531758 A7 B7 五、發明説明（1 ) 發明背景 大致地’本發明有關核反應器且更特別地有關用於增加 沸水反應器之熱功率輸出的方法。 典型之沸水反應器(BWR)包含一壓力容器，此壓力容器 容納有-浸入於可從核燃料移去熱量之循環冷水中的核燃 料心’水會沸騰而產生蒸汽以用於驅動產生電力之蒸汽渦 輪發電機，然*蒸汽會凝結且水會在閉合迴路系統中送回 壓力容器’管路載送蒸汽至涡輪機及載送再循環水或供給 水回到容納有核燃料之壓力容器。 BWR包含若干傳統之閉合迴路控制系統，其控制bwr之 不同的個別操作以響應需求，例如控制棒傳動控制系統 (CRDCS)控制反應器核心内之控制棒且藉此控制該核心内 之棒密度而確定其中之反應性且依序地確定反應器核心之 輸出功率。再循環流量控制系統（RFCS)控制該核心之流率 ’其改變該核心中之蒸汽/水之關係且可使用於改變反應器 核心之輸出功率。在任一所給定之時間點處，該兩控制系 統會相互結合作業以控制反應器核心之輸出功率。渦輪機 控制系統（tcs)依據壓力調整或負荷需求而控制來自BWR之 蒸汽流量到渦輪機。 該等系統以及其他BWR控制系統之操作係利用BWR之不 同的監視參數予以控制，若干監視參數包含··會受到RFCS 影響之核心流量及流率；反應器系統壓力，其係從壓力容 裔排放到渦輪機之蒸汽壓力，該蒸汽壓力可測量於反應器 頂部或測量於至渦輪機之入口；中子通量或核心功率；供531758 A7 B7 V. Description of the invention (1) Background of the invention Generally, the invention relates to a nuclear reactor and more particularly to a method for increasing the thermal power output of a boiling water reactor. A typical boiling water reactor (BWR) contains a pressure vessel containing a nuclear fuel core immersed in circulating cold water that can remove heat from nuclear fuel. The water will boil to generate steam for driving a steam turbine that generates electricity. Generators, but the steam will condense and the water will be returned to the pressure vessel in a closed loop system. The pipeline carries steam to the turbine and carries recirculated or supplied water back to the pressure vessel containing nuclear fuel. BWR includes several traditional closed-loop control systems that control different individual operations of bwr in response to demand, such as a control rod drive control system (CRDCS) that controls the control rods in the reactor core and thereby controls the rod density in the core. Determine the reactivity among them and sequentially determine the output power of the reactor core. A recirculation flow control system (RFCS) controls the flow rate of the core, which changes the steam / water relationship in the core and can be used to change the output power of the reactor core. At any given point in time, the two control systems work in conjunction with each other to control the output power of the reactor core. The Turbine Control System (tcs) controls steam flow from the BWR to the turbine based on pressure adjustment or load demand. The operation of these systems and other BWR control systems is controlled using different monitoring parameters of BWR. Some of the monitoring parameters include the core flow rate and flow rate that will be affected by RFCS; the pressure of the reactor system, which is discharged from the pressure-bearing population The steam pressure to the turbine, which can be measured at the top of the reactor or at the inlet to the turbine; neutron flux or core power;
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給水之溫度及流率；配置於渦輪機之蒸汽流率以及證系 統之不同狀態指示。許多監視參數係直接地測量，而諸如 核心熱功率之其他參數則利用所測量之參數予以計算。來 自感測器之輸出以及所計算之參數係輪入於緊急保護系統 以癌保電薇之安全停機’視需要@隔離反應器於外部周遭 及防止反應器核心於任一緊急事件之期間以免於過熱。 先前地’反應n係設計操作於—比所允許之額定熱功率 準佔更高的熱功率輸出4。為符合規定之特許要領，反應 器係操作於比該反應器所能獲得之最大熱功率輸出更小的 最大熱功率輸出處。該等原始之設計基礎包含分解為設計 因數之大的保留餘地’而在操作數年之後已發現到核子反 應器可安全地操作於比原始所允許之更高的熱功率輸出準 位處。’口亦已確定對於操作參數之改變及/或裝備修正將允許 反應器在相當高的最大熱功率輸出處（直到且在原始所允許 功之120%之上）之安全操作。 為刼作於比核能規定者所允許之額定熱輸出更高的熱功 率輸出處品要該核能規定者所批准之特許修正。典型地 ，在批准取得自核能規定者之前，需要所提出之新的操作 參數處之核反應器的安全分析。 發明概述 在一代表性之實施例中，提供一種允許沸水核反應器增 加功率輸出之电知化方法，該方法包含：從屬性評估之資 料庫選擇屬性安全評估；比較增加功率輸出時之反應器操 作條件與所選擇屬性評估之反應器操作條件；確認屬性評 -5- 本紙張尺度適用中國國家標準(CNS) A4規格(210 X 297公爱) ' ----Feed water temperature and flow rate; steam flow rate configured in the turbine and different status indications of the certificate system. Many monitoring parameters are measured directly, while other parameters such as core thermal power are calculated using the measured parameters. The output from the sensor and the calculated parameters are rounded into the emergency protection system to ensure the safe shutdown of the cancer cell. As needed @Isolate the reactor from the outside and prevent the reactor core from being in any emergency period to avoid overheat. The previous 'reaction n' was designed to operate at a higher thermal power output 4 than the allowable rated thermal power. To comply with the required concession, the reactor is operated at a maximum thermal power output that is less than the maximum thermal power output that the reactor can obtain. These original design foundations include a large reserve factor decomposed into design factors, and after years of operation, it has been found that the nuclear reactor can safely operate at a higher thermal power output level than originally allowed. It has also been determined that changes to operating parameters and / or equipment modifications will allow the reactor to operate safely at a relatively high maximum thermal power output (up to and above 120% of the original allowable work). In order to operate at a higher thermal power output than the rated thermal output allowed by the nuclear energy regulator, a concession amendment approved by the nuclear energy regulator is required. Typically, the safety analysis of the nuclear reactor at the proposed new operating parameters is required before approval to obtain self-nuclear energy regulations. SUMMARY OF THE INVENTION In a representative embodiment, an electro-chemical method is provided that allows a boiling water nuclear reactor to increase power output. The method includes: selecting an attribute safety assessment from a database of attribute assessments; comparing reactor operations when power output is increased Conditions and reactor operating conditions for selected attribute evaluation; confirmation of attribute evaluation-5- This paper size applies to China National Standard (CNS) A4 specification (210 X 297 public love) '----
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線 531758 五、發明説明（3 ) 估之可應用性；以及執 . 包廠评估於所選擇屬性評估及未 包含在屬性評估資料庫中 不 作條件處。 子估之條件外之反應器操 在另一代表性的實施例中， ^ , x, ^ ^ r 杈供種允許沸水核反應器 # 、， 糸流匕3 · 一電腦，建構以模擬 該核反應在增加功率於 革輪出日才之操作及響應，從屬性評估 之舅料庫選擇屬性安+ 士 平估，比較增加功率輸出時之反應 器操作條件與所選㈣性評估之反應ϋ操作條件，確認屬 之可應用丨生’以及執行電廠特定評估於所選擇屬性 评估及未包含在屬性評仕^咨 蜀r wf估貝枓庫中之安全評估之條件外之 反應器操作條件處。 圖式簡單說明 圖1係3有渦輪發電機及沸水核反應器之電力產生系 統之基本組件的概略圖。 圖2係額定熱功率相對核心流量之百分比圖形，描繪第ι 圖中所不之沸水核反應器之擴充的操作域及功率額定值提 1¾ ;以及 圖3係根據本發明實施例之促成第1圖中所示之沸水核反 應器增加功率輸出之電腦控制安全分析的流程圖。 發明之詳細說明 圖1係電力產生系統8之基本組件的概略圖。該系統包含 一含有反應器核心12之沸水核反應器10，水14係利用反應 器核心12之熱功率予以沸騰，通過水-蒸汽相16而變成蒸汽 18，蒸汽18流過蒸汽流動路徑20中之管路到渦輪機流動控 裝 訂 -6- 本纸張尺度適用中國國家標準(CNS) A4規格(210X 297公釐） 五、發明説明（4 ) 制閥22，該閥22將控制進入蒸汽漏輪機24之蒸汽量，蒸汽 18係使用於驅動渴輪機24，該渦輪機24則依序地驅動發電 機26而產生電力，蒸汽18流到冷凝器28而轉換回水14，水 14藉供水泵30抽取而穿過供水路徑32中之管路回到反應器 10 〇 反應裔10之操作域40係藉圖2中所描繪之反應器熱功率 及心W里之圖予以特徵化。典型地，係允許反應器操作於 100%原始額定熱功率及100%額定心流量所界定之操作點44 所特徵化之流量控制/棒線42處或下方。在若干情況中，係 允σ午反應0操作於更大的域’但限制操作於1GG%原始額定 熱功率及75%額定心流量所界定之操作點48所特徵化之流 量控制/棒線46處或下方。 企望於操作在大於丨〇 〇 %原始額定所允許熱功率之有時候 %為功率額定值提高的熱功率處，線5〇代表操作域4〇潛在 的上方界限。為了操作於操作域40之額定值提高地區中， 需要操作條件及/或裝備修正，最適化之功率額定值提高係 根據電廠實際能力及電廠擁有者/操作員之財政目標予以界 定。 圖3係根據本發明實施例之促成沸水反應器〗〇增加功率 輸=之電腦控制安全分析法6〇之流程圖。為獲得功率額定 值提高之允許的修正，用之擁有者需提出電廠特定 之功率額定值提高的安全評估報造給適當的核能規定者， 該報告須詳述在較高功率輸出準佔處執行於電廠安全操作 之表現中的種種技術分析。在安全報告之檢討週期期間可 五、發明説明（5 ) 具有若干來自於核能規定者之額外資訊的要求，該額外資 訊涉及BWR使用之擁有者及其立約者針對於解決之時間及 努力的資訊。一旦安全報告之檢討合乎要求時，則由核能 規定者核准該電廠操作執照之修正以回應更新心核心的熱 功率條件。該執照修正之請求應與適用於—般執照之考量 -致。尤其’並未在電廠之允許基礎上有何改變，且預期 在電U施所發出之廢水或輕射量之上並未因為功率更新 而有效地增加。潛在重大危險之考量在於根據所提出之修 正的設施操作不可涉及先前所評估之意外事件的可能性或 ^ 之胃加不可產生新的或不同種類之意外事件於先 前所評估之任一意外事件的可能性，或不可涉及安全界限 之重大的減少。 方法60包含從已執行於功率額定值條件處之屬性安全詞 ^的資料料擇㈣腦為主的隸安全評估；比較料電處 6又计組態與屬性評估中所假 ^作又疋之電廠特徵的範圍；以及择 認66屬性安全評估之 〇 應用性於特定之電廠應用。方法6< 亦包含執行68特定評估於所 ,^ ^ 遇擇屬〖生评估之應用範圍外或 未包含於屬性評估資料庫中 φ . ^ Τ之反應裔刼作條件，若干該等 電廠特定評估係以根據獲 m 又付目屬性砰估之結果的簡易方式 安全呼估幹…： 擇之屬性安全評估及特定 :二：二報告資料庫中所健存之允許報告板 報止雷早抝目丨^ °用於美交到核能規定者。允許 報口電子板則女置以響應夾 率宏## 一 〜來自核此規定者之用於類似之功 羊額疋值如南之所提出的問題。 _ -8- 本纸張尺度適财g國家標準 531758 A7 B7 五、發明説明（6 ) 典型地，該屬性安全評估已由適當的核能規定者予以檢 視及核准，藉顯示反應器10之操作條件於預先核准之屬性 評估的範圍内可消除需再評估反應器10於增加之核心熱功 率輸出處以用於該屬性評估所涵蓋的條件。詳細之特定評 估僅執行於屬性評估之邊際條件外之條件而簡化了電廠特 定評估以及整個允許程序。 方法60包含評估72增加功率輸出時之核心及燃料性能， 該等評估提供了在正常之穩態操作，預期之操作事件或意 外事件期間之熱及機械整體性之預測，該等評估亦解釋了 電廠操作策略，操作循環的長度以及諸如具有宣告無法運 行之特定裝備或具有降級性能輸出之裝備操作的偶發性模 式。 評估72增加功率輸出時之核心及燃料性能的影響包含確 定74限制了增加之核心熱功率輸出所預期的暫態現象而無 故障停車（ATWS)事件。若干（ATWS)事件包含主蒸汽隔離閥 關閉（MSIVC);壓力調節器故障開啟（PRFO);離該處之功率 耗損（LOOP);以及減壓閥不經意的開啟（IORV)。該分析考 慮到ATWS緩和之特性，諸如再循環泵運行（RPT)，替換棒 之插入（ARI)，及備用液體控制系統（SLCS)之性能，而產生 重要參數之圖以及計算中子通量之峰值，平均之燃料熱通 量及容器壓力以用於各該4個事件，以比較用於增加核心 熱功率輸出所確定的ATWS事件與屬性評估之ATWS事件。 方法60亦包含評估76功率額定值提高條件時之反應器壓 力容器（RPV)内部及外部之系統，結構及組件（SSC)之機械 -9- 本紙張尺度適用中國國家標準(CNS) A4規格(210 X 297公釐） 531758 、發明説明（ 及、σ構的凡整性’包含來自增加溫度’流量，壓力及輻射 的效應’該等SSCs必須在動態負荷或振動效應下維持其社 構之完整性及執行其原先所打算之功能，諸如壓力邊= 分或核心冷却幾何形狀成分。Line 531758 V. Description of the invention (3) The applicability of the assessment; and the implementation of the contractor's assessment in the selected attribute assessment and not included in the attribute assessment database without conditions. In a representative embodiment, the reactor operating outside the conditions described in the example, ^, x, ^ ^ r is provided to allow boiling water nuclear reactor # ,, 糸 流 dagger 3. A computer, constructed to simulate the nuclear reaction in the Increase the power to the operation and response of the genius before it is released, select the property security + spine estimate from the property evaluation library, and compare the reactor operating conditions when the power output is increased with the selected response evaluation and operating conditions. Recognising applicable applications and performing plant-specific assessments at reactor operating conditions outside of selected attribute assessments and conditions not included in the safety assessments in the safety assessment. Brief Description of Drawings Figure 1 is a schematic diagram of the basic components of a power generation system with a turbine generator and a boiling water nuclear reactor. Figure 2 is a graph of the percentage of rated thermal power relative to the core flow rate, depicting the extended operating range and power rating of the boiling water nuclear reactor not shown in Figure ι; and Figure 3 is the result of the first embodiment of the present invention. Flow chart of computer-controlled safety analysis for increasing power output of boiling water nuclear reactor shown in the figure. Detailed description of the invention Fig. 1 is a schematic diagram of the basic components of a power generation system 8. The system includes a boiling water nuclear reactor 10 containing a reactor core 12. The water 14 is boiled by the thermal power of the reactor core 12 and passes through the water-steam phase 16 to become steam 18. The steam 18 flows through the steam flow path 20. Pipe-to-turbine flow control binding-6- This paper size applies to China National Standard (CNS) A4 (210X 297 mm) V. Description of the invention (4) Valve 22, which will control the steam leakage into the turbine 24 The amount of steam, steam 18 is used to drive thirsty turbine 24, which in turn drives generator 26 to generate electricity. Steam 18 flows to condenser 28 and is converted back to water 14, which is extracted by water pump 30. The operating region 40 of the reactor 10 is returned to the reactor 10 through the pipeline in the water supply path 32. The operation area 40 of the reactor 10 is characterized by the graph of the reactor thermal power and the core figure depicted in FIG. Typically, the reactor is allowed to operate at or below a flow control / rod line characterized by an operating point 44 defined by 100% of the original rated thermal power and 100% of the rated cardiac flow. In some cases, the σ noon response is allowed to operate in a larger domain, but limited to 1GG% of the original rated thermal power and 75% of the rated cardiac flow defined by the operating point 48 of the flow control / rod line 46 Below or below. It is expected that the operation will be performed at a temperature greater than the allowable thermal power of the original rating, sometimes% is the thermal power with the power rating increased, and the line 50 represents the potential upper limit of the operating domain 40. In order to operate in the rated area of the operating area 40, the operating conditions and / or equipment needs to be modified. The optimal power rating increase is defined based on the actual capacity of the power plant and the financial goals of the plant owner / operator. FIG. 3 is a flow chart of a computer-controlled safety analysis method 60 for increasing the power input and the boiling water reactor according to an embodiment of the present invention. In order to obtain an allowable correction for the increase in power rating, the owner must submit a safety assessment report of the increase in power rating specific to the power plant to the appropriate nuclear energy regulations. Various technical analyses performed in the performance of safe operation of power plants. During the review cycle of the safety report, the invention description (5) has a request for additional information from the nuclear energy regulator, which relates to the information used by the owner of the BWR and its contractor on the time and effort for resolution. . Once the review of the safety report is satisfactory, nuclear power regulators will approve amendments to the plant's operating license in response to the renewed core thermal power conditions. The request for amendment of the license shall be in accordance with the considerations applicable to the general license. In particular, there has been no change on the basis of the allowance of the power plant, and it is expected that the amount of wastewater or light emission from the power plant is not effectively increased due to the power update. Potentially significant hazard considerations are based on the possibility that the operation of the facility under the proposed amendment must not involve the possibility of a previously assessed contingency or that a new or different type of contingency could not produce any of the previously assessed contingencies. The possibility may not involve a significant reduction in safety margins. Method 60 includes selecting a brain-based slave safety evaluation from the data of the attribute safety word ^ that has been performed at the power rating condition; comparing the material and electricity division 6 with the calculations in the configuration and attribute evaluation ^ The scope of power plant characteristics; and the applicability of 66 attribute safety assessments to specific power plant applications. Method 6 < also includes performing 68 specific assessments at the site, ^ ^ case is a reaction condition of φ. ^ T outside the scope of application of the health assessment or not included in the attribute assessment database, and some of these plant-specific assessments It is a simple way to securely evaluate the results based on the results of the attribute evaluation of m and pay ....: Optional attribute security assessment and specific: two: two allowable report boards stored in the report database to report thunder丨 ^ ° Used by the United States to transfer to nuclear energy regulations. The newspaper electronic board is allowed to be set up in response to the clip rate macro ## 1 ~ From the issue of verifying this rule for similar functions The value of the sheep's forehead is raised by Nan Zhi. _ -8- This paper is a national standard 531758 A7 B7 V. Description of the invention (6) Typically, the property safety assessment has been reviewed and approved by the appropriate nuclear energy regulations, and the operating conditions of the reactor 10 are shown Within the scope of the pre-approved attribute evaluation, the need to re-evaluate the reactor 10 at the increased core thermal power output for the conditions covered by the attribute evaluation can be eliminated. The detailed specific evaluation is performed only on conditions outside the marginal conditions of the attribute evaluation, which simplifies the specific evaluation of the power plant and the entire permitting process. Method 60 includes an assessment 72 of core and fuel performance as power output is increased. These assessments provide predictions of thermal and mechanical integrity during normal steady-state operation, anticipated operational events or unexpected events, and these assessments also explain Power plant operating strategy, length of operating cycles, and occasional modes of operation such as equipment with specific equipment declared to be inoperable or equipment with degraded performance output. Assessing the impact of 72 core and fuel performance when increasing power output includes determining 74 the transient phenomena that are expected to limit the increased core thermal power output without a fault stop (ATWS) event. Several (ATWS) events include main steam isolation valve closing (MSIVC); pressure regulator failure opening (PRFO); power loss (LOOP) from there; and inadvertent opening of the pressure reducing valve (IORV). This analysis takes into account the characteristics of ATWS mitigation, such as recirculation pump operation (RPT), the replacement rod insertion (ARI), and the performance of the standby liquid control system (SLCS), resulting in graphs of important parameters and calculation of neutron flux The peak, average fuel heat flux, and container pressure are used for each of the four events to compare the ATWS event determined to increase core thermal power output and the ATWS event for attribute evaluation. Method 60 also includes assessing the internal and external systems, structures, and components (SSC) of the reactor pressure vessel (RPV) when the 76 power rating is increased. 9- This paper is sized to the Chinese National Standard (CNS) A4 (210 X 297 mm) 531758, description of the invention (and, the normality of the σ structure 'includes the effects of increasing temperature' flow, pressure, and radiation 'These SSCs must maintain their social structure under dynamic loading or vibration effects Completeness and perform its intended function, such as pressure edge = minutes or core cooling geometry components.
」列如RPV内部之SSCs包含核心護罩，核心支撐板，反應 為核心的頂部導口，及蒸汽乾燥器’例如RPV外部之SSCs 包含生物屏壁，管路/間/果系統，及圍 S =冓的完整性，發展出RPV及内部組件之電礙 %月®為主的；^型’而且經由電腦模擬發展出電廢熱水力初 始條件以用於穩態以及暫態及意外情況，計算及比較肌 2所產生之負荷於特定的設計規範以確定在穩態或意設相 个月況下之SSCs機械的完整性。 一 亦包含評估78安全裝備性能之能力以維持電廠於 ’·’、工制之狀態中及在預期之操作性發生或意外事件之期 門使任董十於大眾健康及安全之不利影響最小化。該等評 ::依據原始之系統設計’目前之系統操作資料以及具有 宣告無法運行或降級性能之所選擇裝備操作的贿性模式。 評㈣安全裝備性能包含計算8G反應器核心隔離冷却李 ”:(咖)預防核心免於在供水漏失事件期間未覆蓋之核心 功率的範圍。職系統之主要目的在於維持足夠的冷却劑 於反應器容器中’使得該核心不會 在，爪自反應器供水系統 2冷却劑漏失所伴隨之反應器隔離事件中未予以覆蓋，此 事件係限制之暫態，其將挑戰核心之冷却。 伴Ik者功率額定值提高之更高的核心功率準位將造成反 j__ -10- 二張尺度適财_家標準(CN—(21GX297公复)- 裝 訂 i 應器容器中更多的蒸發以及更低的水準位而 ，蓋的潛在性。該觀系統應提供足夠的補給，使得:應 器核心保持覆蓋有水，直到獲得穩定的情況。 此外，RCIC系統應提供足夠的冷却劑補給 降流管中之水位保持在活性燃料 心σσ 〈了貞σ卩上方。若降流管水 位洛到活性燃料之頂部下方時，則 W緊急&序導弓I線會指示 #作貝減壓該容器及使用低壓緊急核心冷却系統邮⑼來 恢復核心之冷却。此動作之進行並非企望的，因為會造成 超過所建議之容器減壓速率。 為確認提高額定之功率操作期間持續施加之安定性正確 的動作及描繪提高額定之操作於特定長期溶液上之影響， 方法60包含確定82增加之核心功率輸出操作期間安=期 間正確的動作。 進一步地，方法60包含評估84增加之功率輸出操作時之 反應器控制及儀錶系統，由熱功率，蒸汽流量，操作壓力 及輻射中之增加所影響之儀錶設定點係初始地再計算為分 析之界限（ALs)，諸如準確性，漂移及延遲之裝備特定的特 徵係成ALs中之因數而接著轉換為實際的儀錶設定點。 為顯示反應器10之操作係在預核准之屬性評估的包封之 内，方法60包含計算86反應器設定點於增加之功率輸出操 作條件時以確保安全的電廠操作於功率提高額定值之條件 時，用於直接地伴隨安全分析報告（SAR)中所分析之不正常 包礙暫悲或意外事件所感測參數之設定點的決定係依據建 立為部分安全分析之分析界限（AL)，該分析之界限係在初 -11- 531758 A7 ----------B7 五、發明説明（9 ) 始所企望之動作前或當時所感測之過程變數值，該AL係設 笔薇〖生把分析所確認者，使得不會超過適用之允許安 全界限，此分析會考慮到儀錶響應時間，暫態超時及模型 準確性。 ' 當由於功率提高額定值而作成AL之改變時，必須建立新 的可。午可值（AV)，AV係藉提供許可於特定的或預期的校準 食匕力儀器之準確性及過程測量誤差而確定自AL·，然後此 值曰界疋為參數之技術規格（Tech Spec)界限且指定為電廠之 允許條件。 標稱之運行設定點（NTSP)值係藉考慮除了儀錶準確性， 杈準及過程測量誤差之外的儀錶飄移而計算自ALs。八1與 A V間之差異允許用於頻道儀錶準確性，校準準確性，過程 測里準確性，及主要元件準確性。△▽與1^丁8?間的邊限允許 用於會發生於所建立之監視週期期間的儀錶飄移。若，在 監視週期之期間，儀錶設定點飄移在非保守的方向之中時但 亚未超過AV，則儀錶性能仍在電廠安全分析之要求之内。 並非所有參數均具有根據安全分析之相關連的AL(例如 主蒸汽管線輻射監視器），AV或設計基礎之技術規袼界限 可直接地根據電廠允許要件，先前操作經驗或其他適合之 規範予以界定，然後從AV計算NTSp而允許儀錶飄移。當適 合日^，NTSP可直接地根據操作經驗或工程判斷予以確定。 方法60亦包含輸出88資料而在功率提高額定值操作的支 援中促成電廠文獻更新，所輸出之資料作用為促成現場操 作程序，工程圖式及計#，設計基礎文獻，以及包含電廠"SSCs inside RPV include core shields, core support plates, top guides that react as cores, and steam dryers." For example, SSCs outside RPVs include biological barriers, pipes / chambers / fruit systems, and enclosures. = Integrity of 冓, developed the RPV and the internal components of the interference %% ®; ^ type ', and through computer simulation to develop the initial conditions of electric waste hot water power for steady state and transient and unexpected conditions, calculation and The load generated by Muscle 2 is compared to specific design specifications to determine the mechanical integrity of SSCs under steady-state or intended phase conditions. One also includes the ability to assess the performance of 78 safety equipment to maintain the power plant in the state of '·', the system of work, and the expected operational occurrence or unexpected period to minimize the adverse impact of Ren Dongshi on public health and safety . These reviews :: Based on the original system design ’current system operation data and a bribery model of selected equipment operations with declared inoperability or degraded performance. Evaluation of safety equipment performance includes calculation of 8G reactor core isolation cooling plume ": (Ca) Prevents the core from being covered by core power that was not covered during the water supply leakage event. The main purpose of the system is to maintain sufficient coolant in the reactor 'In the container' so that the core will not be in the reactor isolation event accompanying the loss of coolant from the reactor water supply system 2 is not covered. This event is a limited transient that will challenge the cooling of the core. Higher core power levels with higher power ratings will result in inverse j__ -10- two-size-fits-all financial standards (CN— (21GX297))-more evaporation and lower The potential of the cover. The system should provide sufficient recharge so that: the core of the reactor remains covered with water until a stable condition is obtained. In addition, the RCIC system should provide sufficient coolant to replenish the downcomer. The water level stays above the active fuel center σσ <贞 σ 卩. If the downcomer water level falls below the top of the active fuel, the emergency & sequence guide bow I line will instruct # 作 贝 compress the Device and use a low-pressure emergency core cooling system to restore core cooling. This operation is not expected because it will exceed the recommended container decompression rate. In order to confirm that the rated power is continuously applied, the stability is correct during operation The method 60 includes determining the effect of increasing the rated operation on a particular long-term solution. Method 60 includes determining 82 the correct core power output operation during the period = the correct action. Further, method 60 includes evaluating 84 the increased power output during operation. Reactor control and instrumentation systems. Instrument setpoints affected by increases in thermal power, steam flow, operating pressure, and radiation are initially recalculated as analytical limits (ALs), such as equipment specificity for accuracy, drift, and delay. The characteristics are converted into factors in ALs and then converted to actual instrument set points. To show that the operation of reactor 10 is within the envelope of a pre-approved attribute evaluation, method 60 includes calculating 86 reactor set points in increments. Power output operating conditions to ensure safe power plant operation at elevated power ratings The determination of the set point used to directly accompany the abnormality analyzed in the safety analysis report (SAR) to the sensed parameters of temporary or unexpected events is based on the analysis limit (AL) established as part of the safety analysis. The limit is at the beginning of -11- 531758 A7 ---------- B7 V. Description of the invention (9) The value of the process variable is sensed before the desired action or at the time. The analysis shall be performed so as not to exceed the applicable allowable safety limit. This analysis will take into account the instrument response time, transient timeout, and model accuracy. 'When the AL is changed due to the increased power rating, A new value must be established. AV is available. AV is determined from AL · by providing the accuracy and process measurement error of a specific or expected calibrated food force instrument, and then this value is defined as a parameter. Technical Spec (Tech Spec) limits and specified as permissible conditions for power plants. The nominal operating set point (NTSP) value is calculated from ALs by considering instrument drift in addition to instrument accuracy, accuracy, and process measurement errors. The difference between 8 and A V allows for channel meter accuracy, calibration accuracy, process measurement accuracy, and main component accuracy. The margin between △ ▽ and 1 ^ 丁 8? Is allowed for instrument drift that will occur during the established monitoring cycle. If, during the monitoring period, the instrument setpoint drifts in a non-conservative direction but does not exceed AV, the performance of the instrument is still within the requirements of the power plant safety analysis. Not all parameters have an associated AL based on safety analysis (such as the main steam line radiation monitor). The technical or regulatory limits of AV or design foundation can be defined directly based on permitted elements of the power plant, previous operating experience, or other suitable specifications. , And then calculate the NTSp from the AV to allow the meter to drift. When applicable, NTSP can be determined directly based on operating experience or engineering judgment. Method 60 also includes outputting 88 data to facilitate power plant documentation updates in support of power-up rating operations. The output data is used to facilitate field operations, engineering drawings and calculations, design basics, and includes power plants.
裝 訂Binding
線 -12- 五、發明説明（1〇 ) 模擬器之訓練程式的更新。 為評估所增加之功率輸出在電廠緊急操 ，方法60包含計算90界定當需要操作員 的政應 片写Μ下時之變數 制曲線。在電廠緊急操作程序中之操 " ^ ^ ^ ^ ^ σσ 貝勤作亚未依增加 額疋之反應器功率之結果而改變，僅改 又石' 卞所特定之動 作處之條件，再計算之範疇則根據伴隨 ^ ^ 手徒南額定值 之電廠改變的大小，該等再計算係'包含於下述類目中·. I ·僅改變額定的反應器功率。 π ·改變除了額定的反應器功率外 手外之取低的安全/保險閥 提升壓力設定點。 瓜·改變除了額定的反應器功率外之圍堵體操作溫度。 IV·改變除了額定的反應器功率外之燃料形成，但：的燃 料具有相同的峰值線性熱產生速率及相同的燃料棒尺寸。 V·改變除了額定的反應器功率外之燃料形式，以及新的 燃料具有不同的峰值線性熱產生速率及/或燃料棒尺寸。 該等類目包含所有相結合於會影響電廠緊急操作程序變 數及曲線之擴充性功額定值提高的預期改變，例如若功率 額定值提高會造成最低的安全/保險閥提升壓力設定點改變 且具有負荷之新的燃料形式時，則需檢驗類目π&Ιν(或 。然而’當界定電廠特定的額定值提高之程式時，將相對 於需用於電廠緊急操作程序計算之電廠資料來確認所影響 的電廠值以確保不會影響其他值。 進一步地，方法60包含計算92增加核心熱功率輸出時之 可能性危險評估以及比較該評估於屬性評估之可能性危險 -13- 本紙張尺度適用中國國家標準(CNS) Α4規格(210X297公爱)" "' - 2估。追求功率額定值提高之電廠預期請求其執照之修正 付合於管理其目前執照之考量，也就是說，在電廠允許基 礎上並沒有改變。修正涉及無重大危險（NSH)考慮，若根據 所提出之修正之設施的操作不會：涉及重大的增加於先前 所評估之意外事件之可能性或結果；產生先前所評估之任 何意外事件之新的或不同種類的意外事件的可能；或涉及 安全邊限中之重大減少。 保y接受之運行避免係在操作性暫態之期間提供於額定值 提咼之後而解決。 意外事件之可能性並不會藉功率額定值提高而有效地增 加’在操作壓力中之小的增加及在溫度中之小的增加並不 具有重大的效應於LOCA可能性之上。當需要時，意外事件 之先質及暫態之發生頻率係藉施加適合之設定點方法以確 在電廠危險上之功率額定值提高之影響的概括性評估係 藉檢視額定值提高之效應於個別之電廠檢驗（IpE)之上而獲 得，此包含該額定值提高在意外事件及其他事件上的效應 。大部分之核能電廠已藉執行可能性安全評估（psA)而完成 IPE，準位1 PSA使引起核心損壌之事件模型化以及計算核 心損壞頻率，準位2 PSA使核心融化之進展及圍堵體故障模 型化以及計算輻射釋出之頻率及大小。 在電廠IPE上之功率額定值提高之效應評估將考慮到功率 額定值提高在諸如：初始化事件頻率；成功規範；組件故 障率；及操作員動作及裝備恢復之有效時間的IpE輸入及假 設上的效率。 -14 - 本紙張尺度適用中國國家標準(CNS) A4規格(210 X 297公釐） 五、發明説明（12 ) 當作IPE之部分的實用性將識別相關連於核心損壤潛在性 及圍堵體u任何電廠之易受傷害性。在電薇ιρΕ上評估 力率額疋值提π之影響將足以識別由功率額定值提高所引 (之任何新的易又知害@，^識別出新的易受傷害性之時 ，則將在允許報告中報告它 <門，若識別出無新的易受傷害 性之時，則可結論ttj該功率額定值提高在電廠危險上具有 可忽視之影響。在並未增加f受傷害性之意外事件頻率或 有效增加核心損壌頻率中之改變本身係無意義的。 述方法60提Μ #系統化，預核准之方式以用於沸^ 反應器之使用擁有者/操作員而允許熱功率額定值提高及義 此使來自核能電廠操作之收益最大化。方法6()促成峨^ 用擁有者發展出最可靠及經證實之方式而以適時方式且符 合目前規定及允許要件來取得功率額定值提高的執昭修正 。標準化的過程福保了所有BWR功率額定錢高計劃中之 —致性且使整體方< 增加效率。從電力供應之觀點來看， 功率增加量可極有意義，例如在原始允許 20%。 … 干心上 雖然本發明已針對不同之特定實施例予以描述，作孰習 =本項技術之該等人士將理解的ϋ發明可隨申請專 範圍之精神及範轉内之修正來實施。 -15- 本紙張尺度適科肖料標準(CNS) Α4規格⑽χ297公爱) ------Line -12- V. Description of the Invention (10) Update of the training program of the simulator. In order to evaluate the increased power output during emergency operation of the power plant, method 60 includes calculating 90 defining the variable curve when the operator's response film is needed. Operations in the emergency operation procedure of the power plant " ^ ^ ^ ^ ^ σσ Beiqin Zuoya changed according to the result of increasing the reactor power, only changed the conditions of the specific operation place, and then calculated The category is based on the size of the power plant accompanying the rating of Handan South. These recalculations are 'included in the following categories ... I only change the rated reactor power. π • Change the safety / safety valve lift pressure set point to a lower value in addition to the rated reactor power. • Change the operating temperature of the enclosure in addition to the rated reactor power. IV. Change the fuel formation except for the rated reactor power, but: The fuel has the same peak linear heat generation rate and the same fuel rod size. V. Change the fuel form in addition to the rated reactor power, and the new fuel has different peak linear heat generation rates and / or fuel rod sizes. These categories include all anticipated changes that are combined with an increase in the expansive work rating that will affect the emergency operating process variables and curves of the power plant, such as a change in the minimum power / safety valve lift pressure set point if the power rating increase If there is a new form of fuel with a load, the category π & Ιν (or.) Is required. However, when defining the program for the specific rating increase of a power plant, it will be compared with the power plant information that is used for the calculation of the emergency operation procedure of the power plant. To determine the value of the affected power plant to ensure that it will not affect other values. Further, the method 60 includes calculating 92 a risk assessment of the likelihood of increasing the core thermal power output and comparing the risk of the assessment to the attribute assessment. The standard applies to China National Standard (CNS) Α4 specification (210X297 public love) " " '-2. Estimated power plants seeking to increase the power rating are expected to request amendments to their licenses to be considered in consideration of managing their current licenses, that is, It is said that there is no change based on the plant's permission. The amendment involves no significant hazard (NSH) considerations. No: it involves a significant increase in the likelihood or outcome of a previously assessed contingency; the possibility of a new or different kind of contingency resulting in any of the previously assessed contingencies; or a significant reduction in safety margins. Ensure that the accepted operation is not provided during the operational transient period after the rating is increased. The possibility of an accident will not be effectively increased by the increase of the power rating 'small in operating pressure The increase in temperature and the small increase in temperature do not have a significant effect on the likelihood of LOCA. When needed, the precursors of accidents and the frequency of transients are determined by applying appropriate setpoint methods to confirm A general assessment of the impact of increased power ratings on danger is obtained by examining the effects of increased ratings on individual power plant inspections (IpE), which includes the increase in ratings on unexpected and other events The majority of nuclear power plants have completed IPE by performing a Possibility Safety Assessment (psA). Level 1 PSA models events that cause core damage and calculates nuclear power. Cardiac damage frequency, level 2 PSA models the progress of core melting and the failure of containment bodies and calculates the frequency and magnitude of radiation release. The effect assessment of the increase in power rating on the IPE of a power plant will take into account the power rating Improve the efficiency of IpE input and assumptions such as: the frequency of initialization events; successful specifications; component failure rates; and the effective time of operator actions and equipment recovery. -14-This paper standard applies Chinese National Standard (CNS) A4 specifications ( 210 X 297 mm) 5. Description of the invention (12) The practicality of being part of IPE will identify the potential damage to the core soil and the vulnerability of any power plant. Evaluate on electricity The effect of increasing the power rate value by π will be sufficient to identify any new vulnerabilities and harms caused by the increase in power rating (@, ^ When new vulnerabilities are identified, they will be reported in the allowable report If it is recognized that there is no new vulnerability, it can be concluded that the increase of the power rating has a negligible effect on the danger of the power plant. Changes that do not increase the frequency of accidents that are harmful to f or that effectively increase the frequency of core damage are meaningless in themselves. The method 60 TiM # described above is a systematic, pre-approved way to allow the owner / operator of a boiling reactor to allow the thermal power rating to be increased and meaning to maximize the benefits from the operation of a nuclear power plant. Method 6 () enables E ^ to use the owner to develop the most reliable and proven method to achieve an ambitious amendment in a timely manner and in accordance with current regulations and permitted requirements to obtain an increase in power rating. The standardized process guarantees consistency in all BWR power rating schemes and makes the overall approach < increase efficiency. From the point of view of power supply, the amount of power increase can be extremely significant, such as 20% of the original allowance. … With care Although the present invention has been described for different specific embodiments, the practice = those skilled in the art will understand that the invention can be implemented with the spirit and scope of the application. -15- This paper is compliant with the Chinese Standard (CNS) Α4 Specification (⑽χ297 公 爱) ------