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A new method to calculate a standard set of finite cloud dose correction factors for the level 3 probabilistic safety assessment of nuclear power plants

  • Gee Man Lee;Woo Sik Jung
    • Nuclear Engineering and Technology
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    • v.56 no.4
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    • pp.1225-1233
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    • 2024
  • Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations. To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes. This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.

Evaluation of Neutron Flux Accounting for Shadowing Effect Among the Dry Storage Casks (경수로 사용후핵연료 건식저장용기 간 중성자 표면선속 간섭률 평가)

  • Min Woo Kwak;Shin Dong Lee;Kwang Pyo Kim
    • Journal of Radiation Industry
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    • v.18 no.2
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    • pp.133-140
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    • 2024
  • The Korean 2nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

Glass Dissolution Rates From MCC-1 and Flow-Through Tests

  • Jeong, Seung-Young
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2004.06a
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    • pp.257-258
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    • 2004
  • The dose from radionuclides released from high-level radioactive waste (HLW) glasses as they corrode must be taken into account when assessing the performance of a disposal system. In the performance assessment (PA) calculations conducted for the proposed Yucca Mountain, Nevada, disposal system, the release of radionuclides is conservatively assumed to occur at the same rate the glass matrix dissolves. A simple model was developed to calculate the glass dissolution rate of HLW glasses in these PA calculations [1]. For the PA calculations that were conducted for Site Recommendation, it was necessary to identify ranges of parameter values that bounded the dissolution rates of the wide range of HLW glass compositions that will be disposed. The values and ranges of the model parameters for the pH and temperature dependencies were extracted from the results of SPFT, static leach tests, and Soxhlet tests available in the literature. Static leach tests were conducted with a range of glass compositions to measure values for the glass composition parameter. The glass dissolution rate depends on temperature, pH, and the compositions of the glass and solution, The dissolution rate is calculated using Eq. 1: $rate{\;}={\;}k_{o}10^{(ph){\eta})}{\cdot}e^{(-Ea/RT)}{\cdot}(1-Q/K){\;}+{\;}k_{long}$ where $k_{0},\;{\eta}$ and Eaare the parameters for glass composition, pH, $\eta$ and temperature dependence, respectively, and R is the gas constant. The term (1-Q/K) is the affinity term, where Q is the ion activity product of the solution and K is the pseudo-equilibrium constant for the glass. Values of the parameters $k_{0},\;{\eta}\;and\;E_{a}$ are the parameters for glass composition, pH, and temperature dependence, respectively, and R is the gas constant. The term (1-Q/C) is the affinity term, where Q is the ion activity product of the solution and K is the pseudo-equilibrium constant for the glass. Values of the parameters $k_0$, and Ea are determined under test conditions where the value of Q is maintained near zero, so that the value of the affinity term remains near 1. The dissolution rate under conditions in which the value of the affinity term is near 1 is referred to as the forward rate. This is the highest dissolution rate that can occur at a particular pH and temperature. The value of the parameter K is determined from experiments in which the value of the ion activity product approaches the value of K. This results in a decrease in the value of the affinity term and the dissolution rate. The highly dilute solutions required to measure the forward rate and extract values for $k_0$, $\eta$, and Ea can be maintained by conducting dynamic tests in which the test solution is removed from the reaction cell and replaced with fresh solution. In the single-pass flow-through (PFT) test method, this is done by continuously pumping the test solution through the reaction cell. Alternatively, static tests can be conducted with sufficient solution volume that the solution concentrations of dissolved glass components do not increase significantly during the test. Both the SPFT and static tests can ve conducted for a wide range of pH values and temperatures. Both static and SPFt tests have short-comings. the SPFT test requires analysis of several solutions (typically 6-10) at each of several flow rates to determine the glass dissolution rate at each pH and temperature. As will be shown, the rate measured in an SPFt test depends on the solution flow rate. The solutions in static tests will eventually become concentrated enough to affect the dissolution rate. In both the SPFt and static test methods. a compromise is required between the need to minimize the effects of dissolved components on the dissolution rate and the need to attain solution concentrations that are high enough to analyze. In the paper, we compare the results of static leach tests and SPFT tests conducted with simple 5-component glass to confirm the equivalence of SPFT tests and static tests conducted with pH buffer solutions. Tests were conducted over the range pH values that are most relevant for waste glass disssolution in a disposal system. The glass and temperature used in the tests were selected to allow direct comparison with SPFT tests conducted previously. The ability to measure parameter values with more than one test method and an understanding of how the rate measured in each test is affected by various test parameters provides added confidence to the measured values. The dissolution rate of a simple 5-component glass was measured at pH values of 6.2, 8.3, and 9.6 and $70^{\circ}C$ using static tests and single-pass flow-through (SPFT) tests. Similar rates were measured with the two methods. However, the measured rates are about 10X higher than the rates measured previously for a glass having the same composition using an SPFT test method. Differences are attributed to effects of the solution flow rate on the glass dissolution reate and how the specific surface area of crushed glass is estimated. This comparison indicates the need to standardize the SPFT test procedure.

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Evaluation of the Size of Emergency Planning Zone for the Korean Standard Nuclear Power Plants (한국표준형 원전에 대한 방사선비상계획구역 범위 평가)

  • Jeon, In-Young;Lee, Jai-Ki
    • Journal of Radiation Protection and Research
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    • v.28 no.3
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    • pp.215-223
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    • 2003
  • Against major release of radioactive material in nuclear power plant, Emergency Planning Zone(EPZ)s are typically established around nuclear power plants to effectively perform the public protective measures. The domestic methodology to determine the size of the EPZ is similar to that of Japan established in 1980, where calculations were based on the conservative accident source term. The objective of this study is to re-evaluate the validity of established EPZ, the area within the radius of $8{\sim}10km$ around domestic nuclear power plants, using the source terms covering full spectrum of accidents obtained from PSA study of ULJIN 3&4. To evaluate the risks of health effects, the computer code MACCS2(MELCOR Accident Consequence Code System2) was used. The result shows that the existing EPZ can reduce the probability of early fatality adequately for most of the source term categories(STCs) except for STC-14 and STC-19. In case of STC-14 and 19, the evacuation distance of 16km and 13km, respectively, are required. These distances can be reduced by improving emergency preparedness since the sensitivity studies for the public protective actions show that the magnitude of early fatality is largely affected by the time delays in notification and evacuation.

Measurement of the Gap and Grain Boundary Inventories of Cs, Sr in and I in Domestic Used PWR Fuels (국내 PWR 사용후핵연료에서 세슘, 스트론튬과 요오드의 갭 및 입계 재고량 측정)

  • Kim, S.S.;Kang, K.C.;Choi, J.W.;Seo, H.S.;Kwon, S.H.;Cho, W.J.
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.5 no.1
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    • pp.79-84
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    • 2007
  • Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

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A Revisit to the Recent Human Error Events in Nuclear Power Plants Focused to the Organizational and Safety Culture

  • Lee, Yong-Hee
    • Journal of the Ergonomics Society of Korea
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    • v.32 no.1
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    • pp.117-124
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    • 2013
  • Objective: This paper presents additional considerations related to organization and safety culture extracted from recent human error incidents in Korea, such as station blackout(i.e., SBO) in Kori#1. Background: Safety culture has been already highlighted as a major cause of human errors after 1986 Chernobyl accident. After Fukushima accident in Japan, the public acceptance for nuclear energy has taken its toll. Organizational characteristics and culture became elucidated as a major contributor again. Therefore many nuclear countries are re-evaluating their safety culture, and discussing any preparedness and its improvement. On top of that, there was an SBO in 2012 in the Kori#1. Korean public feels frustrated due to the similar human errors causing to a catastrophe like Fukushima accident. Method: This paper reassesses Japan's incidents, and revisits Korea's recent incidents. It focuses on the analysis of the hazards rather than the causes of human errors, the derivation of countermeasures, and their implementation. The preceding incidents and conclusions from Japanese experience are also re-analyzed. The Fukushima accident was an SBO due to the natural disaster such as earthquakes and a successive tsunami. Unlike the Fukushima accident, the Kori#1 incident itself was simple and restored without any loss and radioactive release. However, the fact that the incident was deliberately concealed led to massive distrust. Moreover, the continued violation of rules and organized concealment of the accident are serious signs of a new distorted type of human errors, blatantly revealing the cultural and fundamental weakness of the current organization. Result: We should learn from Japanese experiences who had taken pride in its safety technology and fairly high confidence in safety culture. Japan's first criticality accident in JCO facility splashed cold water on that confidence. It has turned out to be a typical case revealing the problems in the organization and safety culture. Since Japan has failed to gain lessons and countermeasure, the issue persists to the Fukushima incident. Conclusion: Safety culture is not a specific independent element, which makes it difficult to either evaluate it properly or establish countermeasures from the lessons. It may continue to expose similar human errors such as concealment of incident and manipulation of bad data. Application: Not only will this work establish the course of research for organization and safety culture, but this work will also contribute to the revitalization of Korea's nuclear industry from the disappointment after the export contract to UAE.

Subgrouping of N1a Stage Papillary Thyroid Carcinoma with Positive Node Ratio (갑상선유두상암의 중앙림프절 전이율에 따른 N1a병기의 세분화)

  • Lee, Min-Wan;Cho, Jin-Seong;Cho, Dong-Hoon;Ryu, Young-Jae;Park, Min-Ho;Yoon, Jung-Han
    • Korean Journal of Head & Neck Oncology
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    • v.32 no.1
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    • pp.13-19
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    • 2016
  • Background : The 2015 American thyroid association (ATA) guidelines greatly expanded section on risk stratification of thyroid cancer. Definition of "Low risk of recurrence" has expanded, by inclusion of small volume lymph node involvement, such as less than 5 lymph node metastases each smaller than 2mm in central compartment. Purpose : We evaluated the number of positive nodes, Positive node ratio (PNR), recurrence, and radioablation therapy. Also, evaluated the safety of omitting strategy of radioablation after total thyroidectomy with PTC, especially on low-PNR N1a patients compared with high-PNR N1a patients. Methods : Consecutive 147 N1a and 216 N0 patients who underwent total thyroidectomy with central neck dissection between 2003 and 2004 were enrolled. We divided 147 N1a patients into two groups, such as 96 high-PNR versus 51 low-PNR group according to 50% of PNR, and compared these two groups with N0 group. Results: 7.2% (26/363) recurrences were occurred, and 21/147 (14.3 %) recurrences were on N1a patients, and 5/216 (2.3 %) were on N0 patients. Of these 21 recurrences in N1a stage patients, 20 (95.2 %) recurrences were occurred in high-PNR N1a group and only 1 (4.8 %) recurrence was in low-PNR N1a group. The recurrence of low-PNR N1a group was significantly lower than high-PNR N1a group (Log-rank p value = 0.003), but significantly not different from N0 group (Log-rank p value = 0.889). Although this study was a retrospective non-randomized trial with small number of patients, the 10-year recurrence of omitting RAI in low-PNR N1a patients with less than 50% of PNR were shown to be comparable with 216 N0 low risk patients. Conclusion : Positive node ratio could be a useful predictor of recurrence and useful guidance postoperative management -rather than absolute number of positive node.

Effect of engineered barriers on the leach rate of cesium from spent PWR fuel (가압경수로 사용후핵연료 중 세슘의 침출에 미치는 공학적 방벽 영향)

  • Chun Kwan Sik;Kim Seung-Soo;Choi Jong-Won
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.3 no.4
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    • pp.329-333
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    • 2005
  • To identify the effect of engineered barriers on the leach rate of cesium from spent PWR fuel under a synthetic granitic groundwater, the related leach tests with and without bentonite or metals have been performed up to about 6 years. The leach rates were decreased as a function of leaching time and then became a constant after a certain period. The period in a bare spent fuel was much longer than that with bentonite or metal sheets. The cumulative fraction of cesium released from the spent fuel with bentonite or with copper and stainless steel sheets was steadily increased, but the fraction from bare fuel was rapidly and then sluggishly increased. However, the values deducted its gap inventory from the cumulative fraction of cesium released from the bare fuel was almost very close to the others. These suggest that the initial release of cesium from bare fuel might be dependant on its gap inventory and the effect of engineered barriers on the long-term leach rate of cesium would be insignificant but the rate with engineered barriers could be reduced in the initial transient period due to their retardation effect. And the long-term leach rate of cesium from spent fuel in a repository would be approached to a constant rate of $2\times10^{-2}g/m^2-day$.

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Study of morphology on the Oxidation and the Annealing of High Burn-hp $UO_2$ Spent Fuel (고연소도 사용후 핵연료의 가열산화와 고온가열을 통한 미세조직 변화고찰)

  • Kim Dae Ho;Bang Jae Geun;Yang Yong Sik;Song Keun Woo;Lee Hyung Kwon;Kwon Hyung Moon
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.3 no.4
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    • pp.301-307
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    • 2005
  • The morphology of the high burnup $UO_2$ spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent $UO_2$ fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at $1400^{\circ}C$ for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to $500^{\circ}C$ and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the $UO_2$ matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to $10\;{\mu}m$.

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Study on the Code System for the Off-Site Consequences Assessment of Severe Nuclear Accident (원전 중대사고 연계 소외결말해석 전산체계에 대한 고찰)

  • Kim, Sora;Min, Byung-Il;Park, Kihyun;Yang, Byung-Mo;Suh, Kyung-Suk
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.14 no.4
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    • pp.423-434
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    • 2016
  • The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.