JPH01285897A - Automatic pressure reducing device of nuclear reactor - Google Patents
Automatic pressure reducing device of nuclear reactorInfo
- Publication number
- JPH01285897A JPH01285897A JP63113534A JP11353488A JPH01285897A JP H01285897 A JPH01285897 A JP H01285897A JP 63113534 A JP63113534 A JP 63113534A JP 11353488 A JP11353488 A JP 11353488A JP H01285897 A JPH01285897 A JP H01285897A
- Authority
- JP
- Japan
- Prior art keywords
- reactor
- pressure
- water
- signal
- safety valve
- 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
Links
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims abstract description 69
- 230000001629 suppression Effects 0.000 claims abstract description 9
- 238000002347 injection Methods 0.000 claims description 28
- 239000007924 injection Substances 0.000 claims description 28
- 230000006837 decompression Effects 0.000 description 5
- 238000010586 diagram Methods 0.000 description 4
- 241000272814 Anser sp. Species 0.000 description 3
- 230000000694 effects Effects 0.000 description 3
- 239000003758 nuclear fuel Substances 0.000 description 3
- 238000009835 boiling Methods 0.000 description 2
- 239000002826 coolant Substances 0.000 description 2
- 238000001816 cooling Methods 0.000 description 2
- 238000001514 detection method Methods 0.000 description 2
- UGFAIRIUMAVXCW-UHFFFAOYSA-N Carbon monoxide Chemical compound [O+]#[C-] UGFAIRIUMAVXCW-UHFFFAOYSA-N 0.000 description 1
- 241000282693 Cercopithecidae Species 0.000 description 1
- 241000257465 Echinoidea Species 0.000 description 1
- 230000009286 beneficial effect Effects 0.000 description 1
- 150000001875 compounds Chemical class 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
- 210000003127 knee Anatomy 0.000 description 1
- 238000013021 overheating Methods 0.000 description 1
- 238000011084 recovery Methods 0.000 description 1
- 150000003839 salts Chemical class 0.000 description 1
- 229910000679 solder Inorganic materials 0.000 description 1
- 239000007921 spray Substances 0.000 description 1
- 210000004243 sweat Anatomy 0.000 description 1
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
- Y02E30/30—Nuclear fission reactors
Landscapes
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
Description
【発明の詳細な説明】 ［発明の目的１ （産業上の利用分野） 本発明は原子炉の自動減圧装置に関する。[Detailed description of the invention] [Object of the invention 1 (Industrial application field) The present invention relates to an automatic decompression device for a nuclear reactor.
（従来の技術）
本発明が適用される沸騰水型１原子炉は、第４図に示す
ように、格納容器１内にＬモノノ容器２か配置されてｄ
３す、この圧力容器２内には炉心３と冷却（４４とが収
納されでいる。この汁ツノ容器２内の冷却！４ ／ｌ
ｌよ炉心３の核燃料により加熱されて蒸気となり、この
蒸気は格納容器１を貴通する主蒸気管５によってタービ
ン６に導かれる。(Prior Art) A boiling water type 1 nuclear reactor to which the present invention is applied, as shown in FIG.
3. Inside this pressure vessel 2, a reactor core 3 and cooling (44) are housed.
It is heated by the nuclear fuel in the reactor core 3 to become steam, and this steam is guided to a turbine 6 by a main steam pipe 5 passing through the containment vessel 1.
タービン６で仕事をした蒸気は］ンデン＋Ｊ７において
復水となり、圧力容器２に戻される。格納容器１の内部
は上部空間のドライウェル８と上部空間のすＩルッシ」
ンプ〜ル９とに区分されでいる。、また、主蒸気管５に
は必要に応じて主蒸気管内の蒸気を格納容器１内に放出
する逃し安全弁１０か、ｉ、Ｑ ｔｄ−１うれており、
配管１１を介してザブレッションプール９内の１−ル水
に達している。圧力容器２には、高圧注水系（高坏スプ
レィ系等も含む）１２を設りでいる。The steam that has done work in the turbine 6 becomes condensed water at Nden+J7 and is returned to the pressure vessel 2. The inside of the containment vessel 1 consists of a dry well 8 in the upper space and a dry well 8 in the upper space.
It is divided into groups 9 to 9. In addition, the main steam pipe 5 is provided with a relief safety valve 10 or i, Q td-1 for releasing steam in the main steam pipe into the containment vessel 1 as necessary.
It reaches the 1-liter water in the breechion pool 9 via the pipe 11. The pressure vessel 2 is equipped with a high-pressure water injection system (including a high-pressure spray system, etc.) 12.
原子炉の配管か破断じたような場合には、原子炉はスク
ラムされ、格納容器１は密閉隔離されるが、［原子炉の
冷却材は破断［−１より流出し、Ｌトカ容器２内の水位
は低下し炉心３の核燃料が露出され過熱破損する恐れが
ある。ぞのため、圧力容器２内の水位低下検出により高
圧注水系１２を作動させて、圧力容器２内に冷却材を注
入し核燃料の露出を防止するように構成されている。In the event that a reactor piping ruptures, the reactor is scrammed and the containment vessel 1 is hermetically isolated. The water level will drop and the nuclear fuel in the reactor core 3 will be exposed and there is a risk of overheating damage. Therefore, the high-pressure water injection system 12 is activated upon detection of a drop in the water level in the pressure vessel 2, and coolant is injected into the pressure vessel 2 to prevent nuclear fuel from being exposed.
低圧注水系１３は、高圧注水系１２のバックアップとし
て設けてあり、高圧注水系１２が圧力容器２内水位低下
検出にもかかわらず、何らかの原因で作動しなかった時
に起動されるように構成されている。The low-pressure water injection system 13 is provided as a backup for the high-pressure water injection system 12, and is configured to be activated when the high-pressure water injection system 12 does not operate for some reason despite detection of a drop in the water level in the pressure vessel 2. There is.
ところか、圧力容器２内の圧力か高い用台には低圧注水
系１３によって圧力容器２内に冷却材を注入り−ること
はできない。そのため、従来は第５図の論理回路に示す
ごとく、圧力容器内を減圧し、低圧注水系１３による注
水を可能としている。ずなわら、配管が破断してドライ
ウェル８内に高温高圧の蒸気が放出されている場合には
、トライウェル圧力が上背し高圧となっているので、ド
ライウェル圧力高の信号１４と、圧力容器水位低の信号
１５と、低圧注水系ポンプの運転信号１６との三者のア
間１νに白！Ｉ！ｌＩ減圧系１９を起動さ１！（−いた
１、自動減圧系は、多数の逃し安全弁１（）（図示の都
合上１箇のみ示し、−Ｕいる）のいくつかを開放するこ
とにＪ、す、Ｈ−ｈ打器２内の蒸気を逃し安全弁を介し
てサシレッジ」ンブール水に導かれ、ここで凝縮復水さ
１！でいたので、圧力容器？内は減圧されて低１１Ｔ注
水系１２による注入を可能と（−７でいた。However, if the pressure inside the pressure vessel 2 is high, coolant cannot be injected into the pressure vessel 2 by the low-pressure water injection system 13. Therefore, conventionally, as shown in the logic circuit of FIG. 5, the pressure inside the pressure vessel is reduced to enable water injection by the low-pressure water injection system 13. However, if the piping is broken and high-temperature, high-pressure steam is released into the dry well 8, the tri-well pressure will rise to a high level, so the dry well pressure high signal 14 will be White at 1ν between the pressure vessel water level low signal 15 and the low pressure water injection system pump operation signal 16! I! Activate the lI decompression system 19! (-1, automatic depressurization system, J, S, H-h inside the batter 2 by opening some of the many relief safety valves 1 () (only one is shown for convenience of illustration, -U). The steam was released and led to the sacilage water via a safety valve, where it condensed and condensed at 1!, so the pressure inside the pressure vessel was reduced to enable injection by the low 11T water injection system 12 (at -7). there was.
（発明が解決しようとする問題点）
ところで、従来の自動減ｉ［装置１９てｔよ次のＪ、う
な不都合か生じる。すなわち、自動減圧装置１９【。１
予め決められた多数の逃じ弁１０を開放するので、その
冷端は多く第３図（ａ）に小す如く、急速に減圧を行い
低圧系の注水を可能としている１、この操作は、蒸気を
サブレッジｄンＪ−ル水中に放出゛づるので、第３図（
１））に示すＪ、うに１ノブレツジコクプール水温を上
背させることとなる。サプレッションブール水温か高い
場合に（、策、この蒸゛気を凝縮リ−る効果は小さくな
るため、あまりプール水温を上背させることは望ましい
ことてはない。(Problems to be Solved by the Invention) By the way, the conventional automatic reduction device 19 causes the following inconvenience. That is, the automatic pressure reducing device 19 [. 1
Since a predetermined number of relief valves 10 are opened, there are many cold ends, and as shown in FIG. 3(a), the pressure is rapidly reduced and water injection into the low pressure system becomes possible1. This operation: Since the steam is released into the water in the subledge, Figure 3 (
1)) J, sea urchin 1 Noble Tsujikoku pool water temperature will be raised. If the suppression pool water temperature is high, the effect of condensing and releasing this steam will be reduced, so it is not desirable to raise the pool water temperature too much.
このため、減圧操作を行った後、サプレッションブール
水温かこの限界の温度に近づく場合がある。Therefore, after performing a pressure reduction operation, the suppression boule water temperature may approach this limit temperature.
本発明は上記事情に基づきなされたもので、その目的は
、原子炉事故時に圧力容器の急激な減圧を避けるように
した原子炉の自動減圧装置を提供することにある。The present invention has been made based on the above circumstances, and its object is to provide an automatic decompression device for a nuclear reactor that avoids rapid depressurization of a pressure vessel in the event of a nuclear reactor accident.
「発明の構成」
（問題点を解決するだめの手段および作用）本発明は、
Ｌ記目的を達成するために、原子炉格納容器内に設置さ
れた原子炉圧力容器と、前記原子炉圧力容器内の水位を
確保するために設置された注水系と、前記原子炉圧力容
器に接続された主蒸気管に設けた逃し安全弁と、前記原
子炉格納容器内に設けたザブレツション１−ルとからな
る原子炉において、原子炉水位、原子炉圧力、注水系流
量、サプレッションブール水温の各信号に基づいて予め
決められたしきい値と比較して前記逃し安全弁及び注水
系の圧力流量特性に基づき自動減圧系の起動信号を出力
するように構成したことを特徴どづ−るものである。"Structure of the invention" (Means and effects for solving the problem) The present invention includes:
In order to achieve the purpose of Section L, a reactor pressure vessel installed in the reactor containment vessel, a water injection system installed to secure the water level in the reactor pressure vessel, and a water injection system installed in the reactor pressure vessel In a nuclear reactor consisting of a safety relief valve provided in the connected main steam pipe and a suppression valve provided in the reactor containment vessel, each of the reactor water level, reactor pressure, water injection system flow rate, and suppression valve water temperature is The system is characterized in that it is configured to output a start signal for the automatic pressure reduction system based on the pressure flow characteristics of the relief safety valve and the water injection system by comparing the signal with a predetermined threshold value. .
本発明の自動減圧装置においては、逃し安全弁。In the automatic pressure reducing device of the present invention, a relief safety valve is used.
注水系の圧力流量特性を装置内に内蔵しており、予め、
４ｊプレツシヨンゾール水温の上背、原子炉水位の回復
を予測して減圧の為に必要どなる逃し安全弁の個数おに
び開放のタイミングを決定しているので、事故時の原子
炉水位を速やかに回復させることかできる。The pressure flow characteristics of the water injection system are built into the device, and the
4j Pressure Solder The number of safety relief valves required for depressurization and the timing of their opening are determined by predicting the recovery of the reactor water temperature and the reactor water temperature, so the reactor water level can be adjusted quickly in the event of an accident. It is possible to recover it.
（実施例） 本発明の実施例を図面を参照して説明する。(Example) Embodiments of the present invention will be described with reference to the drawings.
第１図は本発明の一実施例の論理回路図である。FIG. 1 is a logic circuit diagram of one embodiment of the present invention.
図に示すように、本発明の自動減圧装置２０は原子炉水
位信号２１、原子炉圧力信号２２、高圧低圧の注水系の
流量信号２３．２４、サルツションプール水温信号２５
をパノノとして、これらの各信号に基づき、第２図に示
覆゛ような逃し安全弁、注水系の圧力流量特性に基づき
、開くべき逃し安全弁個数及びタイミングを決定し、自
動減圧系起動信号２６を発生するように構成しており、
。As shown in the figure, the automatic decompression device 20 of the present invention includes a reactor water level signal 21, a reactor pressure signal 22, a high pressure and low pressure water injection system flow rate signal 23, 24, and a salt water temperature signal 25.
Based on these signals, the number and timing of safety relief valves to be opened are determined based on the pressure flow characteristics of the safety relief valves and water injection system as shown in FIG. 2, and the automatic pressure reduction system start signal 26 is activated. It is configured so that it occurs,
.
自動減圧装置２０は逃し安全弁を最大弁数の１８個、１
３〕個、ε〕個、Ｏ個、開けた場合について、各々第２
図の逃し支仝弁流量特性から原子炉域Ｈ−カーブをＫＩ
Ｅｉし・、この場合の注水系の流量を第２図の注水系
の特性から求める。また、この各々の場合について、原
子炉水位の最小値を注水系の流量の積分値から求めると
同時に、ザブレッジニー１ン７’−ル水温の最高値を逃
し安全弁流量の積分値から求める。前記４ケースのうち
で、原子炉水位最小値が燃料頂部＋３０ｃｍ以下どなる
もの又は−ｊｊプレッションプール水温の最高値か７１
°Ｃ以上となるものは排除して最小の逃し安全弁数のも
のを選択し運転員に表示する。逃し安全弁の開放タイミ
ングは、′１８個間のグーースては現状の遅れ時間と同
じ１２０秒とし、１５個では１００秒、９個Ｃは６０秒
とり−る。開放個数か少ない場合は、その影響か小ざい
のて、運転員の判断のための遅れ時間は少なくてよいか
らである。The automatic pressure reducing device 20 has a maximum number of relief safety valves of 18 and 1.
3] pieces, ε] pieces, O pieces, the second case respectively when opened.
KI the reactor area H-curve from the relief support valve flow characteristics shown in the figure.
Ei.The flow rate of the water injection system in this case is determined from the characteristics of the water injection system shown in FIG. In each of these cases, the minimum value of the reactor water level is determined from the integral value of the flow rate of the water injection system, and at the same time, the maximum value of the water temperature of the bridge knee 1 and 7' is determined from the integral value of the safety valve flow rate. Among the four cases mentioned above, the minimum value of the reactor water level is below the top of the fuel +30cm or the maximum value of the -jj pressure pool water temperature71
Exclude those that exceed °C and select the one with the minimum number of relief safety valves and display it to the operator. The opening timing of the relief safety valve is 120 seconds, which is the same as the current delay time, for the 18 gooses, 100 seconds for the 15 gooses, and 60 seconds for the 9 gooses. This is because if the number of open valves is small, the delay time for the operator's decision may be small, although the influence may be small.
し発明の効果］
以上説明しように、本発明によれば、サブレッショ１ン
１−ルの水温の急上胃、圧力容器の急激な減圧を避ける
ことができるので、原子炉水位を速やかに同復さけるこ
とかでさる。したかって、Ｉ京子炉の安全維持上−極め
てイ°ｊ益である。[Effects of the Invention] As explained above, according to the present invention, it is possible to avoid a sudden rise in the water temperature of the subression tank and a sudden depressurization of the pressure vessel, so that the reactor water level can be quickly restored. A monkey that screams. Therefore, this is extremely beneficial in terms of maintaining the safety of the Ikyo reactor.
第′１図は本発明の一実施例の論理回路図、第２図は本
発明に係る逃し・安全弁及び注水系の圧カー流量特性を
示すグラフ、第３図は減圧装置作動時の圧力・サプレッ
ションブール水温変化を示すグラフ、第１図は本発明が
適用される沸騰水型原子炉の模式的概略図、第５図は従
来の自動減圧装置の論理回路図でおる。
１・・・格納容器
２・・・Ｌ王力容器
３・・・炉心
４・・・冷却（・４
５・・・主蒸気管
６・・・タービン
７・・・］ンア゛ン４Ｊ
ε３・・・ドライウ１−ル
９・・・サブレッジコンプール
１０・・・逃し安全弁
１１・・・配管
１２・・・高圧注水系
１３・・・低圧汗水系
（８７３３）代理人 弁理に 猪 股 祥 晃（はか
１名）
−〇 −
−＠山女伏や
整←Fig. 1 is a logic circuit diagram of an embodiment of the present invention, Fig. 2 is a graph showing the pressure curve flow characteristics of the relief/safety valve and water injection system according to the present invention, and Fig. 3 is a graph showing the pressure/flow characteristics of the relief/safety valve and water injection system according to the present invention. FIG. 1 is a schematic diagram of a boiling water reactor to which the present invention is applied, and FIG. 5 is a logic circuit diagram of a conventional automatic decompression device. 1...Containment vessel 2...L power vessel 3...Core 4...Cooling (・4 5...Main steam pipe 6...Turbine 7...] Annex 4J ε3・...Dry wall 1-rule 9...Subledge compound pool 10...Relief safety valve 11...Piping 12...High pressure water injection system 13...Low pressure sweat water system (8733) Attorney Yoshiaki Inomata (Haka
1 person) −〇 − −＠Yama Onnabushi Yasei ←
Claims (1)
、前記原子炉圧力容器内の水位を確保するために設置さ
れた注水系と、前記原子炉圧力容器に接続された主蒸気
管に設けた逃し安全弁と、前記原子炉格納容器内に設け
たサプレッションプールとからなる原子炉において、原
子炉水位、原子炉圧力、注水系流量、サプレッションプ
ール水温の各信号に基づいて予め決められたしきい値と
比較して前記逃し安全弁及び注水系の圧力流量特性に基
づき自動減圧系の起動信号を出力するように構成したこ
とを特徴とする原子炉の自動減圧装置。(1) A reactor pressure vessel installed in the reactor containment vessel, a water injection system installed to ensure the water level in the reactor pressure vessel, and a main steam pipe connected to the reactor pressure vessel. In a nuclear reactor consisting of a relief safety valve provided in the reactor containment vessel and a suppression pool provided in the reactor containment vessel, predetermined signals are determined based on the reactor water level, reactor pressure, water injection system flow rate, and suppression pool water temperature. An automatic depressurization system for a nuclear reactor, characterized in that it is configured to output a start signal for an automatic depressurization system based on the pressure flow characteristics of the relief safety valve and the water injection system in comparison with a threshold value.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63113534A JPH01285897A (en) | 1988-05-12 | 1988-05-12 | Automatic pressure reducing device of nuclear reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP63113534A JPH01285897A (en) | 1988-05-12 | 1988-05-12 | Automatic pressure reducing device of nuclear reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH01285897A true JPH01285897A (en) | 1989-11-16 |
Family
ID=14614760
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP63113534A Pending JPH01285897A (en) | 1988-05-12 | 1988-05-12 | Automatic pressure reducing device of nuclear reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH01285897A (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN105469840A (en) * | 2015-11-25 | 2016-04-06 | 中广核工程有限公司 | Cooling method, device and system for loss-of-coolant accident of first loop of nuclear power station |
| CN113571211A (en) * | 2021-07-06 | 2021-10-29 | 中国核电工程有限公司 | Reactor overpressure protection system and method, nuclear power system and primary loop system thereof |
-
1988
- 1988-05-12 JP JP63113534A patent/JPH01285897A/en active Pending
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN105469840A (en) * | 2015-11-25 | 2016-04-06 | 中广核工程有限公司 | Cooling method, device and system for loss-of-coolant accident of first loop of nuclear power station |
| CN113571211A (en) * | 2021-07-06 | 2021-10-29 | 中国核电工程有限公司 | Reactor overpressure protection system and method, nuclear power system and primary loop system thereof |
| CN113571211B (en) * | 2021-07-06 | 2023-12-19 | 中国核电工程有限公司 | Nuclear power system and method and primary loop system thereof as well as reactor overpressure protection system and method |
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| JPH02222877A (en) | Warming arrangement of pump of reactor remaining heat removal system | |
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