HIGH TEMPERATURE OXIDATION OF NB-CONTAINING ZR ALLOY CLADDING IN LOCA CONDITIONS

Chuto, Toshinori,Nagase, Fumihisa,Fuketa, Toyoshi Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.2

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

HIGH TEMPERATURE OXIDATION OF NB-CONTAINING ZR ALLOY CLADDING IN LOCA CONDITIONS

TOSHINORI CHUTO,FUMIHISA NAGASE,TOYOSHI FUKETA 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.2

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were M5® and ZIRLO™, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though M5® shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup M5® and ZIRLO™ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

DETERMINATION OF THE TRANSURANIC ELEMENTS INVENTORY IN HIGH BURNUP PWR SPENT FUEL SAMPLES BY ALPHA SPECTROMETRY

Joe, Kih-Soo,Song, Byung-Chul,Kim, Young-Bok,Han, Sun-Ho,Jeon, Young-Shin,Jung, Euo-Chang,Jee, Kwang-Yong Korean Nuclear Society 2007 Nuclear Engineering and Technology Vol.39 No.5

The contents of transuranic elements in high-burnup spent fuel samples were determined. The activity amounts of $^{238}Pu,\;^{239}Pu,\;^{240}Pu,\;^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ were measured by alpha spectrometry using $^{242}Pu\;and\;^{243}Am$ as tracers, respectively. A spike addition method for $^{237}Np$ was established by an alpha and gamma spectrometry using $^{239}Np$ as a spike after the optimum conditions for the measurements of $^{237}Np\;and\;^{239}Np$, respectively, were obtained. A separation system using anion exchange chromatography and diethylhexylphosphoric acid extraction chromatography was applied for the separation of these elements. This method was applied to high-burnup spent nuclear fuel samples $(40{\sim}60GWD/MTU)$. The contents of the transuranic elements were compared with those by ORIGEN-2 code. Measurements and the calculations of the contents of the plutonium isotopes $^{238}Pu,\;^{239}Pu\;and\;^{240}Pu$ agreed to within 10% on average. The contents of $^{237}Np$ agreed to within approximately 5% except for one instance of a calculation, while those of $^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ showed higher values by approximately 19%, 35% and 14% on average, respectively, compared to the calculations according to the burnup.

Study on the effect of long-term high temperature irradiation on TRISO fuel

Shaimerdenov Asset,Gizatulin Shamil,Dyussambayev Daulet,Askerbekov Saulet,Ueta Shohei,Aihara Jun,Shibata Taiju,Sakaba Nariaki 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO2 kernel was carried out under high-temperature gas-cooled reactor (HTGR)- like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950 e1100 C. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel

Modeling of flow blockage due to swell and rupture for analysis of LBLOCA using a multiple fuel rods scheme under high burnup condition

Bang, Young Seok,Lee, Joosuk Elsevier 2019 Annals of nuclear energy Vol.134 No.-

<P><B>Abstract</B></P> <P>The present study discusses a model of Large Break Loss-Of-Coolant Accident (LBLOCA) calculation using the multiple fuel rods modeling scheme which was developed to consider the fuel rods with different burnup in preparation for the recently proposed revision of Emergency Core Cooling System (ECCS) rule in the United States of America and in Korea. The thermal conductivity of the fuel pellet, the initial oxide layer thickness of the cladding, the size of gap and the rod internal pressure, and the effective thermal conductivity of the cladding considering the oxide layer are reflected by an input as functions of the burnup. Further, through the analysis on the division of hydraulic channels in which the multiple fuel rods are located, the influences of division of the hydraulic channel and the crossflow between channels on cladding thermal response are evaluated. Also a modeling scheme to calculate the flow blockage due to swell and rupture of cladding for the current system thermal-hydraulic code such as MARS-KS code is discussed. To consider the uncertainty of the swell and rupture model of the code and the uncertainty in the present flow blockage modeling scheme, a multiplication factor is introduced and used to evaluate the influence of the flow blockage on ECCS performance.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A multiple fuel rods modeling scheme was developed considering the various burnup conditions. </LI> <LI> Effect of hydraulic nodalization of the core including crossflow was evaluated. </LI> <LI> A modeling scheme to calculate the core flow blockage was developed. </LI> <LI> Effect of flow blockage in the core on thermal response of the fuel was evaluated. </LI> </UL> </P>

DETERMINATION OF THE TRANSURANIC ELEMENTS INVENTORY IN HIGH BURNUP PWR SPENT FUEL SAMPLES BY ALPHA SPECTROMETRY-II

Joe, Kih-Soo,Song, Byung-Chul,Kim, Young-Bok,Jeon, Young-Shin,Han, Sun-Ho,Jung, Euo-Chang,Song, Kyu-Seok Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.1

The contents of transuranic elements ($^{237}Np$, $^{238}Pu$, $^{239}Pu$, $^{240}Pu$, $^{241}Am$, $^{244}Cm$, and $^{242}Cm$) in high-burnup spent fuel samples ($35.6{\sim}53.9\;GWd/MtU$) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about $2.6{\sim}32.7%$ on average according to each isotope, and those for americium and curium were also higher by about $35.9{\sim}63.1%$. However, for $^{237}Np$, the measurements were lower by about 52% on average for the samples.

DETERMINATION OF THE TRANSURANIC ELEMENTS INVENTORY IN HIGH BURNUP PWR SPENT FUEL SAMPLES BY ALPHA SPECTROMETRY-II

조기수,BYUNG-CHUL SONG,YOUNG-BOK KIM,YOUNG-SHIN JEON,한선호,정의창,송규석 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.1

The contents of transuranic elements (237Np, 238Pu, 239Pu, 240Pu, 241Am, 244Cm, and 242Cm) in high-burnup spent fuel samples (35.6~53.9 GWd/MtU) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about 2.6~32.7% on average according to each isotope, and those for americium and curium were also higher by about 35.9~63.1%. However, for 237Np, the measurements were lower by about 52% on average for the samples.

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2 AND MOX FUEL

BYUNG-HO LEE,구양현,JAE-YONG OH,Jin-SikCheon,YOUNG-WOOK TAHK,손동성 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.6

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO₂ AND MOX FUEL

Lee, Byung-Ho,Koo, Yang-Hyun,Oh, Jae-Yong,Cheon, Jin-Sik,Tahk, Young-Wook,Sohn, Dong-Seong Korean Nuclear Society 2011 Nuclear Engineering and Technology Vol.43 No.6

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

Validation of Serpent-SUBCHANFLOW-TRANSURANUS pin-by-pin burnup calculations using experimental data from the Temelín II VVER-1000 reactor

Garcia, Manuel,Vocka, Radim,Tuominen, Riku,Gommlich, Andre,Leppanen, Jaakko,Valtavirta, Ville,Imke, Uwe,Ferraro, Diego,Uffelen, Paul Van,Milisdorfer, Lukas,Sanchez-Espinoza, Victor Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.10

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.