台灣研究用反應器(TRR)用過燃料含有武器等級鈽元素，燃料安定化作業將把金屬TRR用過燃料經由熱室燒製處理成氧化物粉末。為了核子保防，本研究將於燃料安定化前後，分別以非破壞性檢測方法測定鈽元素含量。
我們建立兩套中子偵測系統，即Spent Fuel Neutron Counter (SFNC)系統和Neutron Coincidence Counter (NCC)系統，分別用於量測TRR 用過燃料棒和燒製成氧化物粉末中Pu-240含量。鈽元素總量則依MCNPX計算分析TRR用過燃料之鈽元素各同位素含量，求出Pu/240Pu重量比率加以轉換。SFNC 系統偵檢效率校正係以一全新的TRR燃料棒置於一乾淨的水槽中測量為之，並經MCNPX 程式模擬計算驗證。七根TRR 用過燃料棒之測量則在鄰近用過燃料貯存池之相連水槽中進行。除了已氧化和/或斷裂的用過燃料棒外，測得鈽元素含量與根據燃耗資料之MCNPX計算值相當吻合。而NCC 系統的偵檢效率，包含總計數和雙重計數效率之校正係以U3O8 粉末測量搭配MCNPX 程式計算獲得。三根TRR 用過燃料棒經熱室燒製成之氧化物粉末，以NCC 系統在熱室測得之鈽元素含量，與安定化前測定結果，差異皆在18%以內。
總結而論，本論文之非破壞性檢測技術可準確測定TRR用過燃料的鈽元素含量，透過安定化前後的兩次測量，更證明TRR用過燃料的鈽元素含量並無明顯的自然或人為流失。

Taiwan Research Reactor (TRR) spent fuel rods contain weapons-grade plutonium. For the nuclear safeguard purpose, this work aims to determine the plutonium content in the TRR spent fuel rods by using the nondestructive neutron measurement before and after the fuel stabilization process, by which the metallic spent fuel rods will be transformed to oxide power for interim storage.
We established two neutron detection systems, namely, the SPent-fuel Neutron Counter (SPNC) system and the Neutron Coincidence Counter (NCC) system for determining the Pu-240 content in the TRR spent fuel rod and in the transformed oxide powder, respectively. The measured Pu-240 contents in the TRR spent fuel were then converted into total plutonium contents by using the MCNPX-calculated Pu-to-240Pu mass ratio. The detection efficiency of the SPNC system was calibrated by the measurement of a TRR fresh fuel in a water trench filled with clean water and verified by the MCNPX simulation calculation. Seven TRR spent fuel rods were measured in the nearby water trench connected to the spent fuel storage pool. The measured plutonium contents agree well with the MCNPX-calculated values based on the burnup history of each spent fuel rod, except for the broken rods with fuel oxidation. The detection efficiencies consisting of singles rate and doubles rate efficiencies of the NCC system were determined by using measurements using U3O8 powder in couple with MCNPX calculations. Canisters filled with oxide powder converted from three TRR spent fuel rods were measured by the NCC system in the hot cell. The measured plutonium contents of the oxide powder from the spent fuel rods agree well (<18%) with the measured value before the fuel stabilization process.
In conclusion, our nondestructive neutron detection systems can accurately determine the plutonium content in TRR spent fuel. Furthermore, the two measurements for each TRR spent fuel prove that there is no obvious process and artificial loss of plutonium contents during the fuel stabilization process.

第 1 章 緒論 1
1.1 引言 1
1.2 文獻回顧 2
1.3 研究方向 4
第 2 章 TRR 用過核燃料 6
2.1 TRR簡介 6
2.2 TRR用過核燃料 10
2.3 燃料安定化作業 11
第 3 章 原理與方法 16
3.1 中子射源 16
3.2 中子計數量測 18
3.2.1. 中子偵檢器 19
3.2.2. 加馬堆疊效應 24
3.2.3. 中子符合計數 29
3.3 用過燃料特性和中子偵檢器效率計算 34
3.3.1. 計算機程式簡介 34
3.3.2. TRR用過燃料核種組成 34
3.3.3. 射源項分析 41
3.3.4. 中子偵檢器設計 43
第 4 章 用過燃料燃耗計算分析 47
4.1 SAS2H計算模型 47
4.2 MCNPX計算模型 49
4.3 用過燃料核種組成 53
4.4 中子和光子射源項 59
4.5 鈽元素計算值和定量方法 64
第 5 章 安定化前鈽元素定量 68
5.1 中子偵檢系統設計 68
5.2 系統校正 70
5.3 靈敏度分析 73
5.4 結果與討論 77
第 6 章 安定化後鈽元素定量 88
6.1 中子偵檢系統設計 89
6.2 閘門設定 92
6.3 系統校正 95
6.4 功能測試 96
6.5 靈敏度分析 98
6.6 結果與討論 102
第 7 章 安定化前後鈽元素定量 107
第 8 章 結論與建議 108
8.1 結論 108
8.2 未來研究方向 109
參考文獻 111
附錄 A INCC程式輸出檔 115
附錄 B MAKXSF程式輸入檔 118
附錄 C TRR用過燃料棒計算模型(MCNPX)輸入檔 121
附錄 D NCC系統之MCNPX計算輸入檔 134
發表論文 137

1.D. Reilly, N. Ensslin, and H. Smith, Jr. , Passive Nondestructive Assay of Nuclear Materials. NUREG/CR-5550, LA-UR-90-732, Los Alamos Scientific Laboratory, 1991.
2.S.T. Hsue, T.W. Crane, W.L. Talbert, and J. C. Lee, Nondestructive Assay Methods for Irradiated Nuclear Fuels, Los Alamos Scientific Laboratory report LA-6923, 1978.
3.S. Muralidhar, R. Tripathi, B.S. Tomar, G.K. Gubbi, S.P. Dange, S. Majumdar, and S.B. Manohar, Nondestructive Assay of the Fissile Content in FBTR Fuel Pins, Nucl. Instr. and Meth. A, vol. 51, pp. 437-443, 2003.
4.D. Davidson, J. Verplancke, P. Vermeulen, H. O. Menlove, H.G. Wagner, B. Brandalise, and M. Stutz, A new high Accuracy Combined Neutron/Gamma Counter for in-glove Box Measurements of PuO2 and MOX Safeguards Samples (OSL-Counter), Proc. 15th ESARDA Symp. on Safeguards and Nucl. Mater. Management, Rome, EUR 15214 EN, pp. 585-588, 1993.
5.L.C-A Bourva, S. Croft, H. Ottmar and D.R. Weaver, MCNP Modelling of a Combined Neutron/Gamma Counter. Nucl Instrum Meth A, vol. 426 pp. 503-517, 1999.
6.Briesmeister, J. F., Ed. MCNP—A general purpose Monte Carlo N-particle transport code, version 4C, LANL Report LA-13709-M (Los Alamos National Laboratory, Los Alamos, NM), 2000.
7.A. L. Thornton, Development of a Portable Neutron Coincidence Counter for Field Measurements of Nuclear Materials Using the Advanced Multiplicity Capabilities of MCNPX 2.5.F and the Neutron Coincidence Point Model, Master’s Thesis, 2007.
8.O. W. Hermann, S. M. Bowman,M. C. Brady,and C. V. Parks, Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, 2000.
9.J.C. Tait, I. Gauld and A.H. Kerr, Validation of the ORIGEN-S Code for Predicting Radionuclide Inventories in used CANDU Fuel, Journal of Nuclear Materials, vol. 223, pp. 109-121, 1995.
10.J.P. Lestone, J.M. Pecos, J.A. Rennie, J.K. Sprinkle, P. Staples, K.N. Grimm, R.N. Hill, I. Cherradi, N. Islam, J. Koulikov, and Z. Starovich, The passive Nondestructive Assay of the Plutonium Content of Spent-fuel Assemblies from the BN-350 Fast-breeder Reactor in the City of Aqtau, Kazakhstan. Nucl. Instr. and Meth. A, vol. 490, pp. 409-425, 2002.
11.黃盛年，桃園重水式核反應器概要，INER-C0211。
12.李顯謨，桃園重水式核反應器實驗數據與理論值，NER-64-J-T0463，1975。
13.H.C. Hsieh, H.T. Kung, N.N. Hsu, and J.H. Cheng , Experience in TRR Fuel Element Fabrication, Nuclear Science Journal, vol. 13, pp. 53-69, 1976.
14.S.Y. Lee, H.O. Menlove, J.B. Marlow, M.C. Mildder, C.D. Rael, J.B. Wang, and M.C.Yuan, Calibration of the Spent-fuel Pluconium Coincidence Counter at the Taiwn Research Reactor, INMM48, 2007.
15.T. Qian, P. Tonner, N. Keller, and W. Buyers, A Method for Comparing Degradation of Boron Trifluoride and Helium Detectors in Neutron and Gamma Fields, IEEE Transactions on Nuclear Science vol. 45 (3) pp. 636-642, 1998.
16.A.J. Stokes, T.J. Meal, and J. E. Myers, Improved Performance of BF3 Neutron Counters in High Gamma Fluxes, IEEE Transactions on Nuclear Science, vol. 13 (1), pp. 630-635, 1966.
17.D.H. Beddingfield, N.H. Johnson and H.O. Menlove, He-3 Neutron Proportional Counter Performance in High Gamma-ray Dose Environments, Nucl. Instrum. Meth. A, vol. 455 (3) pp. 670-682, 2000.
18.http://140.114.122.100/download.php?filename=94_38a487ec.pdf&dir=news&title=檔案下載
19.C.L. Chen and S.H. Su, The Calibration System of Air Kerma Rate for 137Cs Gamma-ray at 4.23 Meters, INER-5000, 2007.
20.J.E. Swansen, P.R. Collinsworth and M.S. Krick, Shift-Register Coincidence Electronics System For Thermal Neutron Counters, Nuclear Instruments and Methods, vol. 176, pp. 555-565, 1980.
21.J. E. Stewart, S.C. Bourret, M.S. Krick, M.R. Sweet, T.K. Li and A. Gorobets, New Shift Register Electronics for Improved Precision of Neutron Coincidence and Multiplicity Assays of Plutonium and Uranium Mass, LA-UR-99-4927, 1999.
22.SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version 5.1, Vols. I–III, 2006.
23.Denise B. Pelowitz, MCNPXTM User’S Manual, LA-CP-07-1473, 2008.
24.C. J. Chang and J. T. Yang, TRR Spent Fuel Model for the ORIGEN 2 Code, Institute of Nuclear Energy Research, INER-0735, 1987.
25.O. W. Hermann and C. V. Parks, SAS2H: A Coupled One-Dimensional Depletion and Shielding Analysis Module, NUREG/CR-0200, 2000.
26.I.C. Gauld, O.W. Hermann and R.M. Westfall, ORIGEN-S:SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms. Vol. II, Book 1, Sect. F7 ORNL/TM-2005/39, 2006.
27.Michael L. Fensin, John S. Hendricks, Samim Anghaie, The Enhancements and Testing for MCNPX 2.6.0 Depletion Capability, Nuclear Technology, vol. 170 pp. 68-79, 2010.
28.O. W. Hermann and M. D. DeHart, Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel, ORNL/TM-13315, 1998.
29.W. B. Wilson, R. T. Perry, W. S. Charlton, T. A. Parishand E. F. Shores, SOURCES: A Code for Calucluating (a,n), Spontaneous Fission, and Delcayed Neutron Sources and Spectra, Radiation Protection Dosimetry, vol. 115 (14) pp. 117-121, 2005.
30.I. C. Gauld, E. F. Shores, and R. T. Perry, New Neutron Source Algorithms In the ORIGEN-S Code, 12th Biennial RPSD Topical Meeting, April, 2002.
31.W.S. Charlton, R.T. Perry, W.B. Wilson, Benchmarking of the Los Alamos neutron production rate code SOURCES-3A, Nucl. Instrum. Meth. B, vol. 140 pp. 1-8, 1998.
32.John Hendrick, Multiplicity PowerMonte Carlo Sampling of Fission Multiplicity, MC2005, Chattanooga, 2005.
33.Forrest B. Brown, The makxsf Code with Doppler Broadening, LA-UR-06-7002, 2006.
34.Reactor Physics Constants, ANL-5800, 1963.
35.C.C. Tseng, S.C. Cheng, K.Y. Hunag, Analysis of TRR Core Parameters Related to Fuel Irraditation From the Post-Irradiation Gamma Scanning Results, INER-T1150, 1986.
36.C.S. Taso, T.T. Yang, T. Lee and L.F. Lin, Nondestructive Burnup Determination of TRR Testing Fuel Rods FR-01 and S1 by 137Cs Counting Rate and 134Cs/137Cs Activity Ratio from Gamma Spectroscopy, INER-T1251, 1987.
37.C.H. Delegard and A.J. Schmidt, Uranium Metal Reaction Behavior in Water, Sludge, and Grout Matrices, PNNL-17815, 2008.
38.M.S. Krick, Thermal Neutron Multiplicity Counting of Samples with Very low Fission Rates, Los Alamos National Laboratory, LA-UR-97-2649, 1997.