JPH07151899A - Nuclear reactor power controller - Google Patents

Nuclear reactor power controller

Info

Publication number
JPH07151899A
JPH07151899A JP5297710A JP29771093A JPH07151899A JP H07151899 A JPH07151899 A JP H07151899A JP 5297710 A JP5297710 A JP 5297710A JP 29771093 A JP29771093 A JP 29771093A JP H07151899 A JPH07151899 A JP H07151899A
Authority
JP
Japan
Prior art keywords
reactor
signal
scram
safety valve
set pressure
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
JP5297710A
Other languages
Japanese (ja)
Inventor
Hiromitsu Imaruoka
浩充 伊丸岡
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
Toshiba Corp
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 Toshiba Corp filed Critical Toshiba Corp
Priority to JP5297710A priority Critical patent/JPH07151899A/en
Publication of JPH07151899A publication Critical patent/JPH07151899A/en
Pending legal-status Critical Current

Links

Classifications

    • Y—GENERAL 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
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin

Landscapes

  • Monitoring And Testing Of Nuclear Reactors (AREA)
  • Structure Of Emergency Protection For Nuclear Reactors (AREA)

Abstract

PURPOSE:To surely keep a nuclear reactor water level by automatically operat ing a main steam relief safety valve at the time of a reactor scram-impossible transient event to reduce the pressure of a nuclear reactor, and expediting pouring of cooling water from an emergency reactor core cooler. CONSTITUTION:The nuclear reactor power controller comprises a scram- impossible transient event judging unit 1 for judging occurrence of scram- impossible transient event, a dynamic characteristic void reactivity calculator 2 for outputting dynamic characteristic void coefficient signal, a nuclear reactor power predicting unit 3 for outputting a predicted unclear reactor power signal, a main steam relief safety valve switching set pressure altering unit 4 for outputting an opening set pressure altering signal and a closing set pressure altering signal of a main steam relief safety valve, and a main steam relief safety valve opening/closing set pressure controller 5 for controlling opening/ closing set pressure of the valve.

Description

【発明の詳細な説明】Detailed Description of the Invention

【０００１】[0001]

【産業上の利用分野】本発明は沸騰水型原子炉における
原子炉スクラムに係り、特にスクラム不能過渡事象時に
対応する原子炉出力制御装置に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a reactor scrum in a boiling water reactor, and more particularly to a reactor power controller for a non-scram transient event.

【０００２】[0002]

【従来の技術】一般に沸騰水型原子炉では、運転状態に
おいて何らかの理由で給水流量が喪失した場合には、原
子炉水位が低下して原子炉水位低位置（Ｌ−３）に達す
ると原子炉をスクラムさせる。これにより原子炉出力が
急減して、原子炉圧力容器から主蒸気管を通って流出す
る蒸気流量が減少して原子炉水位の低下は緩慢になる。
しかしながら、原子炉はスクラムしても炉心では崩壊熱
が発生しているため、定格蒸気流量に対して約５％の蒸
気発生は続くことになる。2. Description of the Related Art Generally, in a boiling water reactor, when the feed water flow rate is lost for some reason in an operating state, the reactor water level lowers and reaches a reactor water level low position (L-3). To scrum. As a result, the reactor output decreases sharply, the flow rate of steam flowing out from the reactor pressure vessel through the main steam pipe decreases, and the decrease in the reactor water level slows down.
However, since the decay heat is generated in the core of the nuclear reactor even if it is scrammed, about 5% of the rated steam flow rate will continue to be generated.

【０００３】したがって、給水流量の確保ができなけれ
ば、原子炉においては流入量と流出量のミスマッチによ
り原子炉水位は更に低下し、遂には原子炉水位低々位置
（Ｌ−２）に到達する。また、この原子炉水位の低々位
置（Ｌ−２）到達により主蒸気隔離弁が閉止し、原子炉
隔離時冷却系や高圧注水系の緊急炉心冷却装置が自動起
動して、外部より冷却水を注入して原子炉水位のこれ以
上の低下を防止する。Therefore, if the supply water flow rate cannot be secured, the reactor water level further decreases due to the mismatch between the inflow amount and the outflow amount in the reactor, and finally reaches the reactor water level low position (L-2). . When the reactor water level reaches the low position (L-2), the main steam isolation valve closes, and the reactor core isolation cooling system and the high pressure water injection system emergency core cooling system are automatically started, and the cooling water is supplied from the outside. To prevent further reduction in reactor water level.

【０００４】一方、前記主蒸気隔離弁の閉止に伴い、崩
壊熱によって発生した蒸気により原子炉圧力が上昇し
て、主蒸気管に設けられた主蒸気逃し安全弁の設定圧に
達し、主蒸気逃し安全弁が開放されて、この主蒸気逃し
安全弁と排気管を通り、発生蒸気の一部が圧力抑制室へ
放出されて原子炉圧力は低下する。このようにして、原
子炉スクラムに際しては、原子炉圧力及び原子炉水位
が、安全に制御されるようになっている。On the other hand, with the closing of the main steam isolation valve, the steam pressure generated by the decay heat raises the reactor pressure to reach the set pressure of the main steam relief safety valve provided in the main steam pipe, and the main steam escape valve is released. The safety valve is opened, and a part of the generated steam is discharged to the pressure suppression chamber through the main steam relief safety valve and the exhaust pipe, and the reactor pressure is reduced. In this way, the reactor pressure and the reactor water level can be safely controlled during the reactor scram.

【０００５】しかしながら、現実に発生する確率は非常
に低いが、若しも前記原子炉スクラムに失敗した場合に
ついては次のことが考えられる。通常の運転状態から給
水流量が喪失した場合には、原子炉水位は低下し、原子
炉水位低位置（Ｌ−３）により原子炉スクラム信号が発
生するが、この時に何らかの原因で全制御棒の挿入に失
敗したと仮定する。However, although the probability of actually occurring is very low, the following can be considered when the above-mentioned nuclear reactor scrum fails. When the feedwater flow rate is lost from the normal operating state, the reactor water level will drop and a reactor scrum signal will be generated due to the reactor water level low position (L-3). Suppose the insertion failed.

【０００６】これにより蒸気発生量が減少せず、原子炉
水位は更に低下して原子炉水位は低々位置（Ｌ−２）に
到達し、主蒸気隔離弁が閉止する。また、原子炉隔離時
冷却系や緊急炉心冷却装置の自動起動信号が発生するた
めに、原子炉に注水が開始されて原子炉水位は確保され
るが、そのために炉心流量が確保されて原子炉出力の抑
制が不十分になり、圧力抑制室のサプレッションプール
への熱負荷が大きくなる。As a result, the steam generation amount does not decrease, the reactor water level further decreases, the reactor water level reaches the low position (L-2), and the main steam isolation valve closes. Also, since the automatic start signal for the reactor isolation cooling system and the emergency core cooling system is generated, water injection is started in the reactor to secure the reactor water level. The output is insufficiently suppressed, and the heat load on the suppression pool of the pressure suppression chamber increases.

【０００７】したがって、これは原子炉格納容器の健全
性を十分に確保するという観点からは不十分である。そ
こで、このような事態に至るのを避けるために従来から
以下のような手段が考えられていた。すなわち、スクラ
ム不能過渡事象発生当初の原子炉出力を低下させて、発
生蒸気量を少しでも抑えるために、原子炉スクラム不能
時には原子炉再循環ポンプをトリップさせて炉心流量を
減少させる。Therefore, this is insufficient from the viewpoint of ensuring the soundness of the containment vessel. Therefore, in order to avoid such a situation, the following means have been conventionally considered. That is, in order to reduce the reactor power at the beginning of the occurrence of a transient unscrambling event and suppress the amount of generated steam as much as possible, the reactor recirculation pump is tripped when the reactor scram is disabled to reduce the core flow rate.

【０００８】また、給水ポンプをランバックして原子炉
水位を低下させ、自然循環力を低下して炉心流量を絞
り、原子炉出力を抑制させるようにしてスクラム不能過
渡事象発生当初の原子炉出力を抑制するようにしてい
る。さらに、これ以降は運転員の操作により原子炉への
注水流量を制御することにより原子炉水位をできるだけ
低く保ち、自然循環流量を最少にすることにより圧力抑
制室への熱負荷を最小限にするようにしていた。Further, the feed water pump is run back to lower the reactor water level, the natural circulation force is reduced to reduce the core flow rate, and the reactor output is suppressed so that the reactor output at the beginning of the occurrence of a non-scram transient event occurs. Is trying to suppress. Furthermore, from this point onward, the operator's operation controls the flow rate of water injected into the reactor to keep the reactor water level as low as possible, and the natural circulation flow rate is minimized to minimize the heat load to the pressure suppression chamber. Was doing.

【０００９】[0009]

【発明が解決しようとする課題】原子炉スクラム不能時
において、例えば期間はかなり限定されるが初装荷炉心
のサイクル早期のように、ボイド係数の絶対値が非常に
小さい場合には、上記のような原子炉出力を抑制する方
法は、あまり効果的には働かない。これは原子炉水位を
低下させて自然循環流量、すなわち炉心流量を減少して
ボイド率を増加させても、ボイド係数の絶対値が小さい
ために負の反応度が効果的に投入されないからである。
したがって、従来考えられていた方法を用いても原子炉
出力を効果的に抑制することができないという支障があ
る。If the absolute value of the void coefficient is very small, as in the early cycle of the initial loading core, for example, when the reactor scram cannot be performed, the period is considerably limited, but as described above. The method of suppressing a large reactor power does not work very effectively. This is because even if the reactor water level is lowered and the natural circulation flow, that is, the core flow is reduced and the void fraction is increased, the negative reactivity is not effectively injected because the absolute value of the void coefficient is small. .
Therefore, there is a problem that the reactor power cannot be effectively suppressed even by using the conventionally conceived method.

【００１０】例えば、ボイド反応度係数の絶対値が小さ
い（例えば、−３¢/%V)時に、自然循環状態まで炉心流
量を減少させても、原子炉出力は約40％までしか低下し
ない。しかし、現状設備の緊急炉心冷却装置の注水流量
は定格圧力近傍では、ほぼ10％定格給水流量しかない。
たとえ、原子炉隔離時冷却系の流量を加えても13％定格
給水流量程度である。For example, when the absolute value of the void reactivity coefficient is small (for example, -3 ¢ /% V), even if the core flow rate is reduced to the natural circulation state, the reactor power decreases only to about 40%. However, the injection flow rate of the emergency core cooling system of the current equipment is only 10% of the rated feed water flow rate near the rated pressure.
Even if the flow rate of the cooling system during reactor isolation is added, it is about 13% rated feed water flow rate.

【００１１】したがって原子炉水位を維持するため、す
なわち注水流量と発生蒸気量と釣り合わせるためには、
出力を13％程度に抑制するか、または注水流量の容量の
増加を図るしかない。しかし原子炉水位を十分確保でき
なかった場合には、炉心の健全性に重要な影響を及ぼす
という問題があった。Therefore, in order to maintain the reactor water level, that is, in order to balance the water injection flow rate with the generated steam quantity,
There is no choice but to suppress the output to about 13% or increase the capacity of the water injection flow rate. However, there was a problem that if the reactor water level could not be secured sufficiently, it would have an important effect on the integrity of the reactor core.

【００１２】本発明の目的とするところは、ボイド反応
度係数の絶対値の小さい時の原子炉スクラム不能過渡事
象に際して、先行的かつ自動的に主蒸気逃し安全弁を作
動させて原子炉を減圧し、緊急炉心冷却装置からの冷却
水注入を促進させることにより、極めて限定された場合
ではあるが炉心流量によって原子炉出力を十分抑制でき
ない場合でも、原子炉水位を確保して炉心の健全性を維
持することが可能な原子炉出力制御装置を提供すること
にある。The object of the present invention is to depressurize the reactor by actuating the main steam relief safety valve in advance and automatically in the event of a transient reactor scram impossible event when the absolute value of the void reactivity coefficient is small. , By promoting the injection of cooling water from the emergency core cooling system, the reactor water level is secured and the integrity of the reactor core is maintained even if the reactor output cannot be sufficiently suppressed by the reactor flow rate, although it is extremely limited. It is intended to provide a reactor power control device capable of doing so.

【００１３】[0013]

【課題を解決するための手段】上記目的を達成するため
請求項１記載の発明に係る原子炉出力制御装置は、原子
炉出力信号とスクラム作動要求信号を入力としてスクラ
ム不能過渡事象発生を判定するスクラム不能過渡事象判
定部と、前記スクラム不能過渡事象判定部のスクラム不
能過渡事象信号及び原子燃料の燃焼度信号を入力して動
特性ボイド係数信号を出力する動特性ボイド反応度算出
部を備える。In order to achieve the above object, the reactor power control apparatus according to the invention of claim 1 determines the occurrence of a non-scram transient event by inputting a reactor output signal and a scrum operation request signal. A scram impossible transient event determining section and a dynamic characteristic void reactivity calculating section for inputting the scram impossible transient event signal of the scram impossible transient event determining section and the nuclear fuel burnup signal and outputting a dynamic characteristic void coefficient signal are provided.

【００１４】さらに、前記原子炉出力信号と動特性ボイ
ド係数信号、及びスクラム不能事象発生時の炉心流量信
号とから予測原子炉出力信号を出力する原子炉出力予測
部と、前記原子炉出力予測部からの予測原子炉出力信号
を入力して、主蒸気逃し安全弁の開設定圧力変更信号及
び閉設定圧力変更信号を出力する主蒸気逃し安全弁開閉
設定圧変更部と、前記主蒸気逃し安全弁の開設定圧力変
更信号及び閉設定圧力変更信号を入力して主蒸気逃し安
全弁の開閉設定圧力の制御をする主蒸気逃し安全弁開閉
設定圧制御部とからなることを特徴とする。Further, a reactor output predicting section for outputting a predicted reactor output signal from the reactor output signal, the dynamic characteristic void coefficient signal, and the core flow rate signal at the time of a non-scram event, and the reactor output predicting section. Input the predicted reactor output signal from the main steam relief safety valve to output the open set pressure change signal and the closed set pressure change signal of the main steam relief safety valve opening / closing setting pressure change unit and the main steam relief safety valve open setting It is characterized by comprising a main steam relief safety valve opening / closing set pressure control unit for inputting a pressure change signal and a closed set pressure change signal to control the opening / closing set pressure of the main steam relief safety valve.

【００１５】請求項２記載の発明に係る原子炉出力制御
装置は、前記原子炉出力予測部における原子炉出力の予
測を、動特性ボイド反応度算出部からの動特性ボイド係
数信号と、各動特性ボイド反応度係数毎に用意された原
子炉水位対炉心流量曲線と各動特性ボイド反応度係数と
スクラム不能事象発生時の原子炉出力と炉心流量で規定
されるところの炉心状態毎に予め作成されていた原子炉
出力対炉心流量曲線とを用いて、炉心冠水を確保できる
最低原子炉水位に対応する原子炉出力を予測することを
特徴とする。According to a second aspect of the present invention, there is provided a reactor power control device for predicting a reactor power output in the reactor power output predicting section, using a dynamic characteristic void coefficient signal from a dynamic characteristic void reactivity calculating section and each dynamic coefficient. Reactor water level vs. core flow rate curve prepared for each characteristic void reactivity coefficient and each dynamic characteristic void reactivity coefficient, prepared in advance for each core state defined by reactor output and core flow rate at the time of non-scram event It is characterized by predicting the reactor power corresponding to the minimum reactor water level at which core flooding can be secured by using the existing reactor power vs. core flow rate curve.

【００１６】請求項３記載の発明に係る原子炉出力制御
装置は、前記主蒸気逃し安全弁開閉設定圧変更部で設定
する主蒸気逃し安全弁の開閉設定圧力が、前記原子炉出
力予測部において予測したスクラム不能過渡事象時に炉
心冠水を確保できる最低原子炉水位に対応する原子炉出
力と、その原子炉出力と高圧注水系の注水圧力特性より
必要な注水流量を得られる原子炉圧力を算出して注水流
量を最適にする原子炉圧力を設定することを特徴とす
る。In the reactor output control device according to the third aspect of the present invention, the opening / closing set pressure of the main steam release safety valve set by the main steam release safety valve opening / closing set pressure changing unit is predicted by the reactor output predicting unit. Injecting water by calculating the reactor pressure that can obtain the required injection flow rate from the reactor output that corresponds to the minimum reactor water level that can secure core flooding during a non-scram transient event, and the injection pressure characteristics of that reactor output and the high-pressure injection system It is characterized by setting the reactor pressure that optimizes the flow rate.

【００１７】[0017]

【作用】請求項１記載の発明は、スクラム不能過渡事象
判定部において原子炉出力信号とスクラム作動要求信号
とからスクラム不能過渡事象と判定すると、このスクラ
ム不能過渡事象信号及び原子燃料の燃焼度信号により動
特性ボイド反応度算出部では動特性ボイド反応度係数を
算出する。According to the first aspect of the present invention, when the scram non-transient event determining unit determines from the reactor output signal and the scram operation request signal that the scram non-transient event occurs, the scram non-transient event signal and the nuclear fuel burnup signal Thus, the dynamic characteristic void reactivity calculation section calculates the dynamic characteristic void reactivity coefficient.

【００１８】さらに原子炉出力予測部において動特性ボ
イド反応度係数及び炉心流量等から炉心を冠水させるの
に必要な最低原子炉水位を確保に対応する原子炉出力を
予測し、この時の原子炉発生蒸気流量に相当する予測原
子炉出力信号を主蒸気逃し安全弁開閉設定圧変更部に出
力する。Further, the reactor power predicting unit predicts the reactor power corresponding to securing the minimum reactor water level necessary to submerge the core from the dynamic characteristic void reactivity coefficient, the core flow rate, etc., and the reactor power at this time is predicted. The predicted reactor output signal corresponding to the generated steam flow rate is output to the main steam relief safety valve opening / closing set pressure changing unit.

【００１９】主蒸気逃し安全弁開閉設定圧変更部では、
前記予測原子炉出力信号から主蒸気逃し安全弁の開閉設
定圧を算出し、主蒸気逃し安全弁の開閉設定変更信号を
主蒸気逃し安全弁開閉設定圧制御部に出力する。これに
より主蒸気逃し安全弁は、炉心を冠水させるのに必要な
最低原子炉水位が確保されるように開閉設定圧力が主蒸
気逃し安全弁開閉設定圧制御部により制御されることか
ら、原子炉の健全性が維持される。In the main steam relief safety valve opening / closing set pressure changing section,
An opening / closing set pressure of the main steam relief safety valve is calculated from the predicted reactor output signal, and an opening / closing setting change signal of the main steam relief safety valve is output to the main steam relief safety valve opening / closing set pressure control unit. As a result, the main steam relief safety valve controls the opening / closing set pressure by the main steam relief safety valve opening / closing pressure control unit so that the minimum reactor water level necessary to submerge the core is secured. Sex is maintained.

【００２０】請求項２記載の発明では、前記原子炉出力
予測部における原子炉出力の予測を、動特性ボイド反応
度算出部からの動特性ボイド係数と、各動特性ボイド反
応度係数毎に用意された原子炉水位対炉心流量曲線と各
動特性ボイド反応度係数とスクラム不能事象発生時の原
子炉出力と炉心流量で規定されるところの炉心状態毎に
予め作成されていた原子炉出力対炉心流量曲線とを用い
て、炉心冠水を確保できる最低原子炉水位に対応する原
子炉出力を出力する。According to the second aspect of the present invention, the prediction of the reactor power in the reactor power predicting section is prepared for each of the dynamic characteristic void coefficient from the dynamic characteristic void reactivity calculating section and each dynamic characteristic void reactivity coefficient. Reactor water level vs. core flow rate curve and dynamic characteristics Void reactivity coefficient, reactor power at the time of non-scram event and reactor power vs. core created in advance for each core state defined by core flow rate The flow rate curve is used to output the reactor power corresponding to the minimum reactor water level that can secure core flooding.

【００２１】請求項３記載の発明では、主蒸気逃し安全
弁開閉設定圧変更部で設定する主蒸気逃し安全弁の開閉
設定圧力は、前記原子炉出力予測部において予測したス
クラム不能過渡事象時に炉心冠水を確保できる最低原子
炉水位に対応する原子炉出力と、その原子炉出力と高圧
注水系の注水圧力特性より必要な注水流量を得られる原
子炉圧力を算出して注水流量を最適にする。According to the third aspect of the present invention, the opening / closing set pressure of the main steam relief safety valve set by the main steam relief safety valve opening / closing set pressure changing unit is set to the core flooding at the time of the non-scram transient event predicted by the reactor output predicting unit. The reactor output corresponding to the minimum reactor water level that can be secured and the reactor output and the injection pressure characteristics of the high-pressure injection system are used to calculate the reactor pressure that can obtain the required injection flow rate, and the injection flow rate is optimized.

【００２２】[0022]

【実施例】本発明の一実施例を図面を参照して説明す
る。全体構成は図１のブロック構成図に示すように、原
子炉出力信号Ｓ1 及びスクラム作動要求信号Ｓ2 を入力
としてスクラム不能過渡事象が生じたことを判定し、ス
クラム不能過渡事象信号Ｓ3を発するスクラム不能過渡
事象判定部１と、このスクラム不能過渡事象信号Ｓ3及
びプロセスコンピュータからの原子燃料の燃焼度を示す
燃焼度信号Ｓ4 を入力して、動特性ボイド係数信号Ｓ5
を出力する動特性ボイド反応度算出部２を備える。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to the drawings. As shown in the block diagram of FIG. 1, the overall configuration is such that the reactor output signal S 1 and the scram operation request signal S 2 are input to determine that a scram-unusable transient event has occurred, and the scram-unusable transient event signal S 3 is output. The non-scram transient event determination unit 1 to be issued, the scram non-transient event signal S 3 and the burnup signal S 4 indicating the burnup of the nuclear fuel from the process computer are input to input the dynamic characteristic void coefficient signal S 5
The dynamic characteristic void reactivity calculation unit 2 for outputting

【００２３】また、前記原子炉出力信号Ｓ1 、動特性ボ
イド係数信号Ｓ5 及びスクラム不能事象発生時の炉心流
量信号Ｓ6 とから予測原子炉出力信号Ｓ7 を出力する原
子炉出力予測部３と、この原子炉出力予測部３からの予
測原子炉出力信号Ｓ7 を入力して、主蒸気逃し安全弁開
設定圧力変更信号Ｓ10及び主蒸気逃し安全弁閉設定圧力
変更信号Ｓ11を出力する主蒸気逃し安全弁開閉設定圧変
更部４を設ける。Further, a reactor output predicting section 3 for outputting a predicted reactor output signal S 7 from the reactor output signal S 1 , the dynamic characteristic void coefficient signal S 5 and the core flow rate signal S 6 at the time of an unscrambling event. And the predicted reactor output signal S 7 from the reactor output prediction unit 3 are input to output the main steam relief safety valve open setting pressure change signal S 10 and the main steam relief safety valve close setting pressure change signal S 11. A steam relief safety valve opening / closing set pressure changing unit 4 is provided.

【００２４】さらに、前記主蒸気逃し安全弁開設定圧力
変更信号Ｓ10及び主蒸気逃し安全弁閉設定圧力変更信号
Ｓ11を入力して主蒸気逃し安全弁の開閉設定圧力を制御
する主蒸気逃し安全弁開閉設定圧制御部５とで構成され
ている。Further, the main steam relief safety valve open setting pressure change signal S 10 and the main steam relief safety valve close setting pressure change signal S 11 are input to control the main steam relief safety valve open / close setting pressure. It is composed of a pressure controller 5.

【００２５】次に上記構成による作用について説明す
る。原子炉定格出力運転時に何らかの理由により全給水
流量喪失事象が発生すると原子炉水位は低下して、原子
炉水位低位置（Ｌ−３）でスクラム要求信号が発生する
が、何らかの原因で全制御棒挿入に失敗すると仮定す
る。Next, the operation of the above configuration will be described. If the total feedwater flow rate loss event occurs for some reason during the reactor rated output operation, the reactor water level will drop and a scrum request signal will be generated at the reactor water level low position (L-3), but for some reason all control rods will be generated. Suppose the insertion fails.

【００２６】このときに原子炉出力制御装置において、
スクラム不能過渡事象判定部１は、原子炉出力信号Ｓ1
とスクラム作動要求信号Ｓ2 を入力してスクラム不能過
渡事象が生じたことを判定するが、スクラム不能過渡事
象判定部１においては、図２のロジック図に示すように
原子炉出力信号Ｓ1 を入力して、予め定められた出力レ
ベルＰ0 （例えば、定格出力の３％）以上の成立信号
と、スクラム要求信号Ｓ2 のＡＮＤ条件が成立すると、
スクラム不能過渡発生と判定してスクラム不能過渡事象
信号Ｓ3 を発する。At this time, in the reactor power control device,
The non-scrum transient event determination unit 1 determines the reactor output signal S 1
And a scram operation request signal S 2 are input to determine that a non-scrum transient event has occurred. In the scram non-transient event determination unit 1, the reactor output signal S 1 is output as shown in the logic diagram of FIG. When the AND condition of the input signal satisfying the predetermined output level P 0 (for example, 3% of the rated output) or more and the scrum request signal S 2 is satisfied,
A non-scrum transient event signal S 3 is generated when it is determined that a non-scram transient has occurred.

【００２７】スクラム不能過渡事象信号Ｓ3 を入力した
動特性ボイド反応度算出部２では、図示しないプロセス
コンピュータからの原子燃料の燃焼度（Ｂn ）を示す燃
焼度信号Ｓ4 を入力して、予め用意しておいた、図３の
燃焼度対動特性ボイド反応度係数特性図に示すボイド反
応度係数曲線により動特性ボイド係数（Ｖn ）を算出
し、動特性ボイド係数信号Ｓ5 を原子炉出力予測部３に
出力する。In the dynamic characteristic void reactivity calculation section 2 to which the non-scram transient event signal S 3 is input, the burnup signal S 4 indicating the burnup (B n ) of the nuclear fuel is input from a process computer (not shown). prepared in advance to calculate a dynamic characteristic void coefficient (V n) by the void reactivity coefficient curve shown in burnup to dynamic characteristic void reactivity coefficient characteristic diagram of FIG. 3, the dynamic characteristic void coefficient signal S 5 atomic Output to the furnace output predicting unit 3.

【００２８】この原子炉出力予測部３では、これも予め
作成しておいた図４の原子炉水位対炉心流量特性図に示
す、各動特性ボイド反応度係数（Ｖ0 〜Ｖn ）毎に用意
されている原子炉水位対炉心流量の曲線と、図５の炉心
流量対原子炉出力特性図に示す、各動特性ボイド反応度
係数とスクラム不能事象発生時の原子炉出力と、炉心流
量で規定されるところの炉心状態（Ｔ0 〜Ｔn ）毎に予
め作成されていたスクラム不能事象発生時の炉心状態曲
線とを用いる。In the reactor power predicting section 3, for each dynamic characteristic void reactivity coefficient (V 0 to V n ) shown in the reactor water level versus core flow rate characteristic diagram of FIG. 4 which is also prepared in advance. In the prepared curve of reactor water level vs. core flow rate, and core flow rate vs. reactor output characteristic diagram of Fig. 5, the dynamic characteristic void reactivity coefficient, the reactor output at the time of non-scram event, and the core flow rate are shown. A core state curve at the time of occurrence of a scrum-disabled event that is created in advance for each core state (T 0 to T n ) that is defined is used.

【００２９】これにより、スクラム不能事象発生時の原
子炉出力信号Ｓ1 、炉心流量信号Ｓ6 及び動特性ボイド
係数信号Ｓ5 とから、原子炉を冠水できる最低水位位置
（Ｌo ）に対応する原子炉出力（Ｐo ）を予測し、予測
原子炉出力信号Ｓ7 を主蒸気逃し安全弁開閉設定圧変更
部４に出力する。Thus, from the reactor output signal S 1 , the core flow rate signal S 6 and the dynamic characteristic void coefficient signal S 5 at the time of the occurrence of a non-scrum event, the lowest water level position ( Lo ) at which the reactor can be submerged is corresponded. The reactor output (P o ) is predicted, and the predicted reactor output signal S 7 is output to the main steam relief safety valve opening / closing set pressure changing unit 4.

【００３０】この時の原子炉出力（Ｐo ）の推定方法を
図４と図５により説明する。先ず、図４の原子炉水位対
炉心流量曲線により、前記動特性ボイド反応度算出部２
より求めた動特性ボイド反応度係数（Ｖn ）と、原子炉
冠水最低水位（Ｌo ）より原子炉水位対炉心流量曲線か
ら対応する炉心流量（Ｗo ）を求める。A method of estimating the reactor power (P o ) at this time will be described with reference to FIGS. 4 and 5. First, based on the reactor water level versus core flow rate curve of FIG.
The corresponding core flow rate (W o ) is calculated from the reactor water level versus core flow rate curve from the dynamic characteristic void reactivity coefficient (V n ) thus obtained and the minimum reactor flood water level (L o ).

【００３１】次に、今求めた炉心流量（Ｗo ）と動特性
ボイド反応度係数（Ｖn ）と、更に原子炉出力（Ｐn ）
と炉心流量（Ｗn ）から、スクラム不能事象発生時の炉
心状態ｎを同定し、図５に示すように炉心状態ｎに対応
する原子炉出力対炉心流量曲線（Ｔn ）から炉心冠水最
低水位（Ｌo ）に対応する原子炉出力（Ｐo ）を予測す
る。Next, the obtained core flow rate (W o ) and dynamic characteristic void reactivity coefficient (V n ), and further the reactor power (P n ).
From core flow (W n), to identify core state n during scram impossible event occurs, the core flood low water from the reactor output pair core flow rate curve corresponding to the core state n (T n), as shown in FIG. 5 Predict the reactor power (P o ) corresponding to (L o ).

【００３２】主蒸気逃し安全弁の開閉設定圧変更部４で
は、図６のロジック図に示すように前記原子炉出力予測
部３からの予測原子炉出力信号Ｓ7 （原子炉発生蒸気流
量相当信号、すなわち、これが必要とする高圧注水系の
緊急炉心冷却装置による注水流量となる。）を入力し
て、例えば図７の緊急炉心冷却装置の冷却水注入特性図
に示す、注入特性曲線に従って原子炉圧力を計算し、こ
の時の原子炉圧力信号Ｓ8 、すなわち、主蒸気逃し安全
弁の開閉設定圧を算出する信号を出力する。In the main steam relief safety valve opening / closing set pressure changing unit 4, as shown in the logic diagram of FIG. 6, the predicted reactor output signal S 7 (reactor generated steam flow equivalent signal, That is, this is the required water injection flow rate by the emergency core cooling device of the high-pressure water injection system.), For example, according to the injection characteristic curve shown in the cooling water injection characteristic diagram of the emergency core cooling device of FIG. And the reactor pressure signal S 8 at this time, that is, a signal for calculating the opening / closing set pressure of the main steam relief safety valve is output.

【００３３】この時に無駄な主蒸気逃し安全弁の開放を
防ぐため、図６に示すように、緊急炉心冷却装置からの
必要な冷却水注入量を維持するための原子炉圧力信号Ｓ
8 と、スクラム不能過渡事象発生時の原子炉圧力信号Ｓ
9 を比較する。At this time, in order to prevent the unnecessary opening of the main steam relief safety valve, as shown in FIG. 6, the reactor pressure signal S for maintaining the required cooling water injection amount from the emergency core cooling device is shown.
8 and the reactor pressure signal S when a non-scrum transient event occurs
Compare 9

【００３４】この結果からＳ8 ＜Ｓ9 が成立したときの
みに、算出された圧力信号Ｓ8 を算出設定圧力として入
力し、主蒸気逃し安全弁の開閉設定圧を算出設定圧力の
例えば、±５％（開設定圧が算出設定圧の＋５％、閉設
定圧が算出設定圧の−５％）に設定し、主蒸気逃し安全
弁開設定圧力変更信号Ｓ10と、主蒸気逃し安全弁閉設定
圧力変更信号Ｓ11を主蒸気逃し安全弁開閉設定圧制御部
５へ出力する。From this result, only when S 8 <S 9 is satisfied, the calculated pressure signal S 8 is input as the calculated set pressure, and the open / close set pressure of the main steam relief safety valve is set to ± 5 of the calculated set pressure. % (+ 5% open pressure is calculated set pressure,閉設-5% of constant pressure is calculated set pressure) is set to, main steam relief and safety valve open set pressure change signal S 10, main steam safety relief valve closed set pressure change signal S 11 is output to the main steam relief safety valve opening / closing set pressure control unit 5.

【００３５】このようにして本発明では、主蒸気逃し安
全弁の開放設定圧を、予めスクラム不能過渡事象時に炉
心冠水を確保するため、特に定めた原子炉水位に見合っ
た圧力に自動設定することで、原子炉圧力の制御と共に
原子炉への冷却水の注水流量も自動的に制御されるの
で、原子炉が必要とする水位も自動的に制御される。As described above, according to the present invention, the opening set pressure of the main steam relief safety valve is automatically set to a pressure corresponding to a predetermined reactor water level in advance in order to secure core flooding in advance in the event of a scram incapability. Since the flow rate of cooling water injected into the reactor is automatically controlled together with the control of the reactor pressure, the water level required by the reactor is also automatically controlled.

【００３６】これにより、ボイド反応度係数の絶対値の
小さいときの全制御棒挿入失敗等によるスクラム不能過
渡事象時でも、既存の設備である主蒸気逃し安全弁を自
動的に作動させて、原子炉出力及び原子炉水位を自動制
御して炉心の健全性を確保することができる。As a result, the main steam relief safety valve, which is an existing facility, is automatically operated to automatically operate the reactor even at the time of a transient event in which a scrum cannot be performed due to a failure in inserting all control rods when the absolute value of the void reactivity coefficient is small. Power integrity and reactor water level can be automatically controlled to ensure core integrity.

【００３７】[0037]

【発明の効果】以上本発明によれば、ボイド反応度係数
の絶対値の小さいときにおける全制御棒挿入失敗のスク
ラム不能過渡事象においても、原子炉圧力と原子炉出
力、及び原子炉水位を自動的に制御し、炉心の健全性を
既存の設備を用いて確保できると共に、運転員の操作に
よらず原子炉は安全に制御できることから、運転員の負
担も軽減して、原子炉の安全性と運転の信頼性が向上す
る効果がある。As described above, according to the present invention, the reactor pressure, the reactor power, and the reactor water level are automatically adjusted even in the non-scrum transient event of the failure of all the control rod insertions when the absolute value of the void reactivity coefficient is small. Can be controlled to ensure the integrity of the reactor core using existing equipment, and the reactor can be safely controlled regardless of the operator's operation, reducing the burden on the operator and ensuring reactor safety. And there is an effect that the reliability of driving is improved.

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

【図１】本発明に係る一実施例の原子炉出力制御装置の
ブロック構成図。FIG. 1 is a block configuration diagram of a reactor power control device according to an embodiment of the present invention.

【図２】本発明に係る一実施例のスクラム不能過渡事象
判定部のロジック図。FIG. 2 is a logic diagram of a non-scrum transient event determination unit according to an embodiment of the present invention.

【図３】本発明に係る一実施例の燃焼度対動特性ボイド
反応度係数特性図。FIG. 3 is a burn-up dynamic characteristic void reactivity coefficient characteristic diagram of an embodiment according to the present invention.

【図４】本発明に係る一実施例の原子炉水位対炉心流量
特性図。FIG. 4 is a characteristic diagram of reactor water level versus core flow rate according to an embodiment of the present invention.

【図５】本発明に係る一実施例の炉心流量対原子炉出力
特性図。FIG. 5 is a core flow rate vs. reactor power characteristic diagram of one embodiment according to the present invention.

【図６】本発明に係る一実施例の主蒸気逃し安全弁開閉
設定圧変更部のロジック図。FIG. 6 is a logic diagram of a main steam relief safety valve opening / closing set pressure changing unit according to an embodiment of the present invention.

【図７】本発明に係る一実施例の緊急炉心冷却装置の冷
却水注入特性図。FIG. 7 is a cooling water injection characteristic diagram of the emergency core cooling device according to the embodiment of the present invention.

【符号の説明】[Explanation of symbols]

１…スクラム不能過渡事象判定部、２…動特性ボイド反
応度算出部、３…原子炉出力予測部、４…主蒸気逃し安
全弁開閉設定圧変更部、５…主蒸気逃し安全弁開閉設定
圧制御部、Ｓ1 …原子炉出力信号、Ｓ2 …スクラム要求
信号、Ｓ3 …スクラム不能過渡事象信号、Ｓ4 …燃焼度
信号、Ｓ5 …動特性ボイド係数信号、Ｓ6 …炉心流量信
号、Ｓ7 …予測原子炉出力信号、Ｓ8 …原子炉圧力信
号、Ｓ9 …スクラム不能過渡事象発生時の原子炉圧力信
号、Ｓ10…主蒸気逃し安全弁開設定圧変更信号、Ｓ11…
主蒸気逃し安全弁閉設定圧変更信号。DESCRIPTION OF SYMBOLS 1 ... Scrum impossible transient event determination unit, 2 ... Dynamic characteristic void reactivity calculation unit, 3 ... Reactor output prediction unit, 4 ... Main steam relief safety valve opening / closing set pressure changing unit, 5 ... Main steam relief safety valve opening / closing set pressure control unit , S 1 ... Reactor output signal, S 2 ... Scrum request signal, S 3 ... Scrum non-transient event signal, S 4 ... Burnup signal, S 5 ... Dynamic characteristic void coefficient signal, S 6 ... Core flow signal, S 7 ... predicted reactor power signal, S 8 ... reactor pressure signal, S 9 ... scram non transient event occurs when the reactor pressure signal, S 10 ... main steam safety relief valve opening pressure change signal, S 11 ...
Main steam relief safety valve closing set pressure change signal.

Claims (3)

【特許請求の範囲】[Claims] 【請求項１】 原子炉出力信号とスクラム作動要求信号
を入力としてスクラム不能過渡事象発生を判定するスク
ラム不能過渡事象判定部と、前記スクラム不能過渡事象
判定部のスクラム不能過渡事象信号及び原子燃料の燃焼
度信号を入力して動特性ボイド係数信号を出力する動特
性ボイド反応度算出部と、前記原子炉出力信号と動特性
ボイド係数信号、及びスクラム不能事象発生時の炉心流
量信号とから予測原子炉出力信号を出力する原子炉出力
予測部と、前記原子炉出力予測部からの予測原子炉出力
信号を入力して主蒸気逃し安全弁の開設定圧力変更信号
及び閉設定圧力変更信号を出力する主蒸気逃し安全弁開
閉設定圧変更部と、前記主蒸気逃し安全弁の開設定圧力
変更信号及び閉設定圧力変更信号を入力して主蒸気逃し
安全弁の開閉設定圧力の制御をする主蒸気逃し安全弁開
閉設定圧制御部とからなることを特徴とする原子炉出力
制御装置。1. A scram non-transient event determination unit for determining the occurrence of a non-scram transient event by inputting a reactor output signal and a scram operation request signal, and a scram non-transient event signal and a nuclear fuel of the scram non-transient event determination unit. A predictive atom is obtained from a dynamic void reactivity calculation section that inputs a burnup signal and outputs a dynamic void coefficient signal, the reactor output signal and the dynamic void coefficient signal, and a core flow rate signal at the time of a non-scram event. A reactor output predicting unit that outputs a reactor output signal, and a main that outputs the predictive reactor output signal from the reactor output predicting unit and outputs an open set pressure change signal and a closed set pressure change signal of the main steam relief safety valve The opening / closing set pressure of the main steam relief safety valve is input by inputting the open / closed pressure change signal and the closing set pressure change signal of the steam relief safety valve opening / closing set pressure change section. A reactor output control device comprising a main steam relief safety valve opening / closing set pressure control unit for controlling force. 【請求項２】 前記原子炉出力予測部における原子炉出
力の予測は、動特性ボイド反応度算出部からの動特性ボ
イド係数信号と、各動特性ボイド反応度係数毎に用意さ
れた原子炉水位対炉心流量曲線と各動特性ボイド反応度
係数とスクラム不能事象発生時の原子炉出力と炉心流量
で規定されるところの炉心状態毎に予め作成されていた
原子炉出力対炉心流量曲線とを用いて、炉心冠水を確保
できる最低原子炉水位に対応する原子炉出力を予測する
ことを特徴とする請求項１記載の原子炉出力制御装置。2. The prediction of the reactor power in the reactor power prediction unit is performed by the dynamic void coefficient signal from the dynamic void reactivity calculation unit and the reactor water level prepared for each dynamic void reactivity coefficient. A core flow rate curve and each dynamic characteristic void reactivity coefficient, a reactor output at the time of a non-scram event, and a reactor output vs. core flow rate curve that was created in advance for each core state defined by the core flow rate are used. 2. The reactor power control system according to claim 1, wherein the reactor power corresponding to the minimum reactor water level that can secure the core flooding is predicted. 【請求項３】 前記主蒸気逃し安全弁開閉設定圧変更部
で設定する主蒸気逃し安全弁の開閉設定圧力は、前記原
子炉出力予測部において予測したスクラム不能過渡事象
時に炉心冠水を確保できる最低原子炉水位に対応する原
子炉出力と、その原子炉出力と高圧注水系の注水圧力特
性より必要な注水流量を得られる原子炉圧力を算出して
注水流量を最適にする原子炉圧力を設定することを特徴
とする請求項１記載の原子炉出力制御装置。3. The minimum reactor for which the main steam relief safety valve opening / closing set pressure set by the main steam relief safety valve opening / closing set pressure changing unit is capable of securing core flooding at the time of a scram impossible transient event predicted by the reactor output predicting unit. From the reactor power corresponding to the water level and the reactor output and the injection pressure characteristics of the high pressure injection system, calculate the reactor pressure that can obtain the required injection flow rate and set the reactor pressure that optimizes the injection flow rate. The reactor power control system according to claim 1, which is characterized in that.
JP5297710A 1993-11-29 1993-11-29 Nuclear reactor power controller Pending JPH07151899A (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
JP5297710A JPH07151899A (en) 1993-11-29 1993-11-29 Nuclear reactor power controller

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
JP5297710A JPH07151899A (en) 1993-11-29 1993-11-29 Nuclear reactor power controller

Publications (1)

Publication Number Publication Date
JPH07151899A true JPH07151899A (en) 1995-06-16

Family

ID=17850167

Family Applications (1)

Application Number Title Priority Date Filing Date
JP5297710A Pending JPH07151899A (en) 1993-11-29 1993-11-29 Nuclear reactor power controller

Country Status (1)

Country Link
JP (1) JPH07151899A (en)

Similar Documents

Publication Publication Date Title
EP0594384A1 (en) System and method for controlling a nuclear reactor upon steam flow control valve failure
JPWO2011074544A1 (en) Reactor transient mitigation system
JPH07151899A (en) Nuclear reactor power controller
CN111916233A (en) Small pressurized water reactor safety injection system combining passive and active functions
Vahman et al. Investigation of small break LOCA with failure of the emergency core cooling system accident management in VVER-1000 reactor type
Parzer et al. Feed-and-bleed procedure mitigating the consequences of a steam generator tube rupture accident
JP3871733B2 (en) Reactor core monitoring device
Sui et al. Analysis of single failure criteria and operator action effect for different SGTR mitigation strategies of HPR1000
Ha et al. Comparison of the decay heat removal systems in the KALIMER-600 and DSFR
Tusheva et al. Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor
JP3136786B2 (en) Reactor water level control method and apparatus
JPH063479A (en) Control equipment of reactor power
Prošek RELAP5/MOD3. 3 Simulation of LOFT LP-FW-1 Total Loss of Feedwater Test
Yue et al. Investigation the Fuel Damage Fraction of IB-LOCA
Bánáti et al. Simulation of some emergency operating procedure for VVER-440 reactors
JPH02264886A (en) Apparatus for output control of reactor
Leskovar et al. Analysis of Severe Accident in Safety Upgraded Krško NPP with MELCOR 2.2
JPS6128893A (en) Nuclear power plant
JPH09203793A (en) Test method of main steam isolation valve
JPH11211886A (en) Recovering method after full closure of main steam isolation valve
KR20220138755A (en) Control System for safety injection flow to prevent high level of steam generator
JPS6050318B2 (en) Reactor control device
JP2024139441A (en) Water supply control method and water supply control device
CODE THIRD INTERNATIONAL SEMINAR ON HORIZONTAL STEAM GENERATORS 18-20 October 1994, Lappeenranta, Finland
Bencik et al. RELAP5/MOD3. 3 analysis of the Reactor Coolant Pump Trip event at NPP Krsko for different transient scenarios