transport calculation with KARMA 1.2 gives the gamma flux, gamma smeared power, and gamma energy deposition distributions. The results of the KARMA gamma calculations were compared with those of HELIOS and they showed that KARMA 1.2 gives reasonable gamma transport calculation results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of the IPRO-zone for fire PSA and its applications

A PSA analyst has been manually determining fire-induced component failure modes and modeling them into the PSA logics. These can be difficult and time-consuming tasks as they need much information and many events are to be modeled. KAERI has been developing the IPRO-ZONE (interface program for constructing zone effect table) to facilitate fire PSA works for identifying and modeling fire-induced component failure modes, and to construct a one top fire event PSA model. With the output of the IPRO-ZONE, the AIMS-PSA, and internal event one top PSA model, one top fire events PSA model is automatically constructed. The outputs of the IPRO-ZONE include information on fire zones/fire scenarios, fire propagation areas, equipment failure modes affected by a fire, internal PSA basic events corresponding to fire-induced equipment failure modes, and fire events to be modeled. This paper introduces the IPRO-ZONE, and its application results to fire PSA of Ulchin Unit 3 and SMART(System-integrated Modular Advanced Reactor). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Major results from safety-related integral effect tests with VISTA-ITL for the SMART design

A series of integral effect tests (IETs) was performed by the Korea Atomic Energy Research Inst. (KAERI) using the VISTA integral test loop (VISTA-ITL) as a small-scale IET program. Among them this paper presents major results acquired from the safety-related IETs with the VISTA-ITL facility for the SMART design. Three small-break loss-of-coolant accident (SBLOCA) tests of safety injection system (SIS) line break, shutdown cooling system (SCS) line break and pressurizer safety valve (PSV) line break were successfully performed and the transient characteristics of a complete loss of flowrate (CLOF) was simulated properly with the VISTA-ITL facility. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluation plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Decommissioning of the TRIGA mark II and III and radioactive waste management

KAERI has carried out decommissioning projects for two research reactors (KRR-1 and 2). The decommissioning project of KRR-1 (TRIGA Mark II) and 2 (TRIGA Mark III) was launched in 1997 with a total budget of 23.25 million US dollars. KRR-2 and all auxiliary facilities were already decommissioned, and KRR-1 is being decommissioned now. Much more dismantled waste is generated than in any other operations of nuclear facilities. Thus, the waste needs to be reduced and stabilized through decontamination or treatment before disposal. This paper introduces the current status of the decommissioning projects and describes the volume reduction and conditioning of decommissioning waste for final disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Waste Minimization of HyBRID Decontamination Process - 17060

The Korea Atomic Energy Research Institute (KAERI) developed a Hydrazine base reductive metal ion decontamination (HyBRID) and related pre-oxidation techniques for use in decommissioning a nuclear facility. Inorganic chemicals were used to increase the integrity of the radioactive waste in a repository. Decomposition and precipitation of residual reagents with inorganic chemicals were introduced to minimize the final waste volume. The waste volume was calculated theoretically and demonstrated in a small-scale waste treatment system and bench-scale candle filtration system. The total waste volume of the HyBRID process was less than 1/10 of that of commercial processes, and spent ion-exchange resins were significantly reduced to 5% of that of commercial processes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Facile Separation of Sr-90, Fe-55, Nb-94, Ni-59,63 with Large Amount of Ca by Controlled pH Precipitation - 17190

it is important to manage low-and intermediate-level radioactive wastes from nuclear power plants due to a permanent disposal. According to the Nuclear Safety and Security Commission (NSSC), among all the regulatory radionuclides, a few radionuclides from the fission products and corrosion products such as Sr-90, Fe-55, Nb-94, Ni-59,63 which emit low-energy x-ray and beta radiation. Thus, it is necessary to separate beta-emitting nuclides individually. For that reason, KAERI has established analytical procedures which can separate Sr-90, Fe-55, Nb-94, Ni-59,63 from radioactive waste samples. However, when radioactive wastes contain large amount of Ca (above 300 mg), it is not easy to recover the target radionuclides because high concentration of Ca has a strong influence on the effective analyses. In order to combat this problem, there was an attempt to make the precipitate as CaF{sub 2} by using hydrofluoric acid. However, this way had a limitation; Fe-55, Nb-94 were precipitated as fluoride compounds as well as Ca. In this study, we have focused on developing a separation procedure stage by stage in the radioactive wastes with Sr-90, Fe-55, Nb-94, Ni-59,63 and large amount of Ca (1500 mg). The detailed steps are described below; (1) Separation of Fe-55, Nb-94, Ni-59,63 as metal hydroxide precipitates: low solubility product constant (Ksp) between pH 6 and pH 10. (2) Selective elimination of Ca-oxalate complex under strong acidic conditions. Consequently, all the metal sources were sequentially separated and quantitatively recovered by inductively coupled plasma atomic emission spectrometric (ICP-AES) analysis, the recovery is as follows; Sr-90 (3 mg): 89.47±5.48%, Fe-55 (20 mg): 99.80±3.19%, Nb-94 (20 mg): 97.45±2.42%, Ni-59,63 (2 mg): 83.59±15.31%, Ca (1500 mg): 85.32±6.38%, respectively. This methodology will help us separation of radioactive wastes efficiently if apply properly. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Development of the computer code for multi-calculations of source term of spent fuel - 15351

Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Chemical and Foaming Stability of Foam Decontaminating Agents - 16137

This study conducted experiments to evaluate the stability of nano foaming agents by adding various types of surfactants, silica nanoparticles, and viscosifiers with the aim of enhancing the stability of foaming agents containing nanoparticles. Nonionic surfactants such as EM100 showed higher stability as compared to the Sodium Dodecyl Sulfate (SDS) anionic surfactant, and the foam stability decreased as the surfactant concentration increased. The co-surfactant addition did not significantly affect the foam stability, but the addition of Xantan gum-one of the various viscosifiers tested in this study-resulted in the highest foam stability. When silica nanoparticles were added to the surfactant EM100, the foam stability improved further. In particular, the foam stability of the nano foaming agent was observed to be the maximum when a partially hydrophobic nano-porous silica such as KAERI 1 was added, as compared to the addition of a hydrophilic dry silica such as M5. This is because the partially hydrophobic particles were distributed over the liquid film between the foams, thereby preventing drainage and increasing the stability of the foaming agent fluid. The manufactured nano foaming agent can be used not only for the effective decontamination of large-sized equipment or large facilities but also for overcoming the problem of production of a large amount of radioactive waste during chemical decontamination. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Management of Tritium in HANARO (High Flux Advanced Neutron Application Reactor) - 18113

Tritium is generated inevitably from HANARO (High flux Advanced Neutron Application Reactor), a multi-purpose research reactor at the Korea Atomic Energy Research Institute (KAERI), because HANARO uses heavy water as a reflector. The specific activity of tritium in heavy water increases as the operation time increases. The release of tritium into the environment also increases as the specific activity of tritium in heavy water increases. Although the activity of tritium released from HANARO was less than 1/1,000 of the derived release limit regulated by the regulatory authority, for the last decade, significant effort has been exercised to reduce the activity of tritium released into the environment as well as the activity of tritium inside of HANARO to protect its workers. Recently the replacement of heavy water containing tritium by fresh heavy water was reviewed to maintain the activity of tritium in heavy water as low as possible. A replacement plan is being established based on the technical estimates and reviews and will be applied soon. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Decommissioning Concept of Nuclear Fuel Cycle Facility - 18255

As the Korean Nuclear Safety Act was revised, it is urgent to prepare a decommissioning plan for the nuclear fuel cycle facilities operated by the Korea Atomic Energy Research Institute, KAERI. First, the basic concepts necessary for dismantling each facility and the order of dismantling for each dismantling work were determined. In this paper, we consider the disassembly order of the nuclear fuel cycle facilities such as the post irradiation fuel examination facility, the research reactor fuel production facility, the radioactive waste treatment facility, the radioactive waste form examination facility. The decommissioning plan for the research nuclear fuel cycle facilities will be based on the abovementioned dismantling basic concept and disassembly sequences. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Study on Establishment of Scenario Considering to Location and Treatment for Waste Tracking System

Radioactive waste of domestic generated KAERI, KHNP and KNFC etc. Agency of radioactive waste generation have each waste tracking system and waste streaming. We need to establish integrated waste classification for waste tracking and management efficiency. This study main point is the integration of radioactive waste classification and the cording of domestic generators, waste disposal site, etc. We also created a scenario through this coding. First, This study was able to identify the status of domestic radioactive Waste Tracking System. Second, through this study, WTS scenarios were prepared by analyzing each waste management agency and transportation report. If WTS scenario establish from integration of waste streaming, it will be possible tracking, back tracking and management of radioactive waste more efficiency. Third, This study coded all the sources of radioactive waste generators, acceptor agency, waste classification and treatment. Using this code, we tracking the scenario from generator to acceptor and display it as bar-code. Also, if we later convert this bar-code to a Qr-code, it thought that it will be easier tracking radioactive waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Study on Radionuclides Selection for DCGLs Development for Site Release of KRR Units 1 and 2 site in Korea - 20256

The selection of Derived Concentration Guideline Levels (DCGLs) in nuclear facilities decommissioning is important because they serve as criteria for site characterization and final status surveys. DCGLs should be derived for KAERI TRIGA Mark II and III research reactor site clearance and prior to the work, a radionuclides list for DCGLs consideration was created. Radionuclides selection for DCGLs was performed according to the procedure proposed by the US NRC based on the characterization survey related to the decommission site. By evaluating the dose contribution of each radionuclide using D and D computer code, radionuclides with a dose contribution fraction of 0.1% or more for the entire radionuclide were selected. As a result, the selected radionuclides were beta radionuclides H-3, C-14 and Sr-90 and gamma radionuclides Co-60, Cs-137, Eu-152 and Eu-154. The selected radionuclides can be presented for consultation with domestic regulatory agencies in the future and will be used as the technical base of DCGLs for the final status survey. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

An automatic input/output test experience of FPGA-based logic controller with I/O stimulator - 282

This paper describes an automatic testing approach instead of a traditional manual test for an FPGA-based Logic Controller (FLC). It mainly describes an automatic testing of FLC using an I/O stimulator with the initiated manual action. The FLC was developed by Doosan Heavy Industry and KAERI in S. Korea. We also developed a test-bed like I/O stimulator for a burning test based on the operational scenario. Signal sources from the I/O stimulator contain an analog input, an analog output, a digital input and a digital output, a programmable power supply, an RS232C serial interface, and a dual port Ethernet interface. LabVIEW Programming was used for the automatic testing. The FLC consists of a general processor module (FPM-01), a complex processor module (FPM-02), analog input modules (FAI-01), analog output modules (FAO-01), digital input modules (FDI-01), digital output modules (FDO-01), a data link module (FDL-01), and a network module (FNW-01). Among these, in this paper, automatic digital I/O testing and automatic analog I/O testing are described. In addition, for the accuracy of the automatic testing, two graphic recorders are used. One is the output results of general processor based testing, and the other is the output results of complex processor based testing. These are different company products. Redundancy recording was found in the test results. The test results were synchronized internally with the existing I/O stimulator and externally with two graphic recorders. Calibrations of all of the equipment were made before the system testing. Before the testing started, a test plan and the test procedure were set up in advance, and we performed test scenario-based automatic functional testing for the FLC as the qualification process. It can be several executed depending on the scenario and conditional parameters of the test. For validation of the accuracy and reliability in our automated test, manual testing was also performed. Manual tests were conducted in two ways. The first one was a manual test using Fluke 754 and Fluke 715. The other one was using embedded manual features of the I/O stimulator. By performing the automatic input and output testing using the I/O stimulator, the time and expense of FLC testing can be reduced significantly compared with a conventional manual test. We rechecked the automatic test results using Fluke754 manually. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗