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研究生:陳信宇
研究生(外文):Chen, Shin-Yu
論文名稱:應用於清大水池式反應器之數位式寬程中子監測及反應度量測系統
論文名稱(外文):Digital Neutron Monitor and Reactivity Meter for THOR
指導教授:周懷樸
指導教授(外文):Chou, Hwai Pwu
學位類別:碩士
校院名稱:國立清華大學
系所名稱:核子工程與科學研究所
學門:工程學門
學類:核子工程學類
論文種類:學術論文
論文出版年:2014
畢業學年度:102
語文別:中文
論文頁數:143
中文關鍵詞:清華反應器寬程中子監測系統反應度量測
外文關鍵詞:THORWide Range Neutron Monitoring SystemReactivity measurement
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因應世界能源使用量成指數上升,且核能安全的議題更加備受矚目,世界各地電廠與核儀電子相關領域開始有數位化的趨勢,除了能夠更加準確掌握核能,對於整個儀控系統的調校也更加容易之外,最大的主因還是由於半導體技術的進步使得數位電路的演進越來越快,讓儀控系統不容易隨時間而有特性飄移,造成量測上的誤差,也替在反應器功率監測中負責大範圍的功率監測以及週期運算、扮演相當重要角色的寬程中子監測系統開啟了數位化的開端。
藉著分析與量測清大寬程中子監測系統的機會,本研究根據模組特性與線上量測結果,利用可程式邏輯閘(Field Programmable Gate Array, FPGA)開發全新的數位監測系統與研究數位化方法。自製的數位監測系統原型規格為12位元解析度、兩組不同增益ADC通道、488 kHz 取樣頻率、16.7毫秒取樣區段、零無感時間與兩組計數器。另外為了增加量測的準確度,此量測系統的前端電路使用了兩組ADC通道與無交握資料傳輸,使所設計之FPGA數位監測系統成功達到零無感時間與全功率範圍功率量測,特性與原安裝在THOR的類比系統特性相當。相較於過去示波器大於96 %及Photon II資料擷取卡大於80%的無感時間,使用FPGA為架構的數位系統原型利用零無感時間的特點,使得觀測波形更完整,證實自製的數位系統計算的即時資料更貼近類比電路計算結果。
本研究同時也根據核反應器的動態特性,針對自製的數位系統另外建構一套即時反應度量測系統,目前已可順利應用於清華水池式反應器(Tsing-Hua Open-Pool Reactor, THOR)的例行運轉監測。此量測系統的前端數位電路應用DuBridge, R. A. 所發展的Campbell定理計算反應器的即時功率,並利用商業化的FPGA開發板實現Campbell定理的數位化演算,進而得到所需的即時功率資料。另一方面,反應度監測系統應用反應器動力學中的點反應器動態特性(Point kinetics)與延遲中子母核(Delayed neutron precursors)對時間的變化函數,並在適當的取樣時間間隔以及已知的各項運轉參數下,將功率對時間的變化轉換為反應度,同時利用內建的驗證機制將所計算的即時資料與原廠系統監測的數據比對,以提升反應器運轉的安全性。
本研究所建構的數位式寬程中子監測及反應度量測系統使用美商國家儀器公司(National Instruments)的LabVIEW開發出直覺的人機介面,搭配動態歷史圖表提供運轉員更多元的反應器即時資訊。於線上量測計算中,發現所得之功率資料經過60點移動平均後與類比系統輸出功率相關係數高達0.9992以上,而利用60點小最平方化擬合計算後得到的即時週期資訊亦與NM-1000原廠所得的數據相當,顯示自製的即時數位監測及反應度量測系統在精確度與反應時間都非常接近THOR原廠安裝之類比監測系統。

Owing to the consumption of the resources of energy increases exponentially and the conventional topic of nuclear safety becomes more important to be taken into account than before, there is a tendency that nuclear power plants all over the world and their nuclear instruments are in the beginning of being digitalized. In addition to being enough to take over the nuclear technology precisely which makes the calibration of the whole instrumentation and control system becomes easier, the fully-developed technology of semiconductor reduces the errors from experiments with its characteristic of stability. As a result of the instrumentation and control system plays an important role in wide-range reactor power monitoring and reactor period estimation, digitalization of the wide-range neutron monitoring system is conducted in the nuclear field.
A new digital monitoring system is developed based on Field Programmable Gate Array (FPGA) technique according to reports of characteristics of modules and online measurement results through the analysis of the wide-range neutron monitoring system (WRNM) in Tsing-Hua Open-Pool Research Reactor (THOR), with the study of digitalizing methods included. The specification of the self-developed digital 12-bits resolution monitoring prototype is consist of two ADC channels with different gains, a 488.2 kHz sampling frequency, a 16.7 milliseconds sampling time interval, zero dead time and two counters. In order to improve the accuracy of measurement, the front part of the digital circuit includes two ADC channels with no hand-shake during data transfer, which makes the designed FPGA digital monitoring system achieve the criteria of zero dead time and full scale of reactor power measurement successfully, almost same as the original analog monitoring system. Compared with the dead time more than 96% by using the oscilloscope and more than 80% by using the Photon II data acquisition card, wave observation through the FPGA based digital system prototype is better owing to its characteristic of zero dead time. It is proved that the real time data calculated by the self-developed digital system is approximately equal to the result from the analog system.
In the study, a real time reactivity meter is also established according to the dynamic property of nuclear reactor, which works normally in THOR daily operation. Campbell theorem is applied to the real time power data calculation in the front digital circuit to realize the digital Campbell algorithm, which is compiled by the FPGA commercial board. On the other hand, the reactivity meter is accomplished by the phenomena of point kinetics and delayed neutron precursors in nuclear reactor dynamics, which means the reactivity can be obtained by the variation of the power data versus time under appropriate sampling time interval and known operating parameters. Moreover, a verification is compiled in the program to make a comparison of the calculated data and the data from the original monitoring system, and it can enhance the safety of reactor operation.
In addition, an intuitive human-machine interface is developed for the digital wide-range neutron monitoring system by LabVIEW belongs to National Instruments. With dynamic history graph, the display panel can provide operators more information of a nuclear reactor. It is observed that the coefficient of variance (COV) between the data smoothed by 60-points moving average and the output power data of the analog system is up to 0.9992. Similarly, the period calculated by 60-point least square fitting is also matched with the data collected by the original NM-1000 system. It is concluded that not only the precision but also the system response time of the self-developed digital monitoring system has a good agreement with the original analog one installed in THOR.

摘要 i
ABSTRACT iii
誌謝 vi
第 1 章 緒論 1
1.1研究動機 2
1.2未來展望 4
第 2 章 文獻回顧 5
2.1 Campbell定理應用於中子流量估算技術 5
2.1.1 Campbell定理數學式 6
2.1.2應用Campbell 定理的Fission chamber特性 8
2.1.2.1電荷產生(Charge generation) 8
2.1.2.2頻寬(Bandwidth) 10
2.1.2.3分裂腔的飽和(Saturation of fission chamber) 11
2.1.3 Campbell訊號大小(Signal size) 12
2.1.4 Campbell system統計準確度(Statistical accuracy) 12
2.1.5 Campbell system時間響應(Time response) 13
2.1.6高階Campbell 定理衍伸 14
2.1.7使用機率與統計方法驗證高階Campbell定理 15
2.1.7.1二階Campbell定理驗證 15
2.1.7.2三階Campbell定理驗證 16
2.1.7.3四階Campbell定理驗證 17
2.1.7.4五階Campbell定理驗證 18
2.2 GA NM-1000 WRNM 之Campbell 通道架構 19
2.2.1前置放大器模組 20
2.2.2 Campbell模組 21
2.2.3計數與傳送模組 22
2.2.4微電腦模組 22
2.3 GA NM-1000 WRNM 規格總結 22
2.4 GENE NUMAC WRNM 之Campbell 通道架構 24
2.4.1前置放大器 25
2.4.2鑑別器模組 27
2.4.3均方值模組 29
2.4.4類比模組 30
2.4.5微電腦模組 30
2.5 French MARINE之數位WRNM架構簡介 30
2.6 WRNM架構之規格綜合比較 32
2.7核反應器動態特性應用於反應度量測 33
2.7.1點反應器動態定理(Point Kinetics) 33
2.7.2點反應器動態方程式(In-hour Equation)應用 36
2.7.3 Doppler效應對於反應器的影響 38
2.7.4 CAMAC反應度量測系統 40
2.7.5 TRR反應度量測系統 43
第 3 章 即時數位監測系統架構 45
3.1前置訊號處理單元 45
3.2資料擷取硬體實現 46
3.3軟體演算與人機介面建置 47
3.4即時監測系統原型建構 49
3.4.1系統概要 49
3.4.2硬體環境 49
3.4.3前端電路 51
3.4.3.1電壓偏移電路(Voltage Shifter) 51
3.4.3.2 ADC模組 52
3.4.4演算流程 54
3.4.5資料傳遞 55
3.4.6硬體傳輸介面 56
3.4.7週期計算方法 57
3.5數位系統校正 59
第 4 章 反應度量測系統 65
4.1理論模型 65
4.2演算流程 69
4.3驗證機制 70
4.4驗證過程 70
第 5 章 實驗結果與分析 79
5.1量測架設 79
5.2數位系統線下測試 80
5.3線上反應器功率量測與分析 82
5.3.1數位與類比通道比較 83
5.3.2精確度評估 84
5.3.3取樣區段結合 86
5.3.4功率線性度比較 88
5.4線上反應器週期與核反應度量測分析 92
5.4.1反應度計算的波動狀況分析 97
5.4.2由核反應器動態特性分析反應度計算結果 108
5.4.3反應器週期計算吻合度探討 119
5.5驗證結果比較 121
5.5.1探討Inhour Equation的適用範圍 124
5.6數位監測系統原型與其他系統比較 125
第 6 章 結論與建議 126
第 7 章 參考文獻 128
第 8 章 附錄 137
8.1 NM-1000偵測系統區塊圖 137
8.2 Campbell模組電路 138
8.3 PA-15前置放大器電路 142
8.4計數/傳送模組電路 143


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