JP2001074875A - Method for obtaining temperature difference between inlet/outlet of reactor vessel of pressurized water reactor, and method for evaluating performance of pressurized water reactor plant - Google Patents

Method for obtaining temperature difference between inlet/outlet of reactor vessel of pressurized water reactor, and method for evaluating performance of pressurized water reactor plant

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Publication number
JP2001074875A
JP2001074875A JP25561199A JP25561199A JP2001074875A JP 2001074875 A JP2001074875 A JP 2001074875A JP 25561199 A JP25561199 A JP 25561199A JP 25561199 A JP25561199 A JP 25561199A JP 2001074875 A JP2001074875 A JP 2001074875A
Authority
JP
Japan
Prior art keywords
core
temperature
reactor
outlet
reactor vessel
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Withdrawn
Application number
JP25561199A
Other languages
Japanese (ja)
Inventor
Nobunori Fujitsuka
信典 藤塚
Noriyuki Onishi
宣幸 大西
Akihiko Tominaga
明彦 富永
Toshiaki Niihara
俊明 新原
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Shikoku Electric Power Co Inc
Mitsubishi Heavy Industries Ltd
Original Assignee
Shikoku Electric Power Co Inc
Mitsubishi Heavy Industries Ltd
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Shikoku Electric Power Co Inc, Mitsubishi Heavy Industries Ltd filed Critical Shikoku Electric Power Co Inc
Priority to JP25561199A priority Critical patent/JP2001074875A/en
Publication of JP2001074875A publication Critical patent/JP2001074875A/en
Withdrawn legal-status Critical Current

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Classifications

    • YGENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02EREDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00Energy generation of nuclear origin
    • Y02E30/30Nuclear fission reactors

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  • Monitoring And Testing Of Nuclear Reactors (AREA)

Abstract

PROBLEM TO BE SOLVED: To improve reliability when detecting the temperature difference between the inlet/outlet of the reactor vessel of a pressurized water reactor. SOLUTION: The correlation between a measurement reactor core outlet temperature distribution obtained by a detector 57 of temperature in the reactor being arranged at the outlet of a reactor core 47 in a reactor vessel 31 and a calculation reactor core outlet temperature distribution being calculated based on the reactor core output distribution obtained by a detector 55 of neutrons in the reactor is obtained, thus calculating the temperature of the outlet of the reactor core of a cooling material. Assuming that there is no heat transfer from the outlet of the reactor core to that of the reactor vessel, the calculated temperature is set to the temperature of the outlet of the reactor vessel.

Description

【発明の詳細な説明】DETAILED DESCRIPTION OF THE INVENTION

【0001】[0001]

【発明の属する技術分野】本発明は、原子炉、特に加圧
水型原子炉のプラント性能を評価する方法に関する。
[0001] The present invention relates to a method for evaluating the performance of a nuclear reactor, in particular, a pressurized water reactor.

【0002】[0002]

【従来の技術】加圧水型原子炉においては、原子炉冷却
材の原子炉容器出口温度と原子炉容器入口温度の差が運
転制御や性能評価の重要なパラメータとして使用されて
いる。これを概説すると、加圧水型原子力プラントの代
表的な系統図を示す図4において、原子炉容器1内の炉
心3で加熱された冷却材は、高温側配管即ちホットレグ
5を通って蒸気発生器7に入り、後述するような給水と
熱交換してこれを蒸気に変換せしめ、自身は低温になっ
て中間配管即ちクロスオーバレグ9を介して冷却材ポン
プ11に吸い込まれる。冷却材ポンプ11は、その降温
した冷却材を低温側配管即ちコールドレグ13を介して
原子炉容器1に戻し、冷却材は再び炉心3において加熱
される。このように冷却材が循環する系は一次冷却系と
も称されているが、一次冷却系の圧力は加圧器15によ
り所定範囲に制御されている。前述のように蒸気発生器
7内で原子炉冷却材により加熱されて発生された蒸気
は、主蒸気管21を介して蒸気タービン23に送られ、
ここで発電機を駆動して電気を発生する。蒸気タービン
23の排蒸気は復水器25において冷却、凝縮されて復
水となり、二次冷却材ポンプ乃至給水ポンプ27によ
り、再び給水として蒸気発生器7に供給され、二次冷却
材即ち給水は運転中前述の循環系路即ち二次冷却系を循
環する。
2. Description of the Related Art In a pressurized water reactor, a difference between a reactor vessel outlet temperature of a reactor coolant and a reactor vessel inlet temperature is used as an important parameter for operation control and performance evaluation. In summary, in FIG. 4, which shows a typical system diagram of a pressurized water nuclear power plant, a coolant heated in a core 3 in a reactor vessel 1 passes through a high-temperature side pipe or hot leg 5 and a steam generator 7. Then, it exchanges heat with the feed water as described later to convert it into steam, which itself is cooled down and sucked into the coolant pump 11 via the intermediate pipe or crossover leg 9. The coolant pump 11 returns the cooled coolant to the reactor vessel 1 through the low-temperature side pipe, that is, the cold leg 13, and the coolant is heated again in the core 3. The system in which the coolant circulates in this manner is also called a primary cooling system. The pressure of the primary cooling system is controlled by a pressurizer 15 to a predetermined range. As described above, the steam generated by being heated by the reactor coolant in the steam generator 7 is sent to the steam turbine 23 through the main steam pipe 21,
Here, the generator is driven to generate electricity. Exhaust steam of the steam turbine 23 is cooled and condensed in a condenser 25 to be condensed water, and is again supplied as water to the steam generator 7 by a secondary coolant pump to a water supply pump 27. During operation, the above-mentioned circulation system, that is, the secondary cooling system is circulated.

【0003】而して、蒸気発生器7において原子炉冷却
材から給水に与えられる熱量即ち原子炉熱出力、蒸気発
生器7内の蒸気の圧力即ち主蒸気圧力及び蒸気タービン
23に連結された発電機の電気出力が運転制御乃至性能
評価において主要な指標となっている。そして、原子炉
熱出力は、ホットレグ5に設けた温度検出器17による
冷却材の原子炉容器出口温度THOTとコールドレグ13
に設けた温度検出器19による冷却材の原子炉容器入口
温度TINとの差即ち原子炉容器出入口温度差ΔTと原子
炉冷却材の流量の積として求められる。そして原子炉容
器出入口温度差ΔTは、温度検出器17と温度検出器1
9に連絡した減算器18の出力として求められ、実質的
に冷却材流量が変動しない定流量運転においては、原子
炉容器出入口温度差ΔTはプラント性能評価乃至運転制
御において重要なパラメータとして使用される。
The amount of heat supplied to the feed water from the reactor coolant in the steam generator 7, ie, the reactor heat output, the pressure of the steam in the steam generator 7, ie, the main steam pressure, and the power generation connected to the steam turbine 23 The electrical output of the machine is a major indicator in operation control or performance evaluation. The thermal output of the reactor is determined by the temperature detector 17 provided in the hot leg 5 and the reactor vessel outlet temperature T HOT of the coolant and the cold leg 13.
Is obtained as the product of the difference between the coolant and the reactor vessel inlet temperature T IN by the temperature detector 19 provided at the reactor vessel, that is, the reactor vessel inlet / outlet temperature difference ΔT and the flow rate of the reactor coolant. The temperature difference ΔT between the entrance and exit of the reactor vessel is determined by the temperature detector 17 and the temperature detector 1.
In the constant flow rate operation, which is obtained as the output of the subtractor 18 connected to 9 and does not substantially change the coolant flow rate, the reactor vessel inlet / outlet temperature difference ΔT is used as an important parameter in plant performance evaluation or operation control. .

【0004】又、原子炉容器1内の炉心3の炉内出力分
布を運転中モニターするために、炉心3の出口に対応し
て複数の炉内温度検出器乃至熱電対が配設されと共に炉
心3を構成する特定の燃料集合体の中に炉内中性子検出
器が配置されている。図5に炉心3の上面の1象限にお
ける炉内温度検出器26と炉内中性子検出器28の配置
が示されている(炉心3の中心軸に関する炉内温度検出
器26と炉内中性子検出器28の配置は通常回転対称に
なっている。)。図において符号29は燃料集合体を示
し、この図から分かるように、燃料集合体29には、炉
内温度検出器26のみが設けられるもの、炉内中性子検
出器28のみが設けられるもの、炉内温度検出器26と
炉内中性子検出器28の双方が設けられるもの及びいず
れも設けられないものがある。尚、炉内温度検出器26
は燃料集合体29に直接設けられるのでは無く、その上
端に対向した上部炉心板側に設けられる。
In order to monitor the power distribution in the reactor core 3 in the reactor vessel 1 during operation, a plurality of reactor temperature detectors or thermocouples are provided at the outlet of the reactor core 3 and the reactor core is provided. The in-furnace neutron detector is arranged in a specific fuel assembly constituting 3. FIG. 5 shows the arrangement of the in-furnace temperature detector 26 and the in-furnace neutron detector 28 in one quadrant on the upper surface of the core 3 (the in-furnace temperature detector 26 and the in-furnace neutron detector with respect to the central axis of the core 3). The arrangement of 28 is usually rotationally symmetric.) In the figure, reference numeral 29 denotes a fuel assembly, and as can be seen from the figure, the fuel assembly 29 is provided with only the in-furnace temperature detector 26, provided with only the in-furnace neutron detector 28, Some are provided with both the internal temperature detector 26 and the in-furnace neutron detector 28, and some are not. The furnace temperature detector 26
Is not provided directly on the fuel assembly 29, but is provided on the upper core plate side facing the upper end thereof.

【0005】[0005]

【発明が解決しようとする課題】上述の冷却材の原子炉
容器出口温度THOTは、大径のホットレグ5の中に配置
された数個の温度検出器17の平均値として得られるの
であるが、種々の要因の影響を受けるので信頼性に問題
がある。従って、本発明の課題は、他の要因による影響
を受けない高信頼度の原子炉容器出入口温度差ΔTを求
める方法を提供することにある。
The above-mentioned coolant outlet temperature T HOT of the coolant is obtained as an average value of several temperature detectors 17 arranged in the large-diameter hot leg 5. However, there is a problem in reliability because it is affected by various factors. Therefore, an object of the present invention is to provide a method for obtaining a highly reliable reactor vessel inlet / outlet temperature difference ΔT which is not affected by other factors.

【0006】[0006]

【課題を解決するための手段】如上の課題を解決するた
め、本発明方法によれば、加圧水型原子炉の原子炉容器
出入口温度差は、炉内温度検出器で得られる炉心出口温
度分布と炉内中性子検出器で得られる炉心出力分布を用
いたサブチャンネル熱解析コードによる炉心出力分布か
ら、前記炉内温度検出器の各設置位置でのそれぞれの温
度上昇幅を求め、それらの値を用いて炉内温度検出器に
よる炉心出口温度分布とサブチャンネル熱解析コードに
よる炉心出力分布との相関式を求め、該相関式を用いて
サブチャンネル熱解析コードによる炉心出口の平均温度
から炉心の平均温度上昇幅を求め、該平均温度上昇幅に
原子炉容器入口温度を加えると共に、炉心バイパス流量
を考慮して原子炉容器出口温度を求め、該原子炉容器出
口温度から前記原子炉容器入口温度を差し引くことによ
り求められる。更に、本発明によれば、好適には前述の
方法の実施に使用される相関式が、炉内温度検出器の測
定結果とサブチャンネル熱解析コードによる解析結果と
の差を求めて、その差を統計的手法を用いて評価を行
い、適用範囲を求め、前記炉内温度検出器の測定結果と
前記サブチャンネル熱解析コードによる解析結果との前
記差が前記適用範囲にあるような、炉内温度検出器の測
定結果とサブチャンネル熱解析コードによる解析結果を
用いて求められる。そして、上述のようにして求められ
た原子炉容器出入口温度差を用いて、加圧水型原子炉の
性能評価が行われる。
According to the method of the present invention, the temperature difference between the inlet and outlet of a reactor vessel of a pressurized water reactor is determined by the core outlet temperature distribution obtained by an in-core temperature detector. From the core power distribution by the sub-channel thermal analysis code using the core power distribution obtained in the reactor neutron detector, determine the respective temperature rise width at each installation position of the furnace temperature detector, using those values A correlation equation between the core outlet temperature distribution by the in-core temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained, and the average core temperature is calculated from the average core outlet temperature by the sub-channel thermal analysis code using the correlation equation. The reactor vessel outlet temperature is determined in consideration of the core bypass flow rate, and the reactor vessel outlet temperature is calculated from the reactor vessel outlet temperature. Obtained by subtracting the reactor vessel inlet temperature. Further, according to the present invention, preferably, the correlation equation used in carrying out the above method determines the difference between the measurement result of the in-furnace temperature detector and the analysis result by the sub-channel thermal analysis code, and calculates the difference. Evaluate using a statistical method, determine the application range, such that the difference between the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code is within the application range, It is obtained using the measurement result of the temperature detector and the analysis result by the sub-channel thermal analysis code. Then, the performance of the pressurized water reactor is evaluated using the reactor vessel inlet / outlet temperature difference obtained as described above.

【0007】[0007]

【発明の実施の形態】以下添付の図面を参照して本発明
の実施形態を説明する。図1は、本発明方法によって原
子炉容器出入口温度差ΔTを求める対象の加圧水型原子
炉30の原子炉容器内の構造を概念的に示している。原
子炉容器31は、原子炉容器胴33と容器蓋35を有し
ており、原子炉容器胴33は一体的に形成された冷却材
用入口ノズル37と出口ノズル39を備えている。原子
炉容器胴33の上部開口部から炉心槽41が垂下支持さ
れ、入口ノズル37に連通した環状下降流路43を画成
しているが、複数の燃料集合体45からなる炉心47を
内部に支持している。更に上部開口部に支持された上部
炉心支持板49に一体的に連結された上部炉心板51が
炉心47の出口側に位置している。そして2本のみ示す
複数のシンブル案内管53にそれぞれ炉内中性子検出器
55が挿脱できるようになっており、一方1本のみ示す
複数の温度検出器57が上部炉心板51の部分に配設さ
れ、これは温度検出器ポート59の部分まで延びてそこ
から原子炉容器外に延出している。前述の複数の炉内中
性子検出器55及び温度検出器57の炉心内での平面配
置は、前述の図5と同様になっている。
Embodiments of the present invention will be described below with reference to the accompanying drawings. FIG. 1 conceptually shows a structure in a reactor vessel of a pressurized water reactor 30 for which a reactor vessel inlet / outlet temperature difference ΔT is to be obtained by the method of the present invention. The reactor vessel 31 has a reactor vessel body 33 and a vessel lid 35, and the reactor vessel body 33 has a coolant inlet nozzle 37 and an outlet nozzle 39 integrally formed. A core tank 41 is suspended from the upper opening of the reactor vessel body 33 and defines an annular descending flow path 43 communicating with an inlet nozzle 37. I support it. Further, an upper core plate 51 integrally connected to an upper core support plate 49 supported by the upper opening is located on the outlet side of the core 47. The in-furnace neutron detectors 55 can be inserted into and removed from a plurality of thimble guide tubes 53 only showing two tubes, respectively, while a plurality of temperature detectors 57 showing only one tube are arranged in the upper core plate 51. This extends to the temperature detector port 59 and extends out of the reactor vessel therefrom. The planar arrangement of the plurality of in-furnace neutron detectors 55 and temperature detectors 57 in the core is the same as that in FIG. 5 described above.

【0008】以上の構成の加圧水型原子炉30内の冷却
材の流動を説明すると、コールドレグに連絡した入口ノ
ズル37から流入した冷却材(軽水)は、環状下降流路
43を通って下部プレナム61に流入する。そして、冷
却材は下部プレナム61内で流れ方向を上向きに変え、
炉心47内を分散して上向きに流れる。詳言すれば、燃
料集合体45の内外を燃料棒の外面に接触しつつこれと
平行に上向きに流れ、核反応熱を奪って昇温する。図示
はしていないが、炉心47の外周部には、形成板領域が
ありここも少量の冷却材が流れて上昇する。このように
炉心47の内外を上向きに流れた冷却材は、上部プレナ
ム63内で合流して混合し、出口ノズル39を通ってホ
ットレグへ流出する。温度検出器57は対応する燃料集
合体45の出口における冷却材温度を検出する。炉内中
性子検出器55は、炉心47内のそれが挿入されている
燃料集合体45内で中性子束を検出する。温度検出器5
7の測定温度は炉心47の水平方向出力分布にほぼ比例
する。
[0008] The flow of the coolant in the pressurized water reactor 30 having the above-described structure will be described. Flows into. Then, the coolant changes the flow direction upward in the lower plenum 61,
Dispersed inside the core 47 and flows upward. More specifically, the fuel rods 45 flow inside and outside of the fuel assembly 45 in contact with the outer surfaces of the fuel rods and flow upward in parallel with the fuel rods, and take up the heat of nuclear reaction to increase the temperature. Although not shown, a forming plate region is provided on the outer peripheral portion of the core 47, and a small amount of coolant flows there and rises. The coolant that has flowed upward and downward in and out of the core 47 merges and mixes in the upper plenum 63 and flows out to the hot leg through the outlet nozzle 39. The temperature detector 57 detects the coolant temperature at the outlet of the corresponding fuel assembly 45. The in-core neutron detector 55 detects a neutron flux in the fuel assembly 45 in the core 47 into which it is inserted. Temperature detector 5
7 is substantially proportional to the power distribution in the horizontal direction of the core 47.

【0009】図2に本発明方法の基本原理を示す。炉心
出力分布から求めた燃料集合体45のそれぞれの出入口
温度差(ΔT)65と温度検出器57の測定温度から求
めた出入口温度差(ΔT)67の偏差が最小となるよう
に相関を求めることによって、より確からしい炉心出入
口温度差(ΔT)を求めることができる。原子炉容器出
入口温度差(ΔT)を求める手順を以下に説明する。 1.炉心出力分布よりサブチャンネル熱解析コードを用
いて温度検出器57の各設置位置における炉心出口温度
(THOT)を求める。 2.温度検出器57の測定値(THOT T/C)及びサブチ
ャンネル熱解析コードによる炉心出口温度計算値(T
HOT CAL)から原子炉入口温度(TIN)を差し引くこと
により、各温度検出器57の設置位置における測定値ベ
ースでの燃料集合体出入口温度差(ΔTT/C)及びサブ
チャンネル熱解析コードベースでの燃料集合体出入口温
度差(ΔTCAL)を求める。 3.燃料集合体出入口温度差(ΔTT/C)と燃料集合体
出入口温度差( ΔT C AL)は、1対1の関係があるた
め(1)式に示す一次式により両者の相関式を求める。 y=ax + b (1) ここで、y:燃料集合体出入口温度差(ΔTT/C)の測
定結果 x:燃料集合体出入口温度差(ΔTCAL)の解析結果 a:傾き b:切片 4.サブチャンネル熱解析コードにより求めた炉心出口
の平均温度 (TH AVG CAL)から原子炉入口温度(TIN)を引くこ
とにより、サブチャンネル熱解析コードベースでの炉心
出入口温度差(ΔTAVG CAL)73を求める。 5.前述の3項で求めた相関式に4項で求めた炉心出入
口温度差 (ΔTAVG CA L)を代入して、図3に示すように
測定値ベースでの炉心出入口温度差 (ΔTcor e cal
71を求める。 6.炉心出入口温度差(ΔTcore cal)71に原子炉容
器入口温度TINを加えることにより炉心出口温度(T
H0T PC)が得られる。 7.前述の炉心出口温度(TH0T PC)に、炉心バイパス
流量を考慮して原子炉容器出口温度(TH0T P)を求め
る。 8.前述の原子炉容器出口温度(TH0T P)から原子炉
容器入口温度TINを差し引くことにより原子炉容器出入
口温度差(ΔTP)を求める。
FIG. 2 shows the basic principle of the method of the present invention. Finding a correlation such that the difference between the inlet / outlet temperature difference (ΔT) 65 of the fuel assembly 45 obtained from the core power distribution and the inlet / outlet temperature difference (ΔT) 67 obtained from the temperature measured by the temperature detector 57 is minimized. Thus, a more reliable core entrance / exit temperature difference (ΔT) can be obtained. The procedure for obtaining the reactor vessel inlet / outlet temperature difference (ΔT) will be described below. 1. The core outlet temperature (T HOT ) at each installation position of the temperature detector 57 is determined from the core power distribution using the sub-channel thermal analysis code. 2. The measured value of the temperature detector 57 (T HOT T / C ) and the core outlet temperature calculated value (T
By subtracting the reactor inlet temperature (T IN ) from HOT CAL ), the fuel assembly inlet / outlet temperature difference (ΔT T / C ) based on the measured value at the installation position of each temperature detector 57 and the subchannel thermal analysis code base The fuel assembly inlet / outlet temperature difference (ΔT CAL ) is determined. 3. Fuel assembly inlet and outlet temperature difference ([Delta] T T / C) and the fuel assembly inlet and outlet temperature difference ([Delta] T C AL) determines the correlation equation between them by a linear expression shown because of the one-to-one relationship (1). y = ax + b (1) where, y: measurement result of fuel assembly inlet / outlet temperature difference (ΔT T / C ) x: analysis result of fuel assembly inlet / outlet temperature difference (ΔT CAL ) a: slope b: intercept 4 . The reactor inlet / outlet temperature difference (ΔT AVG CAL ) based on the sub-channel thermal analysis code is obtained by subtracting the reactor inlet temperature (T IN ) from the average core outlet temperature (T H AVG CAL ) obtained by the sub-channel thermal analysis code. Ask for 73. 5. The core inlet / outlet temperature difference (ΔT AVG CA L ) obtained in the fourth term is substituted into the correlation equation obtained in the above three terms, and as shown in FIG. 3, the core inlet / outlet temperature difference (ΔT cor e cal )
Find 71. 6. By adding the reactor vessel inlet port temperature T IN to the core inlet / outlet port temperature difference (ΔT core cal ) 71, the core outlet port temperature (T
H0T PC ) is obtained. 7. The aforementioned core outlet temperature (T H0T PC), obtaining the consideration of core bypass flow rate reactor vessel outlet temperature (T H0T P). 8. The reactor vessel inlet / outlet temperature difference (ΔT P ) is obtained by subtracting the reactor vessel inlet temperature T IN from the aforementioned reactor vessel outlet temperature (T H0T P ).

【0010】また、本発明では、次のような手順の処理
を行うことにより更に原子炉容器出入口温度差の予測精
度を上げることが可能である。 9.前述のように炉内温度検出器57と炉内中性子検出
器55は炉心47の燃料集合体45の全部には設けられ
ておらず、ほぼ炉心47全体をカバーするように任意に
配置されている。そのため、炉内温度検出器57と炉内
中性子検出器55の双方が設けられている燃料集合体4
5のみで炉心出口温度の相関を求めると、炉内温度検出
器57と炉内中性子検出器55との測定位置の差による
誤差を小さくすることが可能である。 10.温度検出器57の測定値(THOT T/C)とサブチ
ャンネル熱解析コードによる炉心出口温度計算値(T
HOT CAL)との相関のよい燃料集合体45を摘出するた
めに、温度検出器57の測定値(THOT T/C)とサブチ
ャンネル熱解析コードによる炉心出口温度計算値 (T
HOT CAL) との差 (ΔTS)の絶対値 (|ΔTS|)を求め
て、その差の絶対値(|ΔTS|)の標準偏差(σ)を
求める。評価使用範囲 (|ΔTSR) は、平均値 (|Δ
SAVG)に前述の標準偏差(σ)を加えることにより
求められる。評価使用範囲(|ΔTSR)内にある燃料
集合体45の温度検出器57の測定値 (THOT T/C) と
サブチャンネル熱解析コードによる炉心出口温度計算値
(TH OT CAL)のみで、原子炉容器出入口温度差(ΔT
P)は更に精度良く求められる。尚、標準偏差(σ)の
替わりに確率誤差(0.6745σ)又はその他の統計
的手法を用いた制限範囲を用いても良い。更にまた、前
述の原子炉容器出入口温度差(ΔTP)を求める方法で
は、温度をパラメータとしているが、これに変えてエン
タルピをパラメータとして計算することにより、更に精
度が上げられる。
Further, in the present invention, it is possible to further improve the prediction accuracy of the reactor vessel inlet / outlet temperature difference by performing the following procedure. 9. As described above, the in-furnace temperature detector 57 and the in-furnace neutron detector 55 are not provided on all of the fuel assemblies 45 of the core 47, and are arbitrarily arranged so as to cover substantially the entire core 47. . Therefore, the fuel assembly 4 provided with both the furnace temperature detector 57 and the furnace neutron detector 55 is provided.
If the correlation of the core outlet temperature is obtained only with the reference numeral 5, it is possible to reduce the error due to the difference between the measurement positions of the in-furnace temperature detector 57 and the in-furnace neutron detector 55. 10. The core outlet temperature calculated value (T) using the measured value (T HOT T / C ) of the temperature detector 57 and the sub-channel thermal analysis code
In order to extract the fuel assembly 45 having a good correlation with the HOT CAL ), the measured value (T HOT T / C ) of the temperature detector 57 and the core outlet temperature calculated value (T
HOT CAL ), the absolute value (| ΔT S |) of the difference (ΔT S |) is obtained, and the standard deviation (σ) of the absolute value (| ΔT S |) of the difference is obtained. The evaluation use range (| ΔT S | R ) is the average value (| Δ
T s | AVG ) and the standard deviation (σ) described above. The measured value (T HOT T / C ) of the temperature detector 57 of the fuel assembly 45 within the evaluation use range (| ΔT S | R ) and the calculated value of the core outlet temperature by the sub-channel thermal analysis code
(T H OT CAL) only, the reactor vessel inlet and outlet temperature difference ([Delta] T
P ) is determined with higher accuracy. Note that a probability error (0.6745σ) or a limit range using other statistical methods may be used instead of the standard deviation (σ). Further, in the above-described method for obtaining the reactor vessel inlet / outlet temperature difference (ΔT P ), the temperature is used as a parameter. However, the accuracy is further improved by calculating the enthalpy as a parameter instead.

【0011】前述の相関式に、原子炉容器入口温度TIN
を加えて次式のようにして、温度検出器57の各設置位
置における炉心出口温度(THOT)を求めることができ
る。 yt=ax + b + TIN (2) ここで、yt:炉心出口温度(THOT) x:燃料集合体出入口温度差(ΔTCAL)の解析結果 a:傾き b:切片 TIN:原子炉容器入口温度 又、前述したようにサブチャンネル熱解析コードベース
での炉心平均の原子炉容器出入口温度差(ΔT
CORE CAL)を(2)式のxに代入することにより炉心平
均の炉心出口温度(TH0T PC)を求めることができる。
In the above-mentioned correlation equation, the reactor vessel inlet temperature T IN
Then, the core outlet temperature (T HOT ) at each installation position of the temperature detector 57 can be obtained by the following equation. y t = ax + b + T IN (2) where, y t : core outlet temperature (T HOT ) x: analysis result of fuel assembly inlet / outlet temperature difference (ΔT CAL ) a: slope b: intercept T IN : atom Reactor Vessel Inlet Temperature As described above, the core vessel average inlet / outlet temperature difference (ΔT
By substituting ( CORE CAL ) into x in the equation (2), the core average core exit temperature (T H0T PC ) can be obtained.

【0012】前述の相関式である(1)式を次式のよう
に変形することにより、炉内中性子検出器55による測
定を行っていない場合でも、炉内温度検出器57の測定
結果より炉内出力分布を求めることができる。 z =(y−b)/a × α (3) ここで、z:炉内温度検出器57の測定結果より求めた
炉内出力分布 y:(1)式より求めた燃料集合体出入口温度差(ΔT
T/C) a:相関式の定数 b:相関式の定数 α:換算係数
By transforming the above-mentioned correlation equation (1) into the following equation, even when the measurement by the in-furnace neutron detector 55 is not performed, the reactor The internal power distribution can be determined. z = (y−b) / a × α (3) where, z: in-furnace power distribution obtained from the measurement result of the in-furnace temperature detector 57, y: fuel assembly inlet / outlet temperature difference obtained from equation (1) (ΔT
T / C ) a: Correlation equation constant b: Correlation equation constant α: Conversion coefficient

【0013】前述したような加圧水型原子炉のプラント
性能評価は、蒸気発生器における交換熱量である熱出力
が主要な評価項目になっている。そのうちの原子炉容器
出入口温度差ΔTの代わりに前述の原子炉容器出入口温
度差(ΔTP)が使用されて、より確からしい評価がで
きる。又、原子炉容器出入口温度差(ΔTP)を求める
ことにより、従来使用している原子炉容器出入口温度差
ΔTの信頼性の保証を行うことができる。
In the evaluation of the plant performance of the pressurized water reactor described above, the heat output, which is the amount of heat exchanged in the steam generator, is a main evaluation item. The reactor vessel inlet / outlet temperature difference (ΔT P ) is used instead of the reactor vessel inlet / outlet temperature difference ΔT, and a more reliable evaluation can be made. Further, by obtaining the reactor vessel entrance / exit temperature difference (ΔT P ), the reliability of the reactor vessel entrance / exit temperature difference ΔT used conventionally can be guaranteed.

【0014】[0014]

【発明の効果】以上説明したように、本発明によれば、
原子炉容器内に設けた炉内温度検出器による冷却材の炉
心出口温度と炉内中性子検出器による炉内出力分布から
原子炉炉心出口における冷却材平均温度を求めるので、
より精確且つ信頼性のある原子炉容器出入口温度差を求
めることができる。更に本発明によれば、炉内温度検出
器による冷却材の炉心出口温度と炉内中性子検出器によ
る炉内出力分布から求めた計算炉心出口温度とを統計的
に処理することができ、より精確な原子炉容器出入口温
度差を求めることができる。更に本発明によれば、前述
のように精確な原子炉容器出入口温度差を使用するの
で、精確な性能評価ができる。
As described above, according to the present invention,
Since the average coolant temperature at the reactor core outlet is obtained from the core outlet temperature of the coolant by the reactor temperature detector provided in the reactor vessel and the reactor power distribution by the reactor neutron detector,
A more accurate and reliable reactor vessel inlet / outlet temperature difference can be obtained. Further, according to the present invention, it is possible to statistically process the core outlet temperature of the coolant by the in-furnace temperature detector and the calculated core outlet temperature obtained from the in-core power distribution by the in-furnace neutron detector. It is possible to determine the temperature difference between the reactor vessel inlet and outlet. Further, according to the present invention, since the accurate reactor vessel inlet / outlet temperature difference is used as described above, accurate performance evaluation can be performed.

【図面の簡単な説明】[Brief description of the drawings]

【図1】本発明の方法によって原子炉容器出入口温度差
が求められる加圧水型原子炉の原子炉容器内部の概念図
である。
FIG. 1 is a conceptual diagram of the interior of a reactor vessel of a pressurized water reactor in which a temperature difference between a reactor vessel entrance and an exit is obtained by a method of the present invention.

【図2】本発明の基本原理図である。FIG. 2 is a basic principle diagram of the present invention.

【図3】本発明の作用説明図である。FIG. 3 is an operation explanatory view of the present invention.

【図4】従来の加圧水型原子力プラントの代表的な系統
図である。
FIG. 4 is a typical system diagram of a conventional pressurized water nuclear power plant.

【図5】図4の加圧水型原子力プラントにおける炉内温
度検出器と炉内中性子検出器の配置状況を示す概念図で
ある。
FIG. 5 is a conceptual diagram showing the arrangement of a reactor temperature detector and a reactor neutron detector in the pressurized water nuclear power plant of FIG.

【符号の説明】[Explanation of symbols]

30 加圧水型原子炉 31 原子炉容器 33 原子炉容器胴 35 容器蓋 37 入口ノズル 39 出口ノズル 41 炉心槽 43 環状下降流路 45 燃料集合体 47 炉心 49 上部炉心支持板 51 上部炉心板 53 シンブル案内管 55 炉内中性子検出器 57 炉内温度検出器 61 下部プレナム 63 上部プレナム REFERENCE SIGNS LIST 30 pressurized water reactor 31 reactor vessel 33 reactor vessel body 35 vessel lid 37 inlet nozzle 39 outlet nozzle 41 core vessel 43 annular descending flow path 45 fuel assembly 47 core 49 upper core support plate 51 upper core plate 53 thimble guide tube 55 In-furnace neutron detector 57 In-furnace temperature detector 61 Lower plenum 63 Upper plenum

───────────────────────────────────────────────────── フロントページの続き (72)発明者 富永 明彦 香川県高松市太田上町708−1 四電太田 アパート (72)発明者 新原 俊明 兵庫県神戸市兵庫区和田崎町一丁目1番1 号 三菱重工業株式会社神戸造船所内 Fターム(参考) 2G075 AA05 BA03 CA08 CA40 DA01 DA03 FA11 FA19 FB07 FB15 FC06 GA21  ────────────────────────────────────────────────── ─── Continuing on the front page (72) Inventor Akihiko Tominaga 708-1 Otauemachi, Takamatsu City, Kagawa Prefecture Apartment of Shiden Ota (72) Inventor Toshiaki Niihara 1-1-1, Wadazakicho, Hyogo-ku, Kobe City, Hyogo Mitsubishi Heavy Industry Co., Ltd. Kobe Shipyard F-term (reference) 2G075 AA05 BA03 CA08 CA40 DA01 DA03 FA11 FA19 FB07 FB15 FC06 GA21

Claims (3)

【特許請求の範囲】[Claims] 【請求項1】 炉内温度検出器で得られる炉心出口温度
分布と炉内中性子検出器で得られる炉心出力分布を用い
たサブチャンネル熱解析コードによる炉心出力分布か
ら、前記炉内温度検出器の各設置位置でのそれぞれの温
度上昇幅を求め、それらの値を用いて炉内温度検出器に
よる炉心出口温度分布とサブチャンネル熱解析コードに
よる炉心出力分布との相関式を求め、該相関式を用いて
サブチャンネル熱解析コードによる炉心出口の平均温度
から炉心の平均温度上昇幅を求め、該平均温度上昇幅に
原子炉容器入口温度を加えると共に、炉心バイパス流量
を考慮して原子炉容器出口温度を求め、該原子炉容器出
口温度から前記原子炉容器入口温度を差し引くことによ
り加圧水型原子炉の原子炉容器出入口温度差を求める方
法。
1. A core outlet temperature distribution obtained by a subchannel thermal analysis code using a core outlet temperature distribution obtained by an in-core temperature detector and a core power distribution obtained by an in-core neutron detector. The respective temperature rise widths at the respective installation positions are obtained, and the correlation equation between the core outlet temperature distribution by the in-furnace temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained by using those values. The average temperature rise of the core is determined from the average temperature of the core outlet by the sub-channel thermal analysis code, and the reactor vessel inlet temperature is added to the average temperature rise by adding the reactor vessel inlet temperature and the core bypass flow rate. And calculating the reactor vessel inlet / outlet temperature difference of the pressurized water reactor by subtracting the reactor vessel inlet temperature from the reactor vessel outlet temperature.
【請求項2】 炉内温度検出器の測定結果とサブチャン
ネル熱解析コードによる解析結果との差を求めて、その
差を統計的手法を用いて評価を行い、適用範囲を求め、
前記炉内温度検出器の測定結果と前記サブチャンネル熱
解析コードによる解析結果との前記差が前記適用範囲に
あるような、炉内温度検出器の測定結果とサブチャンネ
ル熱解析コードによる解析結果を用いて前記相関式を求
めることを特徴とする請求項1記載の加圧水型原子炉の
原子炉容器出入口温度差を求める方法。
2. The difference between the measurement result of the in-furnace temperature detector and the analysis result by the sub-channel thermal analysis code is obtained, and the difference is evaluated using a statistical method to obtain an applicable range.
The difference between the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code is within the applicable range, and the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code are 2. The method for determining a temperature difference between a reactor vessel inlet and an outlet of a pressurized water reactor according to claim 1, wherein the correlation equation is determined using the correlation equation.
【請求項3】 炉内温度検出器で得られる炉心出口温度
分布と炉内中性子検出器で得られる炉心出力分布を用い
たサブチャンネル熱解析コードによる炉心出力分布か
ら、前記炉内温度検出器の各設置位置でのそれぞれの温
度上昇幅を求め、それらの値を用いて炉内温度検出器に
よる炉心出口温度分布とサブチャンネル熱解析コードに
よる炉心出力分布との相関式を求め、該相関式を用いて
サブチャンネル熱解析コードによる炉心出口の平均温度
から炉心の平均温度上昇幅を求め、該平均温度上昇幅に
原子炉容器入口温度を加えると共に、炉心バイパス流量
を考慮して原子炉容器出口温度を求め、該原子炉容器出
口温度から前記原子炉容器入口温度を差し引いて原子炉
出入口温度差を求め、該原子炉出入口温度差をその性能
評価指数として用いる加圧水型原子炉プラントの性能評
価方法。
3. A core outlet temperature distribution obtained from a core outlet temperature distribution obtained by an in-core temperature detector and a core power distribution obtained by a subchannel thermal analysis code using a core power distribution obtained by an in-core neutron detector. The respective temperature rise widths at the respective installation positions are obtained, and the correlation equation between the core outlet temperature distribution by the in-furnace temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained by using those values. The average temperature rise of the core is determined from the average temperature of the core outlet by the sub-channel thermal analysis code, and the reactor vessel inlet temperature is added to the average temperature rise by adding the reactor vessel inlet temperature and the core bypass flow rate. Is obtained, the reactor vessel inlet / outlet temperature is subtracted from the reactor vessel outlet temperature to obtain a reactor inlet / outlet temperature difference, and the reactor inlet / outlet temperature difference is used as a performance evaluation index. Performance evaluation method for pressurized water reactor plant.
JP25561199A 1999-09-09 1999-09-09 Method for obtaining temperature difference between inlet/outlet of reactor vessel of pressurized water reactor, and method for evaluating performance of pressurized water reactor plant Withdrawn JP2001074875A (en)

Priority Applications (1)

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Publication number Priority date Publication date Assignee Title
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Cited By (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2007232712A (en) * 2006-03-02 2007-09-13 Westinghouse Electric Co Llc Method of recovering operation margin for overheat delta temperature and overpower delta temperature, and nuclear reactor system using the same
JP4585527B2 (en) * 2006-03-02 2010-11-24 ウエスチングハウス・エレクトリック・カンパニー・エルエルシー TRIP CONTROL METHOD FOR REACTOR SYSTEM AND REACTOR SYSTEM
JP2014074657A (en) * 2012-10-04 2014-04-24 Toshiba Corp Used fuel pool water monitoring device, used fuel pool water monitoring method and used fuel pool water monitoring system
CN103871512A (en) * 2012-12-11 2014-06-18 中国核动力研究设计院 Major loop pressure checking method for calculation of saturation temperature of reactor core
CN108344472A (en) * 2018-02-09 2018-07-31 中核控制系统工程有限公司 A kind of liquid level component thermal response detecting system
CN108344472B (en) * 2018-02-09 2024-04-12 中核控制系统工程有限公司 Liquid level assembly thermal response detection system
JP2019152445A (en) * 2018-02-28 2019-09-12 三菱重工業株式会社 Abnormality reducing facility for nuclear reactor and method for determining fixing of control rod

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