JPS58205891A - Nuclear power control device - Google Patents

Nuclear power control device

Info

Publication number
JPS58205891A
JPS58205891A JP57088806A JP8880682A JPS58205891A JP S58205891 A JPS58205891 A JP S58205891A JP 57088806 A JP57088806 A JP 57088806A JP 8880682 A JP8880682 A JP 8880682A JP S58205891 A JPS58205891 A JP S58205891A
Authority
JP
Japan
Prior art keywords
reactor
core
channel
control rod
stability
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.)
Pending
Application number
JP57088806A
Other languages
Japanese (ja)
Inventor
立道 伸一郎
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.)
Toshiba Corp
Original Assignee
Tokyo Shibaura Electric Co 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 Tokyo Shibaura Electric Co Ltd filed Critical Tokyo Shibaura Electric Co Ltd
Priority to JP57088806A priority Critical patent/JPS58205891A/en
Publication of JPS58205891A publication Critical patent/JPS58205891A/en
Pending legal-status Critical Current

Links

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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  • Excavating Of Shafts Or Tunnels (AREA)
  • Control Of Motors That Do Not Use Commutators (AREA)

Abstract

(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。
(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

【発明の詳細な説明】 〔発明の技術分野〕 本発明は、原子炉出力低下要求を引き起す何らかの故障
発生時に、原子炉の継続運転を可能とする原子炉出力制
御装置に関する。
DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a nuclear reactor power control device that enables continued operation of a nuclear reactor in the event of any failure that causes a request to reduce the reactor power.

〔発明の技術的背景〕[Technical background of the invention]

一般に、沸騰水型原子力発電所においては、プラント外
の電力系統事故発生時に所内単独運転を継続することを
目的として、事故発生と同時に原子炉再循環ポンプをト
リップさせるとともに、あらかじめ指定された制御棒(
選択制御棒)を完全挿入位置まで緊急挿入して原子炉出
力を低下させ、一方タービンバイパス弁を急開させて原
子炉から発生する蒸気を復水器へ逃がすことにより、原
子炉を停止することなく継続運転を行なうことができる
ようにしである。
Generally, in a boiling water nuclear power plant, in order to continue isolated operation within the plant in the event of an accident in the power system outside the plant, the reactor recirculation pump is tripped as soon as an accident occurs, and the control rods are (
To shut down the reactor by urgently inserting the selective control rods (selective control rods) to the fully inserted position to reduce the reactor output, while rapidly opening the turbine bypass valve to release steam generated from the reactor to the condenser. This allows continuous operation without any problems.

〔背景技術の問題点〕[Problems with background technology]

ところが選択制御棒を挿入し自然循環、低出力状態に整
定した状態では以下のような原因により炉心の出力分布
が下方に歪んでしまう。
However, when the selective control rods are inserted and the reactor is set to natural circulation and low power, the power distribution of the reactor core is distorted downward due to the following reasons.

(イ) 自然循環のような低流量時には炉心上方で発生
する蒸気泡(ボイド)が十分に押し流されないため、炉
心上方の水による中性子減速効果が不十分となるため相
対的に炉心下方の出力が大きくなる。
(b) During low flow rates such as natural circulation, steam bubbles (voids) generated above the core are not sufficiently swept away, and the neutron moderating effect of the water above the core is insufficient, resulting in a relative decrease in power output below the core. becomes larger.

(ロ)給水系の給水加熱器への熱源としてはタービン抽
気蒸気が使用されるが、タービンバイパス弁を急開して
原子炉から発生する蒸気が復水器にバイパスされるため
、給水加熱器の熱源が失われ炉心にはサブクール度の比
較的高い水(通常運転時に比較して低温の水)が給水さ
れる。そのため原子炉内での沸騰開始点が上方に移動し
、炉心下方の中性子減速効果が助長され炉心下方の出力
が大きくなる。
(b) Turbine extraction steam is used as a heat source for the feedwater heater in the feedwater system, but since the turbine bypass valve is suddenly opened and the steam generated from the reactor is bypassed to the condenser, the feedwater heater The heat source is lost, and water with a relatively high degree of subcooling (water at a lower temperature than during normal operation) is supplied to the core. Therefore, the boiling point within the reactor moves upward, the neutron moderation effect below the core is promoted, and the output below the core increases.

(ハ)第1図に示すように制御棒反応度価値aはボイド
量が多い程大きくなる。よって制御棒が全挿入されると
ボイド量すが炉心上方はど多いため(第2図の破線す参
照)炉心上方程制負の制御棒反応度価値Cが高く、下方
では低いことになる(第2図の実線C参照)。このため
選択制御棒を完全挿入位置まで急速挿入させると出力分
布は下方に企むことになる。
(c) As shown in FIG. 1, the control rod reactivity value a increases as the amount of voids increases. Therefore, when the control rods are fully inserted, the amount of voids is large in the upper part of the reactor core (see the broken line in Figure 2), so the control rod reactivity value C, which controls the control rods, is higher in the upper part of the reactor core, and lower in the lower part (see the broken line in Figure 2). (See solid line C in Figure 2). For this reason, if the selective control rod is quickly inserted to the fully inserted position, the power distribution will be directed downward.

上記原因が重なり合って選択制御棒挿入時には出力分布
が過度に下方に歪む傾向がある。
Due to the combination of the above causes, the output distribution tends to be excessively distorted downward when selective control rods are inserted.

このように出力分布が過度にF方に歪むと炉心のチャン
ネル安定性、炉心安定性が悪くなる。
If the power distribution is excessively distorted in the F direction in this way, the channel stability and core stability of the reactor core will deteriorate.

ここで、炉心部下部の出力歪みが大きくなると炉心のチ
ャンネル安定性、炉心安定性に悪影響をおよぼすことに
ついて説明する。
Here, it will be explained that when the power distortion in the lower part of the reactor core increases, it adversely affects the channel stability and core stability of the reactor core.

一般にBWRプラントのような大規模な非線形システム
の安定性を解析する場合はその構成要素ないしけサブシ
ステムの安定性をまず調べ、次にそれらの結合からなる
全システムの安定性を調べるのが合理的である。
Generally, when analyzing the stability of a large-scale nonlinear system such as a BWR plant, it is reasonable to first examine the stability of its component barge subsystems, and then examine the stability of the entire system consisting of their combination. It is true.

従って、まず炉心内の個々のチャンネル(水路)の熱水
力的安定性を調べ、その固有の安定性を確認スる。(チ
ャンネル安定性) 次にこれらのチャンネルは水力学的に結合され、炉心の
核的特性、および熱伝達特性と結合されて炉心全体の安
定性を調べる。(炉心安定性)まず、チャンネル安定性
について説明する。
Therefore, we first examine the thermal-hydraulic stability of individual channels within the core to confirm their inherent stability. (Channel Stability) These channels are then coupled hydraulically and combined with core nuclear properties and heat transfer properties to determine overall core stability. (Core stability) First, channel stability will be explained.

BWRの炉心には数百本の燃料集合体が装荷され、それ
ぞれの燃料集合体により形成されるチャンネルは並列チ
ャンネルを形成する。このような状況ではたとえ1つの
チャンネルで流量振動が生じたとしでも、その影響は周
囲の多数のチャンネルに吸収され、チャンネル入口、出
口の圧力は変化しないと考えられる。
A BWR core is loaded with hundreds of fuel assemblies, and the channels formed by each fuel assembly form parallel channels. In such a situation, even if flow rate oscillation occurs in one channel, the effect will be absorbed by many surrounding channels, and the pressure at the channel inlet and outlet will not change.

上記の加熱二相流チャンネルにおいてはたとえ加熱量が
一定でも、流れの振動が生ずることが認められる。
It is observed that in the heated two-phase flow channel described above, even if the amount of heating is constant, flow oscillations occur.

この二相流の不安定性については従来から多くの研究が
あ)、不安定性にも種々のタイプがあることがわかり、
系統的な分類も行われている。
There have been many studies on the instability of two-phase flow), and it has been found that there are various types of instability.
A systematic classification has also been carried out.

この分類に従えば今問題にしているチャンネル安定性は
密度波振動に属し、最も一般的なものである。その振動
のメカニズムは一口で云えばチャンネル内の流量、密度
(ボイド率)、圧損の間に存在する輸送おくれとフィー
ドバック効果によるとされ、特徴は振動の周期が流れに
おける密度波(あるいはボイド率外乱の伝播波)のチャ
ンネル通過時間と関係の深いことである。この振動はか
つて流量・ボイドフィードバック不安定性、あるいは時
間遅れ不安定性と呼ばれたことがあったが、上記の特徴
にもとづいて密度波振動と呼ばれるようになった。
According to this classification, the channel stability in question belongs to density wave vibration, and is the most common type. The mechanism of this vibration is said to be due to the transport delay and feedback effect that exist between the flow rate, density (void ratio), and pressure drop in the channel. This is closely related to the channel transit time of the propagating wave. This oscillation was once called flow rate/void feedback instability or time delay instability, but based on the above characteristics, it is now called density wave oscillation.

さて、BWR燃料チャンネルでは第3図に示すように入
口から単相流部、サブクール沸騰部、ノ(ルク沸騰部が
ある。サブクール沸騰部では水のエンタルピは飽和エン
タルピに達してないが蒸気泡の存在する部分であり、バ
ルク沸騰部では水も飽和に達している。こ\で具体的に
振動に至る場合のメカニズムを説明する。
As shown in Figure 3, in the BWR fuel channel, there is a single-phase flow section, a subcooled boiling section, and a nok boiling section from the inlet.In the subcooled boiling section, the enthalpy of water has not reached the saturated enthalpy, but steam bubbles In the bulk boiling part, water has also reached saturation.Here we will specifically explain the mechanism that leads to vibration.

簡単のためにサブクール沸騰部を無視するが定性的な議
論なので、影響はない。
For simplicity, we ignore the subcooled boiling part, but this is a qualitative discussion, so it has no effect.

今チャンネルの入口流量が振動している場合を考える。Now consider the case where the inlet flow rate of the channel is oscillating.

この振動は単相流部にエンタルピ外乱の流れに沿った伝
播を起こす。水の温度が飽和に達する点である沸騰境界
(以下B、B、と略す)はこのエンタルピ外乱によって
振動する。流量と拳相流動の長さが振動する結果、単相
流部の圧損も振動することとなる。B、B、の振動は、
すなわちこの点におけるボイド率あるいはクォリティー
の振動であり、これが流れに沿って伝播し、また、同時
に二相流部の流速にも外乱を生ぜしめる。このボイド率
、流速の外乱とB、B、の振動による二相流部の長さの
振動は相まって二相流部の圧損の外乱を引起す。ところ
で、チャンネル全体の圧損は外部から境界条件として与
えられ、今の場合数百本の他のチャンネルよ如決まる定
数であるので、二相流部の圧損の外乱はそれがそのまま
大きさが同じで符号が逆の圧損変化を単相流に及ぼす。
This vibration causes enthalpy disturbance to propagate along the flow in the single-phase flow section. The boiling boundary (hereinafter abbreviated as B, B), which is the point at which the water temperature reaches saturation, oscillates due to this enthalpy disturbance. As a result of the oscillation of the flow rate and the length of the fist-phase flow, the pressure drop in the single-phase flow section also oscillates. The vibration of B, B is
In other words, the void ratio or quality oscillates at this point, and this propagates along the flow, and at the same time causes a disturbance in the flow velocity in the two-phase flow section. This disturbance in the void ratio and flow velocity and the vibration in the length of the two-phase flow section due to the vibrations of B and B combine to cause a disturbance in the pressure drop of the two-phase flow section. By the way, the pressure drop of the entire channel is given as a boundary condition from the outside, and in this case it is a constant that is determined like hundreds of other channels, so the disturbance to the pressure drop in the two-phase flow section has the same size as it is. Effects pressure drop changes of opposite sign on single-phase flow.

これが最初に仮定した入口流速の振動を助長(不安定の
場合)したシ、減衰(安定の場合)したりすることにガ
る。
This may encourage (in the case of instability) or dampen (in the case of stability) the initially assumed oscillation of the inlet flow velocity.

さて、臨界振動の場合についてもう少し検討する。この
ときには単相流部の圧損の変化分は二相流部の圧損の変
化分と大きさは同じで符号が逆となる。ところで、BW
Rの運転条件では単相流部の圧損の変化は入口流量の変
化分とほぼ同相であシ、一方二相流部の圧損の変化は出
口流量の変化とはソ同相である。従って流量はこの場合
、入口、出口で大きな位相おくれをもつこ゛とになる。
Now, let's consider the case of critical vibration a little more. In this case, the change in pressure loss in the single-phase flow section has the same magnitude as the change in pressure loss in the two-phase flow section, but has the opposite sign. By the way, B.W.
Under operating conditions R, the change in pressure drop in the single-phase flow section is approximately in phase with the change in inlet flow rate, while the change in pressure drop in the two-phase flow section is in phase with the change in outlet flow rate. Therefore, in this case, the flow rate will have a large phase lag at the inlet and outlet.

このような位相おくれは非圧縮性の単相流では生じるこ
とはなく、沸騰チャンネルの流れにそった大きな密度変
化によるものである。すなわち、入口流量が大きいとき
にチャンネルに入った水はその流速が早いために飽和温
度に達するのに多くの距離を通過しなければならない。
Such phase lag does not occur in incompressible single-phase flow, but is due to large density changes along the flow in the boiling channel. That is, water entering the channel when the inlet flow rate is high has to travel a large distance to reach the saturation temperature due to its high flow rate.

すなわち沸騰境界は下流側に移る。沸騰部においては、
負のボイド率外乱として伝播し、このことは水と蒸気の
密度差から質量流量としては正の外乱として伝播する。
In other words, the boiling boundary moves downstream. In the boiling part,
It propagates as a negative void fraction disturbance, and this propagates as a positive disturbance in terms of mass flow rate due to the density difference between water and steam.

その結果沸1a部の圧損の増加を来たし、単相流部の圧
損が低下し、更に入口流量を減少するように働く。この
ことから振動の1/2周期がはソ流体のチャンネル通過
時間に等しくなることが説明できる。
As a result, the pressure drop in the boiler section 1a increases, the pressure drop in the single-phase flow section decreases, and the inlet flow rate further decreases. From this, it can be explained that 1/2 period of vibration is equal to the passage time of the solenoid fluid through the channel.

このような状態において、2相流部の圧損を大きくする
ことは二相流部の圧損の変化をまし、チャンネルの不安
定度を増す。一方入ロオリフイスの絞りを大匙ぐするこ
とは沸違部圧損の変化による単相部圧損の変化から入口
流量の変化へのゲイン全下げ安定度の向上につながる。
In such a state, increasing the pressure drop in the two-phase flow section outweighs the change in the pressure drop in the two-phase flow section and increases the instability of the channel. Increasing the orifice of the one-way flow orifice by a large spoon will lead to improved stability in reducing the gain completely from changes in single-phase pressure loss due to changes in boiling part pressure loss to changes in inlet flow rate.

これらの傾向は実験的に確かめられて騒る。These trends have been confirmed experimentally.

上記原因によシ、炉心部下部の出力ビーキングが大きく
なると、ボイド率が増加する。このため2相流部の圧損
が大きくな〕、チャンネルの不安定度が増すことになる
Due to the above reasons, when the power peaking in the lower part of the reactor core increases, the void ratio increases. As a result, the pressure drop in the two-phase flow section becomes large] and the instability of the channel increases.

次に、炉心安定性にういて説明する。Next, core stability will be explained.

チャンネル安定性が確保された場合、次に検討するのは
、チャンネルが数百本以上集って形成されている炉心の
安定性でア乙。この場合にはチャンネルの熱水力特性と
ボイドの反応度係数による原子炉核特性との結合の結果
不安定になることが考えられる。?:、\では多数の燃
料チャンネルを熱力学的および水力学的に似たものを集
めていくつかのグループにまとめ、全炉心を模擬する。
Once channel stability is ensured, the next consideration is the stability of the core, which is made up of hundreds of channels or more. In this case, instability may occur as a result of the coupling between the thermal-hydraulic characteristics of the channel and the reactor core characteristics due to the reactivity coefficient of the void. ? In :,\, a large number of fuel channels that are thermodynamically and hydraulically similar are grouped together to simulate the entire reactor core.

各燃料チャンネルについての熱水力学的動特性はチャン
ネル安定性の数学モデルから導かれ、こ\ではチャンネ
ル圧損および、熱流束の入力に対してチャンネルの流量
、およびボイド率を出力する。さて、チャンネル流量の
総和は炉心再循環流量となるが、炉心の再循環ルーズに
対する動特性モデルを求めて炉心入口プレナムの圧力変
化を得ることができる。最後に各チャンネルのボイド率
にボイド反応度係数を乗じ、總和金とることにより炉心
全体の反応度変化がまとまり、これが炉心核熱特性への
入力となってフィードバック系を構成する。
Thermal-hydraulic dynamics for each fuel channel are derived from a mathematical model of channel stability, which outputs channel pressure drop and channel flow rate and void fraction for inputs of heat flux. Now, the sum of the channel flow rates becomes the core recirculation flow rate, and the pressure change in the core inlet plenum can be obtained by finding a dynamic characteristic model for the core recirculation looseness. Finally, the void fraction of each channel is multiplied by the void reactivity coefficient, and the reactivity changes of the entire reactor core are summarized by multiplying the void fraction of each channel by the void reactivity coefficient.This becomes an input to the core thermal characteristics and constitutes a feedback system.

従ってこの場合には通常のフィードバックシステムの安
定性と同様であシ、理解しやすbものであるO 上述のように、チャンネル安定性の場合と同様、炉心部
下部の出力ビーキングが大きくなるとボイド率が増加し
、ボイドの反応度へのフィードバックゲインが高まり、
炉心の不安定度が増すことになる。
Therefore, in this case, the stability is similar to that of a normal feedback system, and is easy to understand. increases, the feedback gain to the void reactivity increases,
The instability of the reactor core will increase.

以上の結果、炉心部下部の出力ビーキングが大きくなる
と、炉心のチャンネル内のボイド量がふえ、2相流の圧
力損失やボイド量が上昇することになる。この結果チャ
ンネル内流量に微少な外乱が生じた場合、ボイドによる
流量外乱遅れが大きくな如振動が継続し易くなる傾向と
なる。・又、反応度的にみてもボイド反応度が増し、不
安定な傾向となる。これが、いわゆる[炉心のチャンネ
ル安定性の悪化]である。
As a result of the above, when the power peaking in the lower part of the reactor core increases, the amount of voids in the channels of the reactor core increases, and the pressure loss and the amount of voids in the two-phase flow increase. As a result, when a slight disturbance occurs in the flow rate in the channel, the vibration tends to continue as the flow rate disturbance delay due to voids becomes large.・Also, in terms of reactivity, the void reactivity increases and tends to become unstable. This is the so-called [deterioration of core channel stability].

〔発明の目的〕[Purpose of the invention]

本発明は沸騰水型原子力発電所における所内又は所外の
原子炉出力低下要求を引き起す何らかの故障発生時にあ
らかじめ選択した制御棒を急速挿入することによりすみ
やかに低出力状態に移行し、かつ継続して原子炉を安定
な状態で運転可能とする原子炉出力制御装Bを提供する
ことにある。
The present invention provides for rapid insertion of preselected control rods to quickly transition to and maintain a low power state in the event of any failure in a boiling water nuclear power plant that causes an on-site or off-site reactor power reduction request. The object of the present invention is to provide a nuclear reactor power control system B that enables a nuclear reactor to operate in a stable state.

〔発明の概要〕[Summary of the invention]

本発明は、電力系統事故発生時等に予じめ選択された一
部の制御棒(選択制御棒)を急速挿入することによって
原子炉の出力を低下せしめる原子炉制御装置において、
上記緊急挿入されるべき選択制御棒を部分挿入位置で停
止させる原子炉出力制御装置である。
The present invention provides a nuclear reactor control device that reduces the output of a nuclear reactor by rapidly inserting some control rods selected in advance (selective control rods) when a power system accident occurs.
This is a nuclear reactor power control device that stops the selected control rods to be urgently inserted at a partial insertion position.

〔発明の実施例〕[Embodiments of the invention]

第4図において、符号1は原子炉であって、その原子炉
1から発生した蒸気は主蒸気管2全通して主蒸気加減弁
3を介してタービン4に送給され、タービン4の駆動が
行なわれる。上記ターヒフ40回転は発電機5に伝えら
れそこで電気に変換され、主変圧器6、主し中断器7を
介して系統8へと送電される。一方、タービン4で仕事
を行なった蒸気は復水器9で復水せしめられ、その後原
子炉給水ポンプ10によって原子炉1へ給水管10aを
介し還流される。また、前記主蒸気管2と復水器9との
間には、主蒸気加減弁3およびタービン4をバイパスす
るとともにバイパス弁11を有するバイパス管12が接
続されてお)、上記バイパス管12を経た蒸気も前記復
水器9で復水せしめられる。
In FIG. 4, reference numeral 1 denotes a nuclear reactor, and the steam generated from the reactor 1 passes through the main steam pipe 2 and is fed to the turbine 4 via the main steam control valve 3, and the turbine 4 is driven. It is done. The 40 rotations of Tahif are transmitted to the generator 5, where it is converted into electricity, and the electricity is transmitted to the grid 8 via the main transformer 6 and the main interrupter 7. On the other hand, the steam that has performed work in the turbine 4 is condensed in a condenser 9, and then returned to the reactor 1 via a water supply pipe 10a by a reactor feed water pump 10. Furthermore, a bypass pipe 12 that bypasses the main steam control valve 3 and the turbine 4 and has a bypass valve 11 is connected between the main steam pipe 2 and the condenser 9. The passed steam is also condensed in the condenser 9.

ところで、上述のように構成された沸騰水型原子力発電
所においては、系統8または発電機5に伺らかの故障が
発生すると、主しe断器7、主変圧器6が開き、主変圧
器6の開信号が負荷しゃ断検出回路13に送られる。こ
のようにして主変圧器6のRm1が負荷しゃ断検出回路
131Cよって検出されると、上記負荷しゃ断検出回路
13から負荷しゃ断信号が発生せしめられ生蒸気加減弁
3が急閉されるとともにバイパス弁11が急開される。
By the way, in the boiling water nuclear power plant configured as described above, when a failure occurs in the system 8 or the generator 5, the main e-breaker 7 and the main transformer 6 open, and the main transformer The open signal of the device 6 is sent to the load cutoff detection circuit 13. In this way, when Rm1 of the main transformer 6 is detected by the load cutoff detection circuit 131C, a load cutoff signal is generated from the load cutoff detection circuit 13, the live steam control valve 3 is suddenly closed, and the bypass valve 11 is suddenly opened.

しかして、原子炉1から夕〜ビン4に送られる蒸気は、
上記主蒸気加減弁3によって急速しゃ断され、バイパス
弁11を経て復水器9へと送給される。
Therefore, the steam sent from reactor 1 to bin 4 is
The main steam control valve 3 quickly shuts off the steam, and the steam is fed to the condenser 9 via the bypass valve 11.

一方、主蒸気加減弁3の急閉は主蒸気加減弁急閉検出装
置14で検出され、上記主蒸気加減弁3の急閉に応じて
原子炉再循環ポンプ15の駆動モータ16が急速停止せ
しめられるとともに、選択制御棒設定挿入装置17が作
動せしめられ、選択された制御棒18のうち、指定数が
部分挿入され、残りは完全挿入され、原子炉1の出力が
低下せしめられる。また、主蒸気加減弁急閉装置14か
らの信号は給水ポンプ制御装置2oに送られ、その給水
ポンプ制御装置2oによって原子炉給水ポンプ1001
台または複数台の停止が行なわれ、原子炉の所内単独運
転への移行が行なわれる。もちろん原子炉10安全性を
確保するため、例えば炉内の水位りが低下して設定値に
達すると、水位検出器21が作動して水位低信号が発せ
られ、緊急停止回路22が作動せしめられ、すべての制
御棒23が炉心19内に緊急に全挿入され、原子炉の停
止が行なわれる。
On the other hand, the sudden closing of the main steam regulating valve 3 is detected by the main steam regulating valve sudden closing detection device 14, and the drive motor 16 of the reactor recirculation pump 15 is rapidly stopped in response to the sudden closing of the main steam regulating valve 3. At the same time, the selective control rod setting and insertion device 17 is activated, a specified number of the selected control rods 18 are partially inserted, and the rest are fully inserted, thereby reducing the output of the reactor 1. Further, the signal from the main steam control valve quick closing device 14 is sent to the feed water pump control device 2o, and the feed water pump control device 2o controls the reactor feed water pump 1001.
One or more reactors are shut down, and the reactor shifts to isolated operation within the station. Of course, in order to ensure the safety of the reactor 10, for example, when the water level inside the reactor drops and reaches a set value, the water level detector 21 is activated to issue a low water level signal, and the emergency shutdown circuit 22 is activated. , all control rods 23 are urgently fully inserted into the reactor core 19, and the reactor is shut down.

このように本発明は炉心内に配置された制御棒の一部又
は全部が部分挿入位置もしくは完全挿入位置のいずれの
位置にも急速挿入可能となっている。すなわち選択制御
棒設定挿入回路17によりあらかじめ選択制御棒を決め
、その中の一部又は全部を部分挿入制御棒として指定を
する。この指定の一例を第5図に示す。第5図は炉心を
上方から見た概略図で、口の中に制御棒が1本、燃料集
合体が4体位置する。選択制御棒の指定は急速挿入後の
径方向の出力分布の歪みをなるべく少くするよう対称性
を考慮して設定する。Aは完全挿入位置まで挿入される
制御棒、Bは部分挿入位置まで挿入される制御棒を示す
。○印は選択制御棒挿入以前から部分挿入されていた制
御棒を示す。この例に於て選択制御棒設定挿入回路17
が作動すると最終状態として第6図に示すように全挿入
された制御棒と部分挿入された制御棒とが混在した状態
となる。部分挿入制御棒の挿入深さは第6図では6/1
0部分挿入であるが一般的には、軸方向の出力分布が比
較的平担になるよう、最適な位置に予め決めておく。
In this way, the present invention allows a part or all of the control rods arranged in the reactor core to be quickly inserted into either the partial insertion position or the complete insertion position. That is, the selected control rod setting and insertion circuit 17 determines the selected control rods in advance, and designates some or all of them as partially inserted control rods. An example of this designation is shown in FIG. Figure 5 is a schematic view of the core seen from above, with one control rod and four fuel assemblies located in the mouth. The selected control rods are designated with symmetry in mind so as to minimize the distortion of the radial power distribution after rapid insertion. A shows the control rod inserted to the fully inserted position, and B shows the control rod inserted to the partially inserted position. The ○ mark indicates a control rod that was partially inserted before the selective control rod was inserted. In this example, the selection control rod setting insertion circuit 17
When activated, the final state is a state in which fully inserted control rods and partially inserted control rods coexist as shown in FIG. The insertion depth of the partially inserted control rod is 6/1 in Figure 6.
Generally, the optimal position is determined in advance so that the output distribution in the axial direction is relatively even though it is a zero-part insertion.

ここでM7図に選択制御棒の先端を軸方向の任意位置で
停止させたときの停止位置と軸方向ピーキング係数との
関係を示す。これによれば選択制御棒の停止位置を少し
ずつ上向に移動させると出力は少しずつ上方に押しあげ
られ出力分布は平担化するが、更に上方に移動させると
制御棒の効き方がボイドの多い炉心上方で大きいため上
方の出力が抑えられ、相対的に炉心下方の出力ビーキン
グは急速に大きくなる。安定性の面からは出力ビーキン
グの最も低いところが停止位置としては最適なのでおる
が、選択制御棒のもともとの機能は所定の大きな負の反
応度を炉心に印加することなのであるから、余りにも浅
い位置で停止させると8 ては炉心の下端よ”10””’10の範囲が適切である
Here, Fig. M7 shows the relationship between the stop position and the axial peaking coefficient when the tip of the selection control rod is stopped at an arbitrary position in the axial direction. According to this, if the stop position of the selected control rod is moved upward little by little, the output will be pushed upward little by little and the output distribution will be flattened, but if it is moved further upward, the effectiveness of the control rod will become void. Since it is large in the upper part of the core where there is a lot of power, the power output in the upper part is suppressed, and the power peaking in the lower part of the core increases rapidly. From the standpoint of stability, the optimal stopping position is at the lowest power beak, but since the original function of the selection control rod is to apply a predetermined large negative reactivity to the core, it is recommended that the stop position be too shallow. If the reactor is stopped at 8, then a range of 10 to 10 from the lower end of the core is appropriate.

選択制御棒としては、選択制御棒挿入前にすでに部分挿
入されていた制御棒を選択することも勿論可能である。
As the selected control rod, it is of course possible to select a control rod that has already been partially inserted before the selection control rod is inserted.

しかして、今系統8または発電機5等が故障し、主変圧
器6等が開かれると、負荷しゃ断検出回路13を介して
主蒸気加減弁3が急閉されるとと屯にバイパス弁11が
急開され、原子炉1からの蒸気は復水器9へ放出される
。一方、前述のように主蒸気加減弁3の急閉に応じて原
子炉再循環ポンプ15が停止し、約30秒後には原子炉
1の出力は50〜60係定格出力まで低下する。さらに
上記主蒸気加減弁3の急閉に対応して、選択制御棒設定
挿入回路17が作動し、数秒で最外周制御棒(図示せず
)の一部または全部が緊急挿入され、約30係定格出力
相当分だけ原子炉1の出力は低下せしめられる。このよ
うに原子炉再循環ポンプ15の停止と最外周制御棒の緊
急挿入によル、原子炉1の出力は20〜30チ定格出力
1で低下し、上記原子炉1で発生した蒸気はバイパス弁
11を介して復水器9へ放出され、上記原子炉1は系統
8と分離した状態の所内単独運転状態となる。このとき
所定の給水ポンプ10の運転停止釦よって原子炉再循環
ポンプ15の停止に伴なう水位りの上昇は抑制され、緊
急停止回路22が作動するのを防止している。
However, if the system 8 or the generator 5 or the like breaks down and the main transformer 6 or the like is opened, the main steam control valve 3 is suddenly closed via the load cutoff detection circuit 13 and the bypass valve 11 is suddenly closed. is suddenly opened, and steam from the reactor 1 is released to the condenser 9. On the other hand, as described above, the reactor recirculation pump 15 is stopped in response to the sudden closing of the main steam control valve 3, and after about 30 seconds, the output of the reactor 1 is reduced to the 50-60 rated output. Furthermore, in response to the sudden closing of the main steam control valve 3, the selective control rod setting and insertion circuit 17 is activated, and in a few seconds, a part or all of the outermost control rods (not shown) are urgently inserted. The output of the reactor 1 is reduced by an amount equivalent to the rated output. As described above, due to the stoppage of the reactor recirculation pump 15 and the emergency insertion of the outermost control rod, the output of the reactor 1 is reduced by 20 to 30 inches at the rated output 1, and the steam generated in the reactor 1 is bypassed. The reactor 1 is discharged to the condenser 9 via the valve 11, and the reactor 1 is isolated from the system 8 and becomes in an isolated state. At this time, the predetermined operation stop button of the water supply pump 10 suppresses the rise in the water level due to the stoppage of the reactor recirculation pump 15, and prevents the emergency stop circuit 22 from operating.

第8図に所内単独運転状態に移行したときの原子炉内の
軸方向出力分布例を示す。
FIG. 8 shows an example of the axial power distribution inside the nuclear reactor when transitioning to the in-station isolated operation state.

従来のように選択制御棒が全挿入しか許されなかった場
合破線dに示すように極端に炉心下部に出力分布が歪む
可能性があったが部分挿入の選択制御棒が存在すると、
第8図の実線eに示すように炉心上半での出力の持分が
大きくなり、又同時に下部での出力が抑えられて全体と
して出力ピーク位置が上方に押しあげられ、又ビ〜りの
絶対値も小さくなる。このため沸騰開始点が上方に移行
し、又ボイドの量も減少するのでチャンネル安定性、炉
心安定性が格段に向上し、系統8等の故障が復旧し出力
運転に移行するまでの時間、安定した状態で所内単独運
転を行うことができる。・又系統の故障以前の出力運転
中に部分挿入していた制御棒も選択制御棒として緊急挿
入可能であるため、制御棒の印加反応度速度を速めるこ
とが可能である。また、第9図に示すように全引抜状態
から制御棒を挿入しても始めのうちはなかなか負の反応
度が印加されず挿入深さ半分ぐらbから急激に立ちあが
る特性をもっている。従って出力運転中すでに途中まで
挿入されていた制御棒を選択制御棒挿入と同時に挿入す
ることによシ短時間により大きな負の反応度を印加する
ことができる。
If selective control rods were only allowed to be fully inserted as in the past, there was a possibility that the power distribution would be extremely distorted in the lower part of the core, as shown by the broken line d, but if selective control rods were partially inserted,
As shown by the solid line e in Figure 8, the share of the output in the upper half of the core increases, and at the same time the output in the lower part is suppressed, pushing the output peak position upward as a whole. The value also becomes smaller. As a result, the boiling start point moves upward and the amount of voids decreases, resulting in a marked improvement in channel stability and core stability. It is possible to perform independent operation within the station in this state.・In addition, control rods that were partially inserted during output operation before the system failure can be inserted as selective control rods in an emergency manner, making it possible to increase the applied reactivity rate of the control rods. Further, as shown in FIG. 9, even when the control rod is inserted from the fully withdrawn state, negative reactivity is not applied at first, but rises rapidly from about half the insertion depth b. Therefore, by inserting a control rod that has already been inserted halfway during output operation at the same time as the selective control rod is inserted, a larger negative reactivity can be applied in a shorter time.

なお系統の故障によシ主蒸気加減弁3は急速閉するがバ
イパス弁11が急開されるまで少し時間遅れがあり、こ
のため炉心内の圧力は高まってボイドがつぶれ正の反応
度が印加されて一時的に出力が上昇する。この様な場合
上述のように出力運転中部分挿入されていた制御棒を他
の選択制御棒と同時に急速挿入することによ)一時的な
出力上昇を抑制する効果があり、所内単独運転に移行す
るまでの過渡特性を緩和出来る。
In addition, due to a failure in the system, the main steam control valve 3 closes quickly, but there is a slight delay until the bypass valve 11 is suddenly opened, and as a result, the pressure inside the core increases, voids are collapsed, and a positive reactivity is applied. output temporarily increases. In such cases, as mentioned above, by rapidly inserting the control rods that were partially inserted during output operation at the same time as other selected control rods), it is effective to suppress the temporary increase in output, and shift to isolated operation within the station. It is possible to alleviate the transient characteristics until

〔発明の効果〕〔Effect of the invention〕

以上のように、選択制御棒挿入時におりても、炉心の出
力分布が平担化され、チャンネル安定性。
As described above, even when selective control rods are inserted, the power distribution in the core is flattened and channel stability is maintained.

炉心安定性が向上するっCore stability will improve

【図面の簡単な説明】[Brief explanation of the drawing]

第1図は一般的にボイド量と制御棒の反応度価値の関係
を示す図、第2図は一般的にボイド量と制御棒反応度価
値の軸方向分布を示す図、第3図は一般的に二相流沸騰
チャンネルの様子を示す図、第4図は本発明の概略系統
図、第5図は本発明にかかる部分挿入と完全挿入の選択
制御棒の選び方の一例、第6図は本発明にかかる選択制
御棒挿入後のパターン、第7図は本発明にかかる選択制
御棒先端停止位置をパラメータとした軸方向ピーキング
係数の特性図、第8図は本発明を実施した場合としない
場合の出力分布の比較図、第9図は本発明にかかる制御
棒の挿入深さと印加反応度の関係を示す図である、 1・原子炉      3・主蒸気加減弁4 タービン
     9・・・復水器10・原子炉給水ポンプ 1
1  バイパス弁13・・・負荷しゃ断検出回路 ・・ 14・主蒸気加減弁急閉検出装置 15・・原子?再循環ポンプ 17・選択制御棒設定挿入装置 18  制御棒      19  炉心20・給水ポ
ンプ制御装置 22・・緊急停止回路   25・異常検出回路(73
17)  代理人 弁理士  則 近 憲 佑 (ほか
1名)11這 第1図   第2図
Figure 1 generally shows the relationship between void volume and control rod reactivity value, Figure 2 generally shows the axial distribution of void volume and control rod reactivity value, and Figure 3 generally shows the relationship between void volume and control rod reactivity value. 4 is a schematic diagram of the present invention, FIG. 5 is an example of how to select a control rod for partial insertion and complete insertion according to the present invention, and FIG. 6 is a diagram showing the state of a two-phase flow boiling channel. The pattern after insertion of the selective control rod according to the present invention, FIG. 7 is a characteristic diagram of the axial peaking coefficient with the selective control rod tip stop position according to the present invention as a parameter, and FIG. 8 is a diagram showing the case where the present invention is not implemented. Figure 9 is a diagram showing the relationship between control rod insertion depth and applied reactivity according to the present invention. 1. Nuclear reactor 3. Main steam control valve 4 Turbine 9. Water device 10/reactor water pump 1
1 Bypass valve 13... Load cutoff detection circuit... 14. Main steam control valve sudden closing detection device 15... Atomic? Recirculation pump 17, selective control rod setting and insertion device 18, control rod 19, core 20, water supply pump control device 22, emergency stop circuit 25, abnormality detection circuit (73
17) Agent: Patent Attorney Noriyuki Chika (and 1 other person) 11 Figure 1 Figure 2

Claims (1)

【特許請求の範囲】 1、予じめ選択された一部の制御棒を急速挿入すること
によって原子炉の出力を低下せしめる原子炉制御装置に
おいて、上記緊急挿入されるべき選択制御棒を部分挿入
位置で停止させることを特徴とする原子炉出力制御装置
。 2、 選択制御棒の全部を部分挿入位置で停止させるこ
とを特徴とする特許請求の範囲第1項記載の原子炉出力
制御装置。 3、選択制御棒の一部を部分挿入位置で停止させること
を特徴とする特許請求の範囲第1項記載の原子炉出力制
御装置。
[Claims] 1. In a nuclear reactor control device that reduces the output of a nuclear reactor by rapidly inserting some control rods selected in advance, the selected control rods to be urgently inserted are partially inserted. A nuclear reactor power control device characterized by stopping at a certain position. 2. The nuclear reactor power control system according to claim 1, characterized in that all of the selected control rods are stopped at the partially inserted position. 3. The nuclear reactor power control system according to claim 1, wherein a part of the selected control rod is stopped at a partially inserted position.
JP57088806A 1982-05-27 1982-05-27 Nuclear power control device Pending JPS58205891A (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
JP57088806A JPS58205891A (en) 1982-05-27 1982-05-27 Nuclear power control device

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
JP57088806A JPS58205891A (en) 1982-05-27 1982-05-27 Nuclear power control device

Publications (1)

Publication Number Publication Date
JPS58205891A true JPS58205891A (en) 1983-11-30

Family

ID=13953121

Family Applications (1)

Application Number Title Priority Date Filing Date
JP57088806A Pending JPS58205891A (en) 1982-05-27 1982-05-27 Nuclear power control device

Country Status (1)

Country Link
JP (1) JPS58205891A (en)

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