Supercritical water cooled reactor (SCWR) system has been regarded as an innovative reactor system for economical electricity generation due to its simplified direct steam cycle system and high thermal efficiency.

SCWR can be composed as either thermal or fast system owing to its lower coolant density in upper part of the core and its high coolability. Supercritical water-cooled fast reactor (SWFR) system can be accomplished with tighter fuel pin arrangement. Additional moderator is not necessary in the fast system. Accordingly, the supercritical fast reactor system has high power density and consequent more compact pressure vessel size. The high power density and compact size of pressure vessel gives further economical potential in aspect of capital cost reduction as well as plant simplification. Furthermore, steam cycle system and safety system is compatible with those of thermal system. Difference between thermal and fast system is the reactor core.

Fuel rod failure modes and associated fuel rod design criteria that are expected to be limiting in SWFR operating condition were established in fuel rod design study. Maximum linear heat rate of 39kW/m and fuel centerline temperature of 1900℃ are used as thermal design criteria for ensuring fuel integrity. Maximum cladding surface temperature of 650℃ is used to avoid cladding overheating. Flow dynamic pressure of 0.02MPa is used to avoid excessive flow induced vibration. As thermo mechanical criteria, cladding collapse and rupture are considered as design criteria.

The fuel rod design parameters are determined with those limiting fuel rod design criteria including thermo-hydrodynamic consideration and thermo-mechanical consideration and with the design goal of high outlet temperature over 500℃. The fuel rod diameter size and its P/D ratio are determined to be 7.6mm and 1.14, respectively, which is mainly limited by the consideration of excessive flow induced vibration.

Thermo-mechanical behaviors of fuel rod and cladding under irradiation have been analyzed by FEMAXI-6 fuel rod performance code based on finite element method (FEM). High coolant system pressure of 25MPa plays a positive role against PCMI and gas pressure loading, which allows smaller gap and is beneficial for reducing fuel centerline temperature.

Available fuel rod design ranges are determined in terms of gas plenum length and its initial pressure for both upper and lower gas plenum location with quantitative evaluation of creep rupture life fraction of fuel cladding. The minimum gas plenum lengths of fuel rod are predicted to be 110cm and 70cm for upper and lower gas plenum locations, respectively. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse, currently available stainless steels or being developed have a potential for application to SWFR. The mechanical strength required for SWFR fuel rod application is also quantitatively determined by fuel rod analysis.

Past core design studies had been based on 2-dimensional R-Z approximation, which yields large uncertainty in principle core performance such as average coolant outlet temperature by mismatch between assembly power and flow rate. The average coolant temperature widely varies with regard to the number of flow rate regions assumed in R-Z approximation. Local power distribution within an assembly could not be taken from those analyses. The R-Z approximation also restricts core arrangement to only radial heterogeneous core.

Three-dimensional nuclear core design procedure fully coupled with thermal hydraulic calculation, which are based on tri-z fine mesh neutron diffusion solution and single channel analysis, is developed in this study. Evaluation of pin-by-pin power distribution and equilibrium cycle search is implemented in core calculation procedure. Three-dimensional design procedure permits more flexible core arrangement for negative void reactivity and eliminates the concern of mismatching between assembly power level and flow rate, which allow more accurate evaluation of core outlet temperature and cladding surface temperature.

Large commercial scale SWFR core design is investigated for negative coolant void reactivity and high core average temperature with ZrH layer. The core arrangement for negative void reactivity is proposed. The final core has composite type arrangement with internal blanket rings and scattered loading of blanket assemblies in outer core regions. Several design method for improving core outlet temperature are also proposed in this study.

SWFR core could be composed as compact system having core average power density of 156W/cm3 including all blanket regions, which is 1.5 times higher than that of current PWR and is 2.6 times higher than that of Super LWR. The designed core has equivalent diameter of 2.7m with similar active core height. High core average outlet temperature of 503℃ is also achieved by employing radial fuel enrichment zoning in seed assembly and downward flow cooling in some portion of seed fuel assemblies while keeping design criteria of MLHGR of 39kW/m and MCST of 650℃. Assembly average discharge burnup is evaluated to be 68.3MWd/kgHM. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency and discharge burnup give an economical potential in aspect of capital and operating cost.

Subchannel analysis has been carried out for SWFR fuel assembly. Peak cladding surface temperature difference arising from coolant channel heterogeneity is calculated by using the subchannel analysis code and is evaluated to be about 18.5℃. Large subchannel heterogeneity in hexagonal fast reactor fuel assembly is controlled by altering coolant channel flow area rather than adjusting P/D ratio.

Maximum cladding surface temperature at nominal condition is evaluated to be 645.3℃ considering inter-channel flow mixing and coolant channel heterogeneity over the cycle. Several design considerations can be made based on the results of subchannel analysis.

Local power peaking in upward fuel assembly should be kept below 1.2 to avoid excessive cladding surface temperature deflection and to keep conservatism of core design calculation coupled with single channel analysis. Employing downward flow in the region having high local power peaking factor is effective to suppress excessively high PCST difference between single and subchannel analyses in peripheral core regions in which seed assemblies have steep power gradient.

Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Uncertainties involved in system parameter variation, nuclear enthalpy rise hot factor and engineering temperature rise hot factor are taken into account in statistical thermal design uncertainty evaluation.

Maximum thermal design uncertainty associated with MCST is evaluated to be 31℃ by Monte-Carlo sampling procedure and is in a good agreement with that from RTDP method. Effect of downward flow in seed region on sensitivity is investigated by improved Monte-Carlo thermal design procedure. Employing downward flow in seed assemblies are expected to be beneficial for reducing contributions of local parameter uncertainties to total uncertainty. The downward flow cooling of seed assemblies reduce total coolant enthalpy rise in hot assembly, which reduce the sensitiveness of local parameter uncertainties.

MCST including statistical uncertainty is predicted to be 681℃ (650℃ for nominal + 31℃) for SWFR which ensure 95/95 limit. This value provides one of limiting thermal condition that might be occurred during normal operation within engineering uncertainty and can be used as initial condition of safety analysis and for evaluation of creep behavior or stress corrosion cracking of fuel cladding under irradiation.

The conceptual design of large scale supercritical pressure light water cooled fast reactor is carried out comprehensively with considerations of fuel rod mechanical behavior, nuclear core physics and thermal hydraulics including statistical uncertainty. High temperature operation of SWFR is feasible within thermal and mechanical design criteria which do not significantly exceed those of LMFBR design.

　本論文は大型超臨界圧軽水冷却高速炉(SWFR)の炉心設計の研究で7章より構成されている。

　第1章は序論で対象とした原子炉の一般的な特徴と特性について述べている。超臨界圧軽水冷却高速炉に関しての先行研究は2次元解析であるため、炉心出口温度、冷却材ボイド反応度、局所ピーキング係数などの熱的、核的性能や制限値の評価に限界があり、炉心設計と性能の正確に評価する為には核熱結合3次元炉心設計が必要であるとしている。

　第2章は炉心の目標性能と熱的・核的制限値と炉心解析法について述べている。炉心解析は核燃料棒設計、炉心核設計、核燃料集合体の内のサブチャネル解析、統計的熱設計で構成されている。

　第3章は核燃料棒設計について述べている。核燃料棒健全性を確保する熱的・機械的設計制限値をもとめるとともに、燃料棒挙動解析コードを用いて核燃料設計のためのパラメータサーベイを行い、SWFRに使われる核燃料設計値を決めている.定常運転時に炉心内の最も厳しくなる条件で核燃料性能を評価した結果ステンレス鋼被覆管で機械的設計制限値を満足することを示している

　第4章は炉心核設計を述べている。炉心核設計の方法として3角燃料格子を扱える3次元微細格子中性子拡散法と単チャンネル解析法を結合した3次元核熱結合計算方法を開発し、平衡炉心に対して炉心の特性を評価している。水素化ジルコニウム層をシードとブランケット集合体の間に置く炉心を用い、様々な装荷パターンに対してボイド反応度を評価して最も負のボイド反応度を持つ炉心を決めている。炉心平均出口温度を高温化するためには核燃料集合体の内の局所出力を制御する必要があるのでブランケット集合体には厚いラッパー管を用い出力のピークを抑えるとともに,シード集合体では集合体内のプルトニウム富化度に分布を持たせている。さらに局所出力分布の制御が難しい集合体に対しては下降流冷却を採用している。これにより650℃の最高被覆管温度の制限値を満足しながら500℃以上の炉心平均出口温度という目標を達成している.

　第5章はサブチャンネル解析による定常運転中の最高被覆管温度を評価している。サブチャンネル面積を燃料棒出力にあわせて調節することで最高被覆管温度上昇を低減できることを示している。集合体内の局所出力ピーキング係数を1.2以下の領域では単チャンネル解析により被覆管温度が保守的に評価されることを示している.全体炉心のサブチャンネル解析の結果、最高被覆管温度は制限値である650℃を満足することを示している。

　第6章はSWFRの統計的熱設計に関して記述している。定常運転時の流量などのシステム変数の変動及び被覆管温度評価に関係する不確定性因子を考慮し、モンテカルロサンプリングにより不確定性を考慮する確率論的方法で解析している。下降流冷却チャンネルを考慮するために単チャンネル解析とサブチャンネル解析を結合して評価している。加圧水型軽水炉または液体金属冷却高速増殖炉に適用されて来た不確実因子参考に解析し工学的不確実度による最高被服管温度上昇は31℃であることを示している。

　第7章は結論であり、本研究のまとめが述べられている。

　以上を要するに本論文は 大型超臨界圧軽水冷却高速炉3次元炉心設計研究を行い、炉心の核的及び熱的な特性を明らかにしている。この成果はシステム量子工学の進歩に貢献することが少なくない。よって本論文は博士(工学)の学位請求論文として合格と認められる。