• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.5
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• pp.451-457
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• 2010

In the present study, the influences of grains and precipitates of microstructural evolution on the ultrasonic nonlinearity have been experimentally investigated. The prior-austenite grain and precipitate size are controlled by the variation in austenitizing and tempering conditions in reactor pressure vessel materials of nuclear power plant, SA508 Gr.3 low alloys. The ultrasonic nonlinearity was found to have strong correlations with grains and precipitates since the ultrasonic nonlinear parameter $\beta$ shows decrease trend with coarsening of grains and precipitates. Although the prior-austenite grain size increased, the $\beta$ changed little due to the effects of subgrains, packets and laths. For the preciptate effects, the $\beta$ decreased sharply due to decrease in $Mo_2C$ causing the coherency stain in addition to the precipitate size. The results in this study may provide a potential for characterizing the microstructural evolution, grains and precipitates, by measuring the ultrasonic nonlinearity.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.274-284
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• 1992

A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.

• Journal of the Korean Magnetics Society
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• v.8 no.5
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• pp.249-254
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• 1998

In relation to the application of magnetic method to the evaluation of irradiation damage (embrittlement) changes in the magnetic parameters(hysteresis loop and Barkhausen noise) and Vickers microhardness due to neutron irradiation and heat treatment were measured and compared. In the case of irradiation $(2.3{\times}10^{19}\;n/cm^2,\; E{\ge}1\;Mev,\; 288{\circ}C)$ hysteresis loop measurements show that susceptibility decreases as coercivity increase. Saturation magnetization do not show any change. Barkhausen noise amplitude and Barkhausen noise energy have decreased while Vickers microhardness has increased. For isothermally heat treated condition of irradiated specimen at 470 $^{\circ}C$ and 540 $^{\circ}C$, Barkhausen noise energy has increased while Vickers microhardness has decreased. Results of BNE and Vickers microhardness are reversed to the results on irradiated condition. All these consistent changes in magnetic parameter and Vickers microhardness measurement, which are thought to be resulted from the interaction between irradiation-induced defects and dislocation, and magnetic domain, respectively, show a possibility that magnetic measurement may be used to the evaluation of material degradation and recovery due to neutron irradiation and heat treatment, respectively, if a relevant large database in prepared.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.56-65
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• 1991

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.

• Korean Journal of Metals and Materials
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• v.48 no.3
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• pp.194-202
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• 2010

Application of a stronger and more durable material for reactor pressure vessels (RPVs) might be an effective way to insure the integrity and increase the efficiency of nuclear power plants. A series of research projects to apply the SA508 Gr.4 steel in ASME code to RPVs are in progress because of its excellent strength and durability compared to commercial RPV steel (SA508 Gr.3 steel). In this study, the microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure that has coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as $M_{23}C_6$ and $M_7C_3$ due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. In addition, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect, and the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.337-350
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• 1990

A post-irradiation annealing study was conducted with use of reactor pressure vessel(RPV) steel A533B Cl.1 base metal irradiated to a dose of 4.84$\times$10$^{18}$ n/$\textrm{cm}^2$ at about 38$0^{\circ}C$. Microhardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurs in the temperature range of 280-3O5$^{\circ}C$, Michrohardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurrs in the temperature range of 280-305$^{\circ}C$. The variations of Ip, Iw and R parameters indicated that the formation of vacancy clusters by vacancy agglomeration and the annihilation of monovacancies are the first recovery process. The second recovery process occurs in the range of 405-49$0^{\circ}C$ and positron annihilation parameters measured indicated that the dissolution of carbon atoms decorated around vacancy-type defects and possible precipitates, and the annihilation of monovacancies give rise to the second recovery process. It was further indicated that radiation anneal hardening (RAH) in the range of 305-405$^{\circ}C$ between the temperature ranges for the two processes occurs due to the formation of carbon-decorated vacancy clusters and precipitates. The activation energies, orders of reaction and other characteristics of recovery processes were determined by the Meechan-Brinkman method. The activation energy for the first recovery process was determined as 1.76 eV and that for the second recovery process as 2.00eV. These values are lower than those obtained by other workers. This difference may be attributed to the lower copper content of the RPV steel used in the present study. The order of reaction for the first recovery process was determined as 1.78, while that for the second recovery process as 1.67 Non-integer orders of reaction for recovery processes seem to be attributed to the fact that several mechanisms for the first order and the second order of reaction are compounded in one process. This result also supports for the above conclusions from measurements of PA parameters.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.127-134
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• 1980

Shielding problems associated with neutron streaming through the reactor vessel cavity of pressurized water reactors are discussed to a certain extent with the actual examples in the currently operating reactors. Various remedial techniques are proposed herein to mitigate the tedious neutron streaming phenomena including piling up in heaps of temporary boron-containing bags and the installation of permanent shield structure making use of a certain refractory materials. In conclusion, optimum cavity shielding design concepts are presented with special emphasis on such major factors as the identification of major neutron streaming path, selection of necessary shielding materials with acceptable constraints, detailed design characteristics and physical configuration as well as the formulation of dependable mathematical tools to predict the final outcome of each design concept proposed in the context.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.5
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• pp.554-562
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• 2004

An empirical relationship between parameters from SP curves and tensile properties has been systematically investigated by experimental tests and FEM simulations. A series of SP and tensile tests were performed. SP tests were also simulated by FE analysis with various tensile properties. It was found that the yield loads(Py) and the maximum loads( $P_{MAX}$) in SP curves were linearly related with the yield strength($\sigma$$_{o}$) and the tensile strength($\sigma$$_{UTS}$), respectively. The yield loads defined from the intersection point of two lines tangent to the elastic bending region and plastic bending region showed better relation to the yield strength than those from offset line. The maximum loads in SP curves showing plastic instability region was linearly related with the tensile strengths. The slope of SP curves in simulation results had a close correlation with the hardening coefficient and hardening strength as well.l.l.l.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.10
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• pp.1567-1574
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• 2001

In order to operate an RPV safely it is necessary to keep the pressure-temperature (P-T) limit during the heatup and cooldown process. While the ASME Code provides the P-T limit curve for safe operation, this limit curve has been prepared under conservative assumptions In this paper the effects of conservative assumptions involved in the P-T limit curve specified in the ASME Code Sec. XI were investigated. Three different parameters the crack depth the cladding thickness and the cooling rate, were reviewed based on 3-D finite element analyses. Also the constraint effect on P-T limit curve generation was investigated based on J- T approach. It was shown that the crack depth and the constraint effect change the safe region in P-T limit curve significantly Therefore it is recommended to prepare a more precise P-T limit curve based on finite element analysis to obtain P-T limit for safe operation of an RPV.