• Nuclear Engineering and Technology
• /
• v.19 no.1
• /
• pp.34-41
• /
• 1987

This work concerns with a comparison of the 1-dimensional (or 1-D) load follow analysis method with reference to the detailed 3-dimensional (or 3-D) computations. For this purpose a 1-D two-group finite difference code, HLOFO, and a 3-D coarse-mesh code based on the modified Borresen's method, CMSNAC, are developed. The CMSNAC code is used to obtain the 3-D power peaks and reactivity parameters in response to power swing from 100 to 50 and back to 100% in the 12-3-6-3 load cycle for the BOL of the KORI Unit 1 PWR core. The 3-D result is then compared with the 1-D HLOFO computations, the cross section and buckling inputs of which are obtained by combining the flux-volume weighting scheme with the approximate flux from the auxiliary 3-D computations. It is shown that the 1-D computation has a limited accuracy, yet it is confirmed that it can describe the core axial average behavior which is fairly consistent with the detailed 3-D computation.

• Journal of Advanced Navigation Technology
• /
• v.23 no.6
• /
• pp.554-560
• /
• 2019

Excore Nuclear Flux Monitoring System in Nuclear Power Plant monitors continuous reactor power up to maximum 200%. The monitoring method, however, has to be different depending on the reactor power level. Because the logarithmic pulse signals must be counted and processed exactly due to large uncertainty if their levels are low, on the other hand, they must be processed through statistical methodolgies if theirs are high to get exact monitoring values, in point of thermal degradation view. Therefore, we developed thermal degradation input signal selection circuit to transfer low level reactor power monitoring circuit to high level reactor power circuit at rated value in this paper. We proved their validities through testing them using real data used in nuclear power plant and analyzed their results. And, These methods will be used to measure the neutron level of excore nuclear flux monitoring system in nuclear power plant.

• Journal of the Korean Vacuum Society
• /
• v.14 no.3
• /
• pp.132-137
• /
• 2005

Neutron reflectometer has been installed at HANARO, research reactor in Korea. It has vertical sample geometry and the wavelength of incident neutron beam is $2.459\;\AA$ Neutron fluxes at monochromator and sample position were $4.5\times10^9\;n/cm^2/sec,\;6.64\times10^6\;n/cm^2/sec4 those were measured by gold wire activation method. Also, some reference thin films such as d-PS, $SiO_2$ were measured and analyzedwith HANARO neutron reflectometer. As result of the work, it was certified that minimum reflectivity and available Q range were $10^{-6},\;and\;0.003\sim0.3\;\AA^{-1}$ respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.2
• /
• pp.153-164
• /
• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

• Nuclear Engineering and Technology
• /
• v.22 no.4
• /
• pp.359-370
• /
• 1990

This study concerns with comparing low-dimensional reactor kinetics methods with a three-dimensional kinetics method to be used for safety analysis of light water reactors in order to suggest means of preparing input parameters required for low-dimensional methods. For this purpose a one-dimensional finite difference two-group diffusion theory code ODTRAN and a third-order Hermit polynomial-based point kinetics code POTRAN are developed and used to obtain low-dimensional solutions to the LRA-BWR kinetics benchmark problem. The results are compared with a three-dimensional modified Borresen's coarse-mesh solution of the kinetics problem by CMSNACK code. Through this comparison some simple but practical means of preparing input parameters of low-dimensional kinetics analysis methods are suggested.

• Nuclear Engineering and Technology
• /
• v.16 no.4
• /
• pp.181-194
• /
• 1984

The in-core fuel management problem was studied by use of the calculus of variations. Two functions of interest to a public power utility, the profit function and the cost function, were subjected to the constraints of criticality, the reactor turnup equations and an inequality constraint on the maximum allowable power density. The variational solution of the initial profit rate demonstrated that there are two distinct regions of the reactor, a constant power region and a minimum inventory or flat thermal flux region. The transition point between these regions is dependent on the relative importance of the profit for generating power and the interest charges for the fuel. The fuel cycle cost function was then used to optimize a three equal volume region reactor with a constant fuel enrichment. The inequality constraint on the maximum allowable power density requires that the inequality become an equality constraint at some points in the reactor. and at all times throughout the core cycle. The finite difference equations for reactor criticality and fuel burnup in conjunction with the equality constraint on power density were solved, and the method of gradients was used to locate an optimum enrichment. The results of this calculation showed that standard non-linear optimization techniques can be used to optimize a reactor when the inequality constraints are properly applied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.1
• /
• pp.121-126
• /
• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.

• Journal of KSNVE
• /
• v.5 no.3
• /
• pp.361-371
• /
• 1995

The vibration of reactor internals should be monitored and diagnosed for the early detection of the failure of reactor pressure vessel. This can be performed by analyzing the time-history signals from the excore neutron flux detertors. The conventional method is an on-demand system which generates power spectra through Fast Fourier Transform(FFT) algorithm. The operator can make his own decision to detect abnormal vibration using these spectra. This post- processing method, however, requires special expertise in the reactor noise analysis and signal processing for random data. It may mislead the operator into erroneous decision-making, if he is a novice in reactor noise analysis. Hence this study is focused on the automated monitoring and diagnosis procedure for the reactor noise analysis, especially on the Fuzzy algorithm to recognize the pattern of the vibration of Core Suport Barrel. The excore neutron signals of Yonggwang Nuclear Power Plant unit 3 is acquired and analyzed using conventional FFT spectra and tested to adopt the Fuzzy method. An Automated Monitoring and Diagnosis System for CSB Vibration using this Fuzzy method is proposed. Furthermore, vibration data for CSB of Youggwang Nnclear Power Plant unit 3 is presented.

• Nuclear Engineering and Technology
• /
• v.3 no.3
• /
• pp.148-154
• /
• 1971

Some formulae have been derived for the handy calculation of the formation of radioisotope when a reactor is operated regularly on a usual on-off pattern. In particular, the case of isotope production with the present operation condition of tile Korean reactor, which is in operation for 8.2 hours from Monday to Thursday and is not operated on friday and Sunday but is back in operation on Saturday only for 3.2 hours, is discussed herein with special emphasis. Should there be no secondary nuclear reaction resulting in the transformation of produced nuclide, the formula for the calculation of its activity could be derived as follows: (equation omitted) where A: activity (dps), $\Phi$： neutron flux (n cm$^{-2}$ sec$^{-1}$ ), No : number of atoms before the irradiation, $\sigma$ : activation cross section ($\textrm{cm}^2$), λ : disintegration constant of the radioactiveisotope formed (hr$^{-1}$ ), t : elapsed time of target in the reactor (hr), n : number of elasped days of target in the reactor, m : number of days from the first day of sample irradiation to Friday, s, r, q: number of weekday of Friday, Saturday and Sunday, respectively. Since the above formula consists of many invariables on the whole, the activity of each radioisotope to be produced can he easily and conveniently made available from the chart in advance which is made of the invariable terms calculated.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 2008.04a
• /
• pp.69-75
• /
• 2008

기포반응도는 월성발전소를 비롯한 CANDU형 원자로의 주된 안전성 쟁점사안으로 끊임없이 논의되어 왔다. 이는 설계기준사고가 노심에서 열에너지 불균형이 원인이 되어 기준이상의 핵연료 파손과 방사성물질 누출로 발전할 위험이 있는 사건들로 정의될 때, 사건 진행 과정에 기포반응도 증가는 조기에 운전중단을 실패할 경우 출력폭주로 이어지므로 사건의 결말이 중대사고로 전환될 위험이 크기 때문이다. 본 연구는 공개된 최신 핵자료인 ENDF/B-VII.0를 NJOY.99로 처리한 연속에너지 반응단면적 라이브러리를 구축하고 MCNP-4C에 접속하여 37봉 천연우라늄 핵연료다발의 표준노심격자에 대한 기포반응도를 시뮬레이션하여, 지금까지 각종문헌에 제시된 값들과 비교, 종합하므로 내제된 불확실성을 추정하는 내용이다. ENDF/B-VII.0 기반 MCNP-4C의 CANDU 노심격자 모델은 동일한 핵자료와 핵종농도를 사용한 WIMS-IAEA 모델과 비교할 때, 초기 노심의 임계도 오차 약 3.51mk가 연소 진행에 따라 $7.5\times10^{-4}mk$/MWD/teU의 비율로 감소하는 것으로 나타났다. 또한 MCNP-4C 예측기포반응도는 초기노심에서 기포율 50% 및 100%에 대해 각각 8.38 및 15.96mk, 평형노심에서 7.68 및 14.72mk로 계산된다. 이는 월성 2, 3, 4 FSAR의 초기노심 및 평형노심에서 100% 기포상태에 대한 값, 약15.0 및 10.6mk와 비교할 때, 초기노심은 약 1.0mk 평형노심은 약4, 1mk 보수적이지만, 다른 연구결과들과는 최대오차 ${\pm}1{\sim}2mk$ 이내에서 잘 일치하는 것으로 평가되었다. 본 연구는 CANDU 노심의 기포반응도 불확실성 요인의 규명 및 영향평가를 위한 노력의 일부로서 앞으로 감속재의 붕산농도 변화, 감속재 및 냉각재의 중수 순도 변화, 기기노화에 의한 격자 구조 및 물성 변화, 중성자속 및 출력 분포 불균형, 반응도조절장치의 위치, 등 주요 설계변수의 변화에 대한 반응도영향 분석연구를 계속할 계획이다.