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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 42, Issue 6 - Dec 2010
Volume 42, Issue 5 - Oct 2010
Volume 42, Issue 4 - Aug 2010
Volume 42, Issue 3 - Jun 2010
Volume 42, Issue 2 - Apr 2010
Volume 42, Issue 1 - Feb 2010
Selecting the target year
MANAGING SPENT NUCLEAR FUEL FROM NONPROLIFERATION, SECURITY AND ENVIRONMENTAL PERSPECTIVES
Choi, Jor-Shan ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 231~236
DOI : 10.5516/NET.2010.42.3.231
The growth in global energy demand and the increased recognition of the impacts of carbon dioxide emissions from fossil fuel plants have aroused a renewed interest on nuclear energy. Many countries are looking afresh at building more nuclear power stations to deal with the twin problems of global warming and the need for more generating capacity. Many in the nuclear community are also anticipating a significant growth of new nuclear generation in the coming decades. If there is a nuclear renaissance, will the expansion of nuclear power be compatible with global non-proliferation and security? or will it add to the environmental burden from the large inventory of spent nuclear fuel already produced in existing nuclear power reactors? We learn from past peaceful nuclear activities that significant concerns associated with nuclear proliferation and spent-fuel management have resulted in a decrease in public acceptance for nuclear power in many countries. The terrorist attack in the United States (US) on September 11, 2001 also raised concern for security and worry that nuclear materials may fall into the wrong hands. As we increase the use of nuclear power, we must simultaneously reduce the proliferation, security and environmental risks in managing spent-fuel below where they are today.
REVIEW AND FUTURE ISSUES ON SPENT NUCLEAR FUEL STORAGE
Saegusa, T. ; Shirai, K. ; Arai, T. ; Tani, J. ; Takeda, H. ; Wataru, M. ; Sasahara, A. ; Winston, P.L. ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 237~248
DOI : 10.5516/NET.2010.42.3.237
The safety of metal cask and concrete cask storage technology has been verified by CRIEPI through several research programs on demonstrative testing for the interim storage of spent fuel. The results have been reflected in the safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy, Trade and Industry) of the Japanese government. On top of that, spent fuel integrity has been studied by the Japan Nuclear Energy Safety Organization (JNES). This paper reviews these research programs. Future issues include the long-term integrity of cask components and high burn-up spent fuel.
THE EFFECTS OF CREEP AND HYDRIDE ON SPENT FUEL INTEGRITY DURING INTERIM DRY STORAGE
Kim, Hyun-Gil ; Jeong, Yong-Hwan ; Kim, Kyu-Tae ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 249~258
DOI : 10.5516/NET.2010.42.3.249
Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at
and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.
DEVELOPMENT AND VALIDATION OF COUPLED DYNAMICS CODE 'TRIKIN' FOR VVER REACTORS
Obaidurrahman, K. ; Doshi, J.B. ; Jain, R.P. ; Jagannathan, V. ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 259~270
DOI : 10.5516/NET.2010.42.3.259
New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.
MONTE CARLO DEPLETION UNDER LEAKAGE-CORRECTED CRITICAL SPECTRUM VIA ALBEDO SEARCH
Yun, Sung-Hwan ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 271~278
DOI : 10.5516/NET.2010.42.3.271
While the deterministic lattice physics/depletion codes use leakage-corrected critical spectrum (although approximate due to the B1 buckling search employed), Monte Carlo depletion codes currently in use do not have such a feature in spite of their heterogeneity and continuous-energy modeling capability. This paper describes an approach to Monte Carlo depletion with leakage-corrected critical spectrum derived from first principles. This is based on the concept of albedo eigenvalue treated as weight of the reflected neutron in Monte Carlo simulation.
DEVELOPMENT AND PRELIMINARY ASSESSMENT OF A THREE-DIMENSIONAL THERMAL HYDRAULICS CODE, CUPID
Jeong, Jae-Jun ; Yoon, Han-Young ; Park, Ik-Kyu ; Cho, Hyoung-Kyu ; Lee, Hee-Dong ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 279~296
DOI : 10.5516/NET.2010.42.3.279
For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.
ENHANCEMENT OF DRYOUT HEAT FLUX IN A DEBRIS BED BY FORCED COOLANT FLOW FROM BELOW
Bang, Kwang-Hyun ; Kim, Jong-Myung ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 297~304
DOI : 10.5516/NET.2010.42.3.297
In the design of advanced light water reactors (ALWRs) and in the safety assessment of currently operating nuclear power plants, it is necessary to evaluate the possibility of experiencing a degraded core accident and to develop innovative safety technologies in order to assure long-term debris cooling. The objective of this experimental study is to investigate the enhancement factors of dryout heat flux in debris beds by coolant injection from below. The experimental facility consists mainly of an induction heater, a double-wall quartz-tube test section containing a steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of the particle bed was achieved in the radial direction and the axial variation was within 20%. This paper reports the experimental data for 3.2 mm and 4.8 mm particle beds with a 300 mm bed height. The dryout heat density data were obtained for both the top-flooding and the forced coolant injection from below with an injection mass flux of up to
. The dryout heat density increased as the rate of coolant injection increased. At a coolant injection mass flux of
, the dryout heat density was
for the 4.8 mm particle bed and
for the 3.2 mm particle bed. The enhancement factors of the dryout heat density were 1.6-1.8.
PROTECTION SEQUENCE OF AC/DC CONVERTERS FOR ITER PF MAGNET COILS
Oh, Byung-Hoon ; Hwang, Churl-Kew ; Lee, Kwang-Wang ; Jin, Jeong-Tae ; Chang, Dae-Sik ; Oh, Jong-Seok ; Choi, Jung-Wan ; Suh, Jae-Hak ; Tao, Jun ; Song, In-Ho ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 305~312
DOI : 10.5516/NET.2010.42.3.305
The protection sequence of an AC/DC converter for an ITER PF coil system is developed in this study. Possible faults in the coil system are simulated and the results reflected in the design of a sequence to protect the coil and converter. The inductances of the current sharing reactors and thyristor numbers in parallel with the bridge arms are optimized with the designed protection sequence.
APPLICATION OF ALANINE/ESR SPECTRUM SHAPE CHANGE IN GAMMA DOSIMETRY
Choi, Hoon ; Kim, Jeong-In ; Lee, Byung-Ill ; Lim, Young-Ki ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 313~318
DOI : 10.5516/NET.2010.42.3.313
Alnine pellets were installed in a nuclear power plant for one or two operation cycles and measured by electron spin resonance (ESR) spectrometers for dosimetry. Dose and "x/y ratio", i.e., satellite peak over main center peak ratio, were measured for the returned alanine dosimeters from the nuclear power plant and compared to the values of reference alanine dosimeters exposed only to gamma rays. The variation of the x/y ratio change depended on the population of radicals from each radiation component with different LET. The gamma dose in a mixed radiation field was estimated by an additive gamma ray irradiation experiment and the measured dose rate at specified locations in the containment building.
A SENSITIVITY ANALYSIS OF THE KEY PARAMETERS FOR THE PREDICTION OF THE PRESTRESS FORCE ON BONDED TENDONS
Jang, Jung-Bum ; Lee, Hong-Pyo ; Hwang, Kyeong-Min ; Song, Young-Chul ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 319~328
DOI : 10.5516/NET.2010.42.3.319
Bonded tendons have been used in reactor buildings at some operating nuclear power plants in Korea. Assessing prestress force on these bonded tendons has become an important pending problem in efforts to assure continued operation beyond their design life. The System Identification (SI) technique was thus developed to improve upon the existing indirect assessment technique for bonded tendons. As a first step, this study analyzed the sensitivity of the key parameters to prestress force, and then determined the optimal parameters for the SI technique. A total of six scaled post-tensioned concrete beams with bonded tendons were manufactured. In order to investigate the correlation of the natural frequency and the displacement to prestress force, an impact test, a Single Input Multiple Output (SIMO) sine sweep test, and a bending test using an optical fiber sensor and compact displacement transducer were carried out. These tests found that both the natural frequency and the displacement show a good correlation with prestress force and that both parameters are available for the SI technique to predict prestress force. However, displacements by the optical fiber sensor and compact displacement transducer were shown to be more sensitive than the natural frequency to prestress force. Such displacements are more useful than the natural frequency as an input parameter for the SI technique.
DEVELOPMENT OF REACTOR POWER CONTROL LOGIC FOR THE POWER MANEUVERING OF KALIMER-600
Seong, Seung-Hwan ; Kang, Han-Ok ; Kim, Seong-O ;
Nuclear Engineering and Technology, volume 42, issue 3, 2010, Pages 329~338
DOI : 10.5516/NET.2010.42.3.329
We developed an achievable control logic for the reactor power level during a power maneuvering event and set up some constraints for the control of the reactor power in a conceptual sodium-cooled fast reactor (KALIMER-600) that was developed at KAERI. For simulating the dynamic behaviors of the plant, we developed a fast-running performance analysis code. Through various simulations of the power maneuvering event, we evaluated some suggested control logic for the reactor power and found an achievable control logic. The objective of the control logic is to search for the position of the control rods that would keep the average temperature of the primary pool constant and, concurrently, minimize the power deviation between the reactor and the BOP cycle during the power maneuvering. In addition, the flow rates of the primary pool and the intermediate loop should be changed according to the power level in order to not violate the constraints set up in this study. Also, we evaluated some movement speeds of the control rods and found that a fast movement of the control rods might cause the power to fluctuate during the power maneuvering event. We suggested a reasonable movement speed of the control rods for the developed control logic.