[en] TID-14844 source term, which has been used as the licensing basis for siting dose evaluation since 1962, was completely replaced by the physically-based NUREG-1465 source term. In contrast to the TID-14844, NUREG-1465 specifies a categorized release in terms of phenomenological accidental phase and also defines that the dominant form of the fission product iodines will be as finely divided airborne particulate (aerosol). With these dramatic changes, the application of NUREG-1465 source term to the KNGR design transfers the focus to the removal of radioactive aerosols from the containment atmosphere. The models in SRP 6.5.2 or ANSI/ANS-56.5 are too conservative in the estimation of the particulate iodine removal. Therefore, STARNAUA computer code, which includes more realistic aerosol removal models for containment sprays and natural processes, was used for the analysis of the aerosol removal rate in the KNGR containment. This paper presents the resulting time-dependent aerosol removal coefficients for the application of NUREG-1465 source term to the KNGR design and compares the accumulated and depleted iodine masses in the containment atmosphere with those estimated using TID-14844 source term. (author)

[en] The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

[en] Systematic procedure of developing radionuclide release scenarios was established based on FEP list and Interaction Matrix for the near-surface LILW repository. The relevant FEPs were screened by experts' review in terms of domestic situation and combined into scenarios on the basis of Interaction Matrix analysis. A computer program named IMFEP_NS was developed to view and select project FEPs, to make its Interaction Matrix at user's disposal, and to visualize the interaction between FEPs and Interaction Matrix. The concept of approach to generate scenarios for entire domain is to divide the whole system domain into three sections: Near-field, Far-field, and Biosphere. Possible sub-scenarios were generated within each sectional subscenario set composed by assembling relevant FEPS and Interaction Matrix in advance, and then scenarios for entire system were built up with sub-scenarios of each section. As an application of established scenario generation approach, sixteen design scenarios and two alternative scenarios for near-surface repository were evaluated. Finally, a reference scenario and other noteworthy scenarios were selected through experts' scenario screening

[en] For the disposal of high-level radioactive waste, it is required to develop a disposal concept. The geological disposal research group in KAERI had developed a reference disposal concept and now is studying for developing a Korean disposal concept. In order to develop a Korean disposal concept, the validation of disposal concept and performance safety analysis are essential. For the validation of disposal concept, it is necessary to construct an underground research laboratory. For doing that, the research team for the validation of disposal concept studies for constructing an underground research facility in KAERI. In order to design the underground research facility with computer simulation, disposal concept, rock characteristics, and topology should be considered. In this study, the reference disposal concept, which is necessary to be considered in the design of underground research tunnel, will be introduced first. After then, the important factors related to the underground research tunnel will be discussed. In the case of the site, where the underground tunnel are expected to be located, the surface topology is varying with thick weathered zone. In order to make the research modules in deep location with limited tunnel excavation, it is recommended to excavate a declined access tunnel. This study is for investigating the influence of different geological conditions, tunnel slope, tunnel size, and sequential excavation. In this study, mechanical stability analysis for the conceptual design of the underground research tunnel for the validation of Korean disposal concept had been carried out. Investigation of the influence of important parameters on mechanical stability was performed. The results from this study will be utilized for the geological investigation at the site, where an underground research tunnel will be located, as well as the design of the tunnel. Important conclusions from the study are as followings: (1) If the underground research tunnel is located at KAERI site, it is not expected to develop any plastic zone around the tunnel, mainly due to the shallow depth. The maximum stress will be compressive stress and might be as high as 5MPa. The exact location, where the maximum compressive stress will be developed, will be about 20m to the dead end of the tunnel. There is no difference on stress and displacement distributions between the case with simultaneous excavation and the case with sequential excavation. (2) It is expected to have stress release in the roof and floor after the excavation of the tunnel. Tensile stress would be developed at some areas in the floor. (3) There is no influence of weathered zone size on the stress and displacement distributions around the tunnel. It is, however, recommended to carry out further analysis with actually measured properties of the weathered zone for more accurate decision. (4) With increase of tunnel size, the stress around the tunnel was increased. In the case of von-Mises stress, the magnitude increase due to 1m increase of tunnel size was found to be only about 4%

[en] Several types of gamma-ray sensors were constructed by packing different numbers of fibers into aluminum tubes, and tested to detect the ¹³⁷Cs gamma ray. It has been found from this investigation that tapered fibers can be more efficient to collect the lights produced inside the sensor and transfer them into the transmitting fiber. In order to predict the functioning of the tapered fiber, tapered plastic scintillators, composed of polystyrene with minute amount of dPOPOP and PPO or dPBD, were fabricated and tested for the detection of gamma rays from 1.0, 1.5, 3.0. 5.0 μCi ¹³⁷Cs sources, and the pulse height spectrum and the relationship between the radioactivity and the total counts are analyzed. It has been found from this study that the tapered scintillating optical fiber, if manufactured, can be practically applied to the development of gamma-ray sensors which can be deployed in μCi-level radiation fields

[en] Scintillating optical fibers have several advantages over other conventional materials used for radiation detection. We have used glass and plastic scintillating fibers to detect gamma rays emitted from ⁶⁰Co and ¹³⁷Cs, and beta rays from ⁹⁰Sr. The sensors are constructed of single strand or multi-strand fibers of 1 mm diameter. The glass scintillating fiber used contains cerium-activated lithium-silicate as scintillating material and the plastic scintillating fiber used is Bicron model BCF-12. In this paper, we report the pulse-height spectra obtained by both sensor types, and analyze them in the aspect of their usability for radiation detectors. Our investigation suggests that the glass fiber can be used to develop gamma ray detectors which will function in high and low gamma ray flux environments. Use of the sensor for the beta ray detection was not satisfactory. The plastic fiber sensor did not work satisfactorily for the weak gamma sources, but did produce somewhat promising results. The scintillating plastic fiber offers some feasibility as beta ray sensor material

[en] Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distributions in fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors

[en] CONTAIN 2.0 code has been employed to simulate the KAEVER Test series K123A, K148A, K186A, K188A that were proposed as International Standard Problem-44 by OECD/CSNI, and the accuracy of the calculation results has been analyzed. All of these tests were conducted to investigate the behavior of the aerosol depletion with steam condensing on the particle surface under highly saturated steam conditions. The code predicts considerably slower aerosol depletion than the experiment for the CsOH aerosol which is highly hygroscopic, and also showed very similar results for the CsI aerosol which is moderately hygroscopic. For the Ag aerosol which is non-hygroscopic, however, the code predicts much faster depletion compared with the experimental data. For the mixed aerosol of hygroscopic CsOH and non-hygroscopic Ag, the calculation results show the same depletion pattern for both components, which is different from usual anticipation, and do show remarkably slower depletion compared with the experimental data

[en] Systematic development procedure of radionuclide release scenarios was established based on FEP list and Interaction Matrix for near-surface LILW repository. FEPs were screened by experts' review in terms of domestic situation and combined into scenarios on the basis of Interaction Matrix analysis. Under the assumption of design scenario, system domain was divided into three sections. Near-field, Far-field and Biosphere. Subscenarios for each section were developed, and then scenarios for entire system were composed of sub-scenarios from each section. Finally, 6 reference scenarios for near-surface repository were selected through scenario screening

[en] The influence of the gas phase turbulent stress to the characteristic speeds (i.e., eigenvalues) of two-phase system in CANDU reactor channel has been theoretically investigated. The average two-fluid model has been constituted with closure relations for horizontally stratified flow. A vapor phase turbulent stress model for the regular interface geometry has been included. It is found that the second order waves (i.e., eigenvalues) propagate in opposite direction with almost the same speed in the fixed frame of reference. Using the well-posedness limit of the two-phase system, the dispersed-stratified flow regime boundary has been modeled. Two-phase Froude number has been found to be a convenient parameter in quantifying the flow regime boundary as a function of the void fraction. The influence of the vapor phase turbulent stress found to stabilize the flow stratification. This study clearly shows that the average two-fluid model can be used for a mechanistic determination of horizontal flow regimes in CANDU reactor channel if appropriate closure relations are developed. (author) 11 refs., 3 figs