在沸水式反應器中，冷卻水會在反應器中發生沸騰的現象，在高功率低流量的情況下容易發生結合熱流與中子效應的不穩定事件，通常以頻域的分析方法界定穩定運轉區域，是防範不穩定事件的主要控制方法。為了因應日新月異的電廠相關組件與新型的核燃料設計，因此引進了頻域的穩定性分析程式LAPUR6.0。本論文的主要目的就是參考之前建立的LAPUR5.2穩定性分析技術與理論，建立自有的LAPUR6.0穩定性分析模式並更新自動化程式DRASM。文中首先說明新增卡號的功能與理論模式並交代卡號建立的過程，然後介紹目前的LAPUR6.0分析程序，最後將比較新舊版程式分析的結果差異並利用LAPUR6.0分析1.使用半長燃料棒之後，2.利用不同方法模擬局部壓降的計算，3.選用不同摩擦模型以上這三個狀況對於穩定性的影響。從核一廠二號機週期25燃料裝填的數據比較中可以發現，LAPUR6.0的衰減率會比LAPUR5.2小，尤其是異相衰減率的差距更大。原因除了程式本身差異的影響外，主要就是因為後面分析的這三種情況造成的，使用摩擦模型II對於穩定性的影響最大，算出來的衰減率比模型I低很多；使用半長棒也會降低系統的衰減率，但是不像摩擦模型的影響這麼大；加入局部壓降的計算則對衰減率沒有很明顯的影響。

In boiling water reactors, the cooling water may takes place of boiling phenomenon in the core. At low flow and high power conditions, it may easily induce instability by unstable power or flow oscillation. It’s coupled mechanisms of neutronic and two-phase flow thermal-hydraulic behaviors. In order to fit in with the newest fuel designs, we import a new version of frequency domain stability program, LAPUR6.0. Based on the LAPUR5.2 methodology, the main subjects of this paper are to establish a methodology and analytic mode of LAPUR6.0 and to update the auto search program DRASM.
At first, this paper illustrates the new input cards and its setting process, and introduces the current analytical procedures of LAPUR6.0. Then compare the results between new version and old version. Finally, this paper uses LAPUR6.0 to see the differences of decay ratio at the following situations, the use of part length rods, using different method to calculate local loss and choosing different friction models. From the results of CS2C25 reload design, it can found that decay ratios by LAPUR6.0 are smaller than LAPUR5.2, especially the regional mode decay ratios. Except for the influence of program calculation variation, the other reasons are those three situations. The use of friction model II cause the greatest impact in stability, results in smaller decay ratio. The use of part length rods also reduce the decay ratio, but not as big as friction model II. As for local loss, there is no apparent difference by adding the calculation of local loss.

[1] Carlos J. Gavilan Moreno, “Using the Hurst’s exponent as a monitor and predictor of BWR reactor instabilities”, Ann. Nucl. Energy, vol.37, 2010, pp.434-442.
[2] Umbager J. A., Digiovine A. S., “SIMULATE-3, Advanced Three Dimensional Two-Group Reactor Analysis Code. User’s Manual”, Studsvik/SOA-92/01, 1992.
[3] Fukuda, M., and Kobori T. (1979) J. Nucl. Sci. Technol., 16, 95.
[4] Van Der Hagen, T.H.J.J., D.D.B. Van Bragt, F.J. Van Der Kaa, “Exploring the Dodewaard Type-I and Type-II Stability; From Start-Up to Shut-down, From Stable to Unstable”, Ann. Nucl. Energy, vol. 24, 1997, pp.659-669.
[5] Nayak, A.K., Vijayan, P.K., Saha, D., Venkat Raj, V., Aritomi, M., “Study on the Stability Behavior of a Natural Circulation Pressure Tube Type Boiling Water Reactor”, Nuclear Engineering and Design, vol.215, 2002, pp.127-137.
[6] Anderson, T.T., “Hydraulic impedance: A tool for predicting boiling loop stability”, Nucl. Appl. & Tech., vol.9, 1970, pp.422-433.
[7] Carmichael, L. A., Neimi, R. O., “Transient and Stability test at Peach Bottom Atomic Power Station Unit 2 at end of cycle 2”, EPRI NP-564, 1978.
[8] Woffinden, F. B. and Niemi, R. O., “Low-flow stability tests at Peach Bottom Atomic Power Station Unit 2 at end of cycle 3”, EPRI NP-972, 1981.
[9] Enomoto, T., Muto, S., Ishizuka, T., Tanabe, A., Mitsutake, T. and Sakurai, M., “Thermal Hydraulic Stability Experiments in Rod Bundle, Proceedings of the Third Int. Topical Meeting on Reactor Thermal Hydraulics”, Newport, R. I., USA, Vol. 1, 1985, Paper 9.B.
[10] Kruijf, W.J.M., Sengstag, T., Haas, D.W., Van der Hagen, T.H.J.J.,” Experimental Theermohydraulic Stability Map of a Frenon-12 Boilig Water Reactor Facility with High Exit Friction”, Nuclear Engineer and Design, vol.229 , 2004, pp.75-80.
[11] Wallis, G. B. and Heasly, J. H., “Oscillation in Two-Phase Flow Systems”, J. Heat Transfer, ASME, 1961, pp. 363-369.
[12] Lahey, R. T. and Moody, F. J., “The Thermal Hydraulics of a Boiling Water Nuclear Reactor”, American Nuclear Society, Hinsdale, 1977, I11.
[13] Saha, P., Zuber, N., “An Analytical study of the Thermally Induced Two-Phase Flow Instabilities Including the Effect of Thermal Nonequilibrium”, Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 415-216.
[14] March-Leuba, J., Pere, R.B. and Cacuci, D. G., “Calculation of Limit Cycle Amplitudes in Commercial Boiling Water Reactors”, Trans. Am. Nucl. Soc., Vol. 46, 1984.
[15] Takigawa, Y., Takeuchi, Y., Tsunoyama, S., Ebata, S., Chan. K.C., Tricoli, C., “Coarso limit cycle analysis with three-dimensional transient code TOSDYN-2”, Nucl. Technol., vol.79, 1987, pp.210-218.
[16] Yunlin Xu, Thomas Downar, R et. “Application of TRACE/PRACS to BWR stability analysis”, Annals of Nuclear Energy, vol.36, 2009, pp.317-323.
[17] Jose E. T.-F., Alfonso P.-G., Gilberto E.-P. “Decay ratio estimation based on time-frequency representations”, Annals of Nuclear Energy, vol. 37, 2010, pp.93-102.
[18] “BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2”, EMF-CC-074(P)(A) ,Rev.0, 2000.
[19] “STAIF：A Computer Progarm for BWR Stability Analysis in the Frequency Domain”, EMF-2279(P) ,Rev.0, 2001
[20] “Cycle –Specific DIVOM Methodology Using the RAMONA5-FA Code”, BAW-10255PA, Rev.2, 2008
[21] RAMONA5 Version 2.4, “Users and Theory Manuals”, Studsvik-Scandpower.
[22] J. March-Leuba, and Otaduy, P. J., “A Comparison of BWR Stability measurements with calculations using the code LAPUR-ΙV”, ORNL/TM-8546, Oak Ridge National Laboratory, January 1983
[23] Umbager J. A., Digiovine A. S., “SIMULATE-3, Advanced Three Dimensional Two-Group Reactor Analysis Code. User’s Manual”, Studsvik/SOA-92/01, 1992.
[24] Escriva, A. and Munoz-Cobo, J. L., “PAPU Models, Correlations, and User’s Mannul”, ThermalHydraulic and Nuclear Engineering Group, GTIN-02/001, March 2002.
[25] Alberto E. and Jose L. M. C., “LAPUR 6.0 User’s Manual”, NUREG/CR-6958, ORNL/TM-2007/233, October 2008.
[26] 「核一廠CS1-C25 ATRIUM-10燃料SIMULATE-3流量分佈模式計算書」， NED-CM-96B16809-CCS-029-01.
[27] 「核二廠KS2C21 ATRIUM-10燃料SIMULATE-3流量分佈模式計算書」， NED-CM-96B16809-CCS-031-01
[28] AREVA NP, Inc., “Thermal-Hydraulic Evaluation of ATRIUM™-10 Fuel in the Chinshan Units”, KSQ: 06:009, August 30 2006.
[29] AREVA NP, Inc., “Chinshan Hydraulic Characteristics for Modeling ATRIUMTM-10 Fuel Assemblies with Advanced Fuel Channel and FUELGARDTM Lower Tie Plate in MICROBURN-B2”, 51-9075557-000, March 14 2008.
[30] 謝昌倫，「沸水式核反應器穩定性分析之LAPUR方法論發展與驗證」，國立清華大學，博士論文，民國97年12月。
[31] Dix et al., “Two phase pressure drop reduction BWR assembly design”, United States Patent：Patent Number 5017332, May 21, 1991.
[32] Lotfi Belblidia, Gerardo Grandi and Carlos Aguirre, “SIMULATE-3K stability benchmarking and predictive calculations of Leibstadt”, International Conference on Reactor Physics, September 2008.
[33] Masahiro Furuya et al., “Development of BWR Regional Stability Experimental Facility SIRIUS-F”, NURETH-11 paper-233, Avignon, France, October 2-6, 2005
[34] March-Leuba J., “LAPUR benchmark against in-phase and out-phase stability tests”, NUREG/CR-5605, ORNL/TM-1162, 1990.
[35] Hotta Akitoshi, et al., “Regional Instability Evaluation of Ringhals Unit 1 Based on Extended Frequency Domain Model”, Nucl. Eng. Design, vol.200, 2000, pp. 201-220.
[36] Wallter Ambrosini, “Assessment of flow stability boundaries in a heated channel with different fluids at supercritical pressure”, Annals of Nuclear Energy, vol.38, 2011, pp.615-627.
[37] 「台灣電力公司第一核能發電廠營運程序書」，程序書編號1002.8，版次21，民國97年6月。
[38] 苑瑞盈，「台灣電力公司98年度OPRM與爐心穩定性訓練課程講義」，2009年1月。
[39] 「核四訓練教材」，章節1.3，爐心與核燃料。