本研究透過系統性的分析探討混凝土在各式加速器環境中的屏蔽與活化特性，為配合台灣主要加速器應用的輻射防護需求，特別針對放射性物質產製的迴旋加速器、質子/碳離子癌症治療的加速器以及高能電子同步加速器等設施輻射場進行深入比較與統合分析。研究成果主要展現在以下四項重要觀察或結論：(1)有自屏蔽設計的放射性物質生產設施可大幅降低室內中子通率並簡化未來設施除役問題，考量11種(p,xn)核種產製情節所誘發的混凝土活化潛勢，經中子產率歸一化處理後，活化潛勢呈現高度相似分布；(2)質子治療加速器運轉引發的輻射場以高能中子為主導，加入金屬夾層的混凝土屏蔽效果有限，遠不如其在光子為主之直線加速器輻射場的表現；(3)計算26種不同組成之混凝土在三種加速器輻射場下的屏蔽表現，輻射衰減長度與混凝土密度具有高度線性相依特性；(4)比較二種常見簡化情節與接近實際運轉之週期性照射，基於核種活化的預測差異，據以建立合適的核種活度修正因子，並引入二個無因次的關鍵因子，解決簡化情節活化評估引起的誤差。
第一部份探討台灣13座放射性物質生產設施的運轉現況，7座有自屏蔽設計者可大幅降低室內中子通率並減少混凝土活化，簡化未來設施除役問題；彙整11種在迴旋加速器常見之(p,xn)產製情節的二次中子特性，包含18O(p,n)18F、15N(p,n)15O、68Zn(p,n)68Ga與68Zn(p,2n)67Ga、69Ga(p,2n)68Ge、89Y(p,n)89Zr、100Mo(p,pn)99Mo與100Mo(p,2n)99mTc、112Cd(p,2n)111In、124Xe(p,2n)123Cs以及203Tl(p,3n)201Pb。11種情節下的二次中子特性經產率歸一化後高度相似，同時由二次中子所引發之混凝土活化潛勢亦呈現相似的空間分布，此一特性有利於簡化設施歷年運轉情節的評估設定。第二部份探討中國醫質子治療設施夾層式隔間牆(0.9公尺混凝土/0.2公尺鐵/0.9公尺混凝土)的屏蔽表現。研究結果顯示此一夾層式設計在高能中子主導之輻射場下，相比於同厚度全混凝土的屏蔽，只有多提供2倍的劑量衰減；相較之下，此一夾層設計在光子為主之直線加速器環境可多提供20倍的劑量衰減。第三部份彙整比較26種不同組成之混凝土在三種加速器輻射場下的屏蔽表現，並將屏蔽表現與四個混凝土的材料特性進行分析，包含密度(2.1 – 5.9 g/cm3)；原子序(7.1 – 19.6)；氫含量(0.05 – 2.2 wt%)及重元素含量(0.6 – 88.1wt%)。分析結果發現，衰減長度與密度的變化呈現高度相關，透過線性迴歸定義各種輻射場下的變化方程式，此成果可提供相關設施迅速評估混凝土密度改變後的屏蔽表現差異。另外，本研究結果顯示，氫含量的增加對於高能中子場中的屏蔽表現幾乎沒有貢獻。第四部份比較二種常見簡化情節與接近實際運轉之週期性照射，建立適當的修正因子，解決簡化連續照射情節所產生的核種活化預測誤差。本研究並進一步引入兩個關鍵的無因次參數(工作因子與半衰期因子)，使修正因子適用於大部份照射情節的加速器設施，此一成果有利於使用連續照射的假設並減少簡化情節對活化預測的誤差。

This research systematically investigates the shielding performance and activation susceptibility of concrete against radiation fields induced by various accelerators, including radioisotope production facilities, proton/carbon-ion therapy facilities, and GeV electron synchrotron facilities. The main contributions are as follows: (1) Characteristics of neutron production and concrete activation in cyclotron vaults for self-shielded and non-self-shielded facilities were examined in detail. The effects of various (p,xn) production routes on the neutron-induced long-lived radioactivity in the concrete walls of a typical cyclotron vault were compared. (2) In comparison with that of a concrete wall of the same thickness, the shielding performance of a concrete-iron-concrete sandwich wall was found only marginally better, further reducing the transmitted dose rate by approximately a factor of 2 against secondary neutrons generated in a proton therapy facility. (3) The performance of 26 types of concrete in three accelerator radiation environments were evaluated. The concrete density was found to have the most direct and correlated effect on the resulting dose attenuation length in concrete. (4) Two simplified scenarios with continuous irradiation are commonly used instead of a periodic irradiation scenario. This study examined residual radionuclide productions under these three scenarios and introduced correction factors to enhance the practical applicability of the two simplified models.
In the first part, this study examined the induced activity of concrete in 13 radioisotope production cyclotrons (seven have self-shielding designs and six are non-self-shielded types) in Taiwan. Eleven (p,xn) related nuclear reactions, 18O(p,n)18F, 15N(p,n)15O, 68Zn(p,n)68Ga and 68Zn(p,2n)67Ga, 69Ga(p,2n)68Ge, 89Y(p,n)89Zr, 100Mo(p,pn)99Mo and 100Mo(p,2n)99mTc, 112Cd(p,2n)111In, 124Xe(p,2n)123Cs, 203Tl(p,3n)201Pb, were involved in their operation. For the considered self-shielded facility, the induced activities in the walls and roof of the cyclotron room were substantially lower than the clearance level; only the concrete activation in the floor could be a problem. In addition, the effects of various (p,xn) production routes on the neutron-induced long-lived radioactivity in concrete were compared. After neutron yield normalization, there is no significant discrepancy among these reactions. This observation offers valuable insights for professionals involved in decommissioning of cyclotrons used for radioisotope production.
In the second part, the shielding performance and activation susceptibility of a sandwich wall in the CMUH proton therapy facility were investigated. In comparison with that of a concrete wall of the same thickness, the shielding performance of the concrete-iron-concrete wall was only marginally better, further reducing the transmitted dose rate by approximately a factor of 2, as compared with an approximately 20-fold increase in the dose attenuation for 18-MV photons.
Different types of concrete contain various elements in different proportions and therefore may exhibit different shielding performances. In the third part, the performance of 26 types of concrete in three accelerator radiation environments were evalauted. The concrete density was found to have the most direct and correlated effect on the resulting dose attenuation length in concrete. Fitting parameters of a linear relation using the calculated data were determined and the result can be used to estimate the effect of concrete composition on its shielding performance.
The residual radioactivity in accelerator components depends on the irradiation profile employed. As an alternative to a fully periodic scenario, in practice, researchers frequently resort to two simplified continuous irradiation scenarios. In the fourth part, this study offered a systematic comparison of residual radionuclide productions under these three scenarios and introduced correction factors to enhance the practical applicability of the two simplified models.

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