응력이완 열처리된 지르코늄 피복관의 부식 및 인장특성에 미치는 수소화물 영향 평가

방제건 ( Je Geon Bang ),백종혁 ( Jong Hyuk Baek ),이명호 ( Myung Ho Lee ),정용환 ( Yong Hwan Jeong ) 한국열처리공학회 2004 熱處理工學會誌 Vol.17 No.6

N/A The hydrides in cladding affect the corrosion and tensile properties. In this study corrosion and tensile properties were evaluated with varying the hydrogen concentration. The charged hydrogen contents were ranged from 200 to 1000 ppm. The corrosion rate in water and LiOH solution increases with the hydrogen concentration. The hydride did not affect the corrosion mechanism in the pre-transition region, but in the post-transition region the corrosion rate was accelerated. Cladding E contained higher Niobium content was slowly accelerated compared with other claddings. The yield and ultimate strengths were independent on the hydrogen content. How-ever, the total elongation decreased gradually with increasing the hydrogen content. SEM observation of fracture surface showed that an average depth of voids decreased with increasing the hydrogen content and small secondary cracks were observed.

Separation of Fission Products by Ion Exchange Method

Lee,Byung-Hun,Bang,Je-Geon 대한방사선 방어학회 1983 방사선방어학회지 Vol.8 No.1

Ru-103, Cs-137 그리고 Ce-144를 차례로 분리하는 방법을 유기양이온 교환체 Amberite CG-120, 무기이온 교환체, silica gel과 montmorillonite를 가지고 검토하였다. Amberite CG-120으로 Ru-103, Cs-137 그리고 Ce-144를 흡착 및 탈착하는 최적조건은 각각 0.01M-, 0.01M-, 0.1M-HCl과 3M-, 3 M-, 5 M-HCl이다. Silica gel로 Ru-103, Cs-137 그리고 Ce-144를 흡착 및 탈착하는 최적 조건은 각각 pH 8, pH 8, pH 8과 3M-, 1 M-, 1 M-HCl이다. Montmorillonite로 Ru-103, Cs-137 그리고 Ce-144를 흡착 및 탈착하는 최적 조건은 각각 pH 8, 0.01 M-HCl, pH 4와 1 M-, 5 M-, 3 M-HCl이다. 낮은 이온강도에서 일어나는 흡착 그리고 탈착 이온강도의 차이는 연속적인 실험에서 혼합핵종의 분리를 위하여 이용된다. Ru-103, Cs-137 그리고 Ce-144 혼합물의 개별분리는 Amberite CG-120, silica gel에서보다 montmorillonite에서 더욱 효과적이다. The sequential separation of Ru-103, Cs-137 and Ce-144 was carried out by organic cation exchange, Amberite CG-120, and inorganic ion exchanges, silica gel and montmorillonite. The optimum conditions of Ru-103, Cs-137 and Ce-144 on Amberite CG-120 are 0.01M-, 0.01M- and 0.1M- hydrochloric acid for the adsorption , and 3M-, 3M- and 5M-hydrochloric acid for the desorption, respectively. The optimum conditions of Ru-103, Cs-137 and Ce-144 on silica gel are pH 8, pH 8 and pH 8 for the adsorption, and 3M-, 1M- and 1M-hydrochloric acid for the desorption, respectively. The optimum conditions of Ru-103, Cs-137 and Ce-144 on montmorillonite are pH 8, 0.01M-hydrochloric acid and pH 4 for the adsorption, and 1M-, 5M- and 3M-hydrochloric acid for the adsorption, respectively. The adsorption which occurs at lower ionic strength and the differences in desorption ionic strength are utilized for the separation of tracer mixture in continuous experiments. The individual separation of Ru-103, Cs-137 and Ce-144 can be carried out more efficiently with montmorillonite than with silica gel and Amberite CG-120.

Separation of Fission Products by Ion Exchange Method

Byung-Hun Lee,Je-Geon Bang 대한방사선방어학회 1983 방사선방어학회지 Vol.8 No.1

Nb가 첨가된 신형 지르코늄 피복관의 열적 크리프 거동

김준환,방제건,정용환,Kim Jun Hwan,Bang Je Geon,Jeong Yong Hwan 한국재료학회 2004 한국재료학회지 Vol.14 No.7

Thermal creep properties of the zirconium tube which was developed for high burnup application were evaluated. The creep test of cladding tubes after various final heat treatment was carried out by the internal pressurization method in the temperature range from $350^{\circ}C to 400^{\circ}C$ and from 100 to 150 MPa in the hoop stress. Creep tests were lasted up to 900days, which showed the steady-state secondary creep rate. The creep resistance of zirconium claddings was higher than that of Zircaloy-4. Factors that affect creep resistance, such as final annealing temperature, applied stress and alloying element were discussed. Tin as an alloying element was more effective than niobium due to solute hardening effect of tin. In case of advanced claddings, the optimization of final heat treatment temperature as well as alloying element causes a great influence on the improvement of creep resistance.

HIGH BURNUP FUEL TECHNOLOGY IN KOREA

Song, Kun-Woo,Jeong, Yong-Hwan,Kim, Keon-Sik,Bang, Je-Geon,Chun, Tae-Hyun,Kim, Hyung-Kyu,Song, Kee-Nam Korean Nuclear Society 2008 Nuclear Engineering and Technology Vol.40 No.1

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

지르코늄 합금의 대량수소화에 미치는 표면산화막의 영향

김선기,방제건,김대호,임익성,양용식,송근우,김용수,Kim, Sun-Ki,Bang, Je-Geon,Kim, Dae-Ho,Lim, Ik-Sung,Yang, Yong-Sik,Song, Kun-Woo,Kim, Yong-Soo 한국재료학회 2008 한국재료학회지 Vol.18 No.11

Oxide effects experiments on massive hydriding reactions of Zr alloy with hydrogen gas were carried out at $400^{\circ}C$ under 1 atm in a $H_2$ environment with a thermo-gravimetric apparatus (TGA). Experimental results for oxide effects on massive hydriding kinetics show that incubation time is not proportional to oxide thickness. The results also show that the massive hydriding kinetics of pre-filmed Zr alloys follows linear kinetic law and that the hydriding rates are similar to that of oxide-free Zr alloys once massive hydriding is initiated. Unlikely microstructure of the oxide during incubation time, physical defects such as micro-cracks and pores were observed in the oxide after incubation time. Therefore, it seems that the massive hydriding of Zr alloys can be ascribed to short circuit paths and mechanical or physical defects, such as micro-cracks and pores in the oxide, rather than to hydrogen diffusion through the oxide resulting from the increase of oxygen vacancies in the hypo-stoichiometric oxide.

HIGH BURNUP FUEL TECHNOLOGY IN KOREA

KUN WOO SONG,YONG HWAN JEONG,KEON SIK KIM,JE GEON BANG,TAE HYUN CHUN,HYUNG KYU KIM,KEE NAM SONG 한국원자력학회 2008 Nuclear Engineering and Technology Vol.40 No.1

High burn-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, UO2 pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain UO2 pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to UO2 powder, were prepared by oxidizing and heat-treating scrap UO2 pellets. A UO2 pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

링 인장시험을 이용한 지르코늄 피복관의 반응도 사고(RIA) 시 연성 평가

김준환,이명호,최병권,방제건,정용환,Kim Jun Hwan,Lee Myoung Ho,Choi Byoung Kwon,Bang Je Geon,Jeong Yong Hwan 한국재료학회 2005 한국재료학회지 Vol.15 No.2

Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.