For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

In ROK, when designing a spent nuclear fuel (SNF) storage facility and cask, criticality safety analysis is performed assuming that the SNF is a fresh fuel in order to ensure conservatism. Storage and transportation capacity can be increased by more than 30% by applying the burnup credit, but it has not been applied to the management of SNF. On the other hand, currently in criticality safety analysis, average burnup value is applied to axial burnup profiles, and it is not conservative because burnup of the middle of SNF is greater than average value. Thus, measuring burnup of SNF with high accuracy contributes to the economics and safety of the management of SNF. In this paper, nondestructive burnup evaluation methods for SNF are reviewed in order to study how to measure burnup more accurately. Gamma ray spectrometry and neutron counting have been used as non-destructive burnup evaluation methods of SNF. Gamma spectrum analysis uses the ratio of Cs-134/Cs-137 or Eu-154/Cs-137. The ratio of Cs-134/Cs-137 is used to SNF with cooling time less than 20 years, and the ratio of Eu- 154/Cs-137 is used to SNF with cooling time more than 20 years due to their half-life. In spectrum analysis, detector sensors with high efficiency and energy resolution are needed to clarify each spectrum. High-purity germanium (HPGe) detector has high energy resolution. However, it is not suitable for the analysis of the SNF in the spent fuel pool because it requires separate cooling system and large volume. Thus, CdZnTe (CZT) detector, which has medium energy resolution, is used as a detector of gamma ray spectrometry for the analysis of the SNF in the spent fuel pool. Recently, LaBr3 detector has been commercialized. Although it is difficult to compare clearly due to different conditions such as detector volume and crystal size, LaBr3 detector showed better resolution than CZT in the entire energy region. Neutron counting method has a large error compared to gamma spectrometry because the neutron flux is lower than gamma ray, and neutron absorption reaction, induced fission, and pool environment have to be considered. Large quantity of gamma energy is deposited in the detector by the fission fragments near the SNF. Therefore, fission chambers, which have the highest insensitivity to gamma rays, must be used as neutron detector in order to avoid noise from gamma rays.

경수로 사용후핵연료 수송/저장용기의 핵임계 해석은 사용후핵연료내의 악티나이드핵종 및 핵분열생성물 함유량에 대한 불확실성을 이유로 신연료로 가정된 가상의 연료를 선정하여 평가해오고 있다. 그러나 이러한 평가방법은 용기 설계 시 과 도한 임계여유도를 유도하여 경제적 손실을 유발할 수 있는 단점이있다. 이와 같은 단점을 극복하기 위하여 최근 연소도이 득효과(burnup credit, BUC)를 반영한 수송저장용기의 설계 및 상용화를 위한 연구가 추진되었다. 이에 본 연구에서는 한국 원자력환경공단에서 개발중인 금속겸용용기를 대상으로 연소도 이득효과적용 시 핵임계 안전성(criticality safety)에 영향을 미칠 것으로 예상되는 ‘노심 운전인자’, ‘축방향 연소도 분포’, ‘오장전 사고상황’에 대하여 핵임계 평가를 수행하였다. 그 결과 노심운전인자 중 저농축, 고연소도일 때 비출력에 따른 핵임계 변화가 크게 평가되었으며, 고연소도 사용후핵연료에서 End effect가 양의 값을 나타내었다. 특히 오장전에 의한 유효증배계수는 최대 0.18467증가하였으므로, 연소도이득효과를 적용 할 경우 필수고려사항임을 확인하였다. 본 연구결과는 국내모델(금속겸용용기)의 연소도 이득효과 적용기술 개발 및 사용 후핵연료 장전 시 일어날 수 있는 오장전 사고를 방지하기 위한 운영절차 개발에 참고자료로 활용될 수 있다.