JPH0752232B2 - Boiling water reactor - Google Patents
Boiling water reactorInfo
- Publication number
- JPH0752232B2 JPH0752232B2 JP61196645A JP19664586A JPH0752232B2 JP H0752232 B2 JPH0752232 B2 JP H0752232B2 JP 61196645 A JP61196645 A JP 61196645A JP 19664586 A JP19664586 A JP 19664586A JP H0752232 B2 JPH0752232 B2 JP H0752232B2
- Authority
- JP
- Japan
- Prior art keywords
- core
- reactor
- boiling water
- shroud
- water level
- 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.)
- Expired - Lifetime
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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
-
- 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
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Organic Low-Molecular-Weight Compounds And Preparation Thereof (AREA)
Description
【発明の詳細な説明】 〔産業上の利用分野〕 本発明は、沸騰水型原子炉に係り、特に、負荷変動を容
易に追従するのに好適な自然循環型の沸騰水型原子炉に
関する。Description: TECHNICAL FIELD The present invention relates to a boiling water nuclear reactor, and more particularly to a natural circulation boiling water nuclear reactor suitable for easily following load fluctuations.
沸騰水型原子炉（以下、BWRともいう。）は、既存プラ
ントが最も代表的に知られている。第５図に、既存BWR
プラントの一例を示す。The boiling water reactor (hereinafter, also referred to as BWR) is most commonly known as an existing plant. Fig. 5 shows the existing BWR
An example of a plant is shown.
BWRプラントは、圧力容器１の中を炉心シユラウド３で
大きく２分割し、炉心シユラウド３内側に核燃料を装荷
した炉心２を設けてある。炉心２で発生した熱は、圧力
容器１内の冷却水に伝わり、炉心シユラウド３内で蒸気
を発生する。In the BWR plant, the pressure vessel 1 is roughly divided into two parts with a core shell 3 and a core 2 loaded with a nuclear fuel is provided inside the core shell 3. The heat generated in the core 2 is transmitted to the cooling water in the pressure vessel 1 and steam is generated in the core shell 3.
この蒸気は、冷却水とともに炉心シユラウド２上部に設
けられたスタンドパイプ（ならびに気水分離器）４で蒸
気と冷却水に分離される。蒸気は、最終的にタービン系
に送られ、発電に供される。圧力容器１内の冷却水から
持ち出された分は、給水スパージヤ５から補給される。This steam is separated into steam and cooling water together with the cooling water by a stand pipe (and steam separator) 4 provided above the core shroud 2. The steam is finally sent to a turbine system for power generation. The water taken out from the cooling water in the pressure vessel 1 is replenished from the water supply sparger 5.
このような既存BWRは、一定出力（定格出力）を対象と
して設計されており、負荷変動に追従する出力の制御は
困難であつた。一部の負荷変動に対しては、再循環系16
の流量を制御し追従可能であつたが、再循環系16のポン
プやモータのほかに種々の補器設備が必要となり、例え
ば、都市近郊を対象として中小型炉を設計しようとして
も、設置スペースの点で非現実的なシステムとなつてい
た。Such an existing BWR was designed for constant output (rated output), and it was difficult to control the output to follow load fluctuations. For some load fluctuations, recirculation system 16
Although it is possible to control and follow the flow rate of the recirculation system 16, various auxiliary equipments are required in addition to the pump and motor of the recirculation system 16. It was an unrealistic system.
また、既存BWRでは、制御棒および再循環系16以外での
出力制御機能はなく、スタンドパイプ４部は、炉心シユ
ラウド３内で発生する蒸気と水の二相流を蒸気と冷却水
とに分離する機能しか持つていない。そのために、炉心
シユラウド３外側の水位は、気水分離器の最も効率の良
い点近傍に維持することが不可欠であつた。In addition, in the existing BWR, there is no output control function other than the control rod and the recirculation system 16, and the stand pipe 4 part separates the two-phase flow of steam and water generated in the core shell 3 into steam and cooling water. It has only the function to do. Therefore, it was essential to maintain the water level outside the core shell 3 near the most efficient point of the steam separator.
負荷追従性を勘案した自然循環型の沸騰水型原子炉とし
ては、例えば、特開昭60−169793号がある。第６図にそ
の考えかたに沿つたシステム構成の一例を示す。核燃料
を装荷した圧力容器１と、圧力容器１内で発生した蒸気
を凝縮する熱交換器14と、熱交換器14内で凝縮した冷却
水を前記圧力容器１に戻す再循環系ポンプ９等を設けた
沸騰水型原子炉である。As a natural circulation type boiling water reactor in consideration of load followability, for example, there is JP-A-60-169793. FIG. 6 shows an example of a system configuration along the way of thinking. A pressure vessel 1 loaded with nuclear fuel, a heat exchanger 14 for condensing vapor generated in the pressure vessel 1, a recirculation system pump 9 for returning the cooling water condensed in the heat exchanger 14 to the pressure vessel 1, and the like. It is a boiling water reactor provided.
従来技術では、熱交換器１次系内の液相部と蒸気相部の
割合を調整し、負荷に応じた熱を２次系に供給可能であ
り、炉心入口サブクールの減少に伴い、炉心部での出力
を増加させることができる。In the conventional technology, the ratio of the liquid phase portion and the vapor phase portion in the primary system of the heat exchanger can be adjusted to supply heat according to the load to the secondary system, and as the core inlet subcool decreases, the core portion can be reduced. The output at can be increased.
しかし、この従来技術では、大型の熱交換器が必要であ
り、熱交換器１次系の水位を制御することから、制御
性，運転操作性、特に応答性が良くなかつた。However, this conventional technique requires a large-sized heat exchanger and controls the water level of the primary system of the heat exchanger, so that controllability, driving operability, and particularly responsiveness are not good.
最近は、更に小型化ししかも応答性が良いものが要求さ
れ、熱交換器を用いずに負荷変動に迅速に追従可能な発
電用原子炉の必要性が高まつている。Recently, there has been a demand for further downsizing and a good responsiveness, and there is a growing need for a nuclear reactor for power generation that can quickly follow load changes without using a heat exchanger.
上記従来技術は、大型の再循環系または熱交換器を必要
とし、また、熱交換器１次系の水位を調整し負荷変動に
追従するために、次のような問題点があつた。The above prior art requires a large-scale recirculation system or a heat exchanger, and has the following problems in order to adjust the water level of the primary system of the heat exchanger and follow the load fluctuation.
（１）冗長性を勘案すると、100％容量の熱交換器が２
台以上必要であるなど、大型化する傾向にあり、現実の
要求に合わない。(1) Considering redundancy, two 100% capacity heat exchangers
There is a tendency to increase the size, such as the need for more than one unit, which does not meet the actual demand.
（２）熱交換器付きのプラントは、転倒モーメントが大
きく、丈夫な土台が要求され、熱交換器設置地盤やスペ
ース等に多大な考慮を払う必要がある。(2) A plant with a heat exchanger has a large overturning moment and a strong foundation is required, and it is necessary to give great consideration to the ground and space where the heat exchanger is installed.
（３）熱交換器１次系内で液面を形成するために、エロ
ージヨンに対する対策が必須であり、熱交換器コストが
増大する。(3) In order to form the liquid level in the primary system of the heat exchanger, it is necessary to take measures against erosion, which increases the cost of the heat exchanger.
本発明の目的は、熱交換器等の大型付帯設備を用いるこ
となく負荷変動に対して迅速に応答可能な沸騰水型原子
炉を提供することである。An object of the present invention is to provide a boiling water reactor capable of responding promptly to load changes without using large-scale auxiliary equipment such as a heat exchanger.
本発明は、上記目的を達成するために、縦方向に複数個
の孔を有するスタンドパイプを炉心シユラウド上部に少
なくとも１本設けるとともに、圧力容器内の炉心シユラ
ウド外側水位を変える手段を設けた沸騰水型原子炉を提
案するものである。In order to achieve the above object, the present invention provides at least one stand pipe having a plurality of holes in the longitudinal direction at the upper part of the core shell, and a means for changing the water level outside the core shell in the pressure vessel. Type reactor.
前記スタンドパイプ縦方向に設けた複数の孔の合計面積
は、次式に従つて決める。The total area of the plurality of holes provided in the vertical direction of the stand pipe is determined according to the following equation.
ここに、 A:スタンドパイプの縦孔合計面積 ρ：冷却材密度 ΔP:シユラウド内外差圧 W:設計炉心流量 g:重力加速度 ξ：設計データ また、前記炉心シユラウド外側水位を変える手段は、炉
心シユラウド外側水位計と炉心シユラウド外側圧力計と
それらの検出信号により圧力容器への給水制御弁の開度
を制御する手段とからなる。 Where: A: Total area of vertical holes in the standpipe ρ: Coolant density ΔP: Differential pressure inside and outside the shroud W: Design core flow rate g: Gravitational acceleration ξ: Design data It consists of an outer water level gauge, a core shell outer pressure gauge, and means for controlling the opening of the feed water control valve to the pressure vessel by the detection signals of these.
上記構成では、炉心からスタンドパイプ孔を通りダウン
カマに至る自然循環ループが形成されダウンカマ水位を
変えれば、自然循環流量が変えられ、その自然循環流量
に対応する炉心出力が得られることになる。In the above configuration, if a natural circulation loop from the core to the downcomer is formed through the stand pipe hole and the downcomer water level is changed, the natural circulation flow rate is changed, and the core output corresponding to the natural circulation flow rate is obtained.
原子炉圧力容器内にそのための可動部を設けることな
く、自然循環流量および炉心出力の制御を、ダウンカマ
の水位のみで行うことに特徴がある。The feature is that the natural circulation flow rate and the core output are controlled only by the water level of the downcomer without providing a movable part for that purpose in the reactor pressure vessel.
第２図に自然循環時の炉心出力と自然循環流量との関係
を示す。自然循環流量は、出力が０から増加するに従つ
て増加するが、極大点を越えると逆に減少する傾向を持
つ。一方、炉心出力を自然循環状態におけるボイド反応
度のみで制御すると、自然循環時の出力は、自然循環流
量の極大点近傍で安定する。これら極大点は、炉心シユ
ラウド内外の差圧が極大になる点に対応している。この
現象は、シユラウド外側圧力がほぼ水頭のみであるのに
対し、シユラウド内ではボイド発生に伴う水頭の減少
と、二相抵抗の増加がバランスすることに対応する。Figure 2 shows the relationship between core power and natural circulation flow rate during natural circulation. The natural circulation flow rate increases as the output increases from 0, but tends to decrease on the contrary when the maximum point is exceeded. On the other hand, when the core output is controlled only by the void reactivity in the natural circulation state, the output during natural circulation becomes stable near the maximum point of the natural circulation flow rate. These maximum points correspond to the points where the differential pressure inside and outside the core shell becomes maximum. This phenomenon corresponds to the fact that the outer pressure of the shroud is almost only the head of the head, while the head of the head decreases with the occurrence of voids and the increase of the two-phase resistance is balanced in the head.
すなわち、極大点より出力が増加すると、ボイドが増え
るとともに、二相流抵抗が増して、自然循流量が減少す
る。このボイド増加で出力が抑制される。一方、極大点
から出力が減少すると、自然循環流量が減少し、ボイド
が減少し、出力が回復する。よつて、自然循環状態の出
力は、ボイド反応度のフイードバツクで、自然循環特性
の極大点近傍に安定することが判る。一方、自然循環特
性の安定点は炉心シユラウド内の水頭と二相抵抗の和の
極小値に対応するために、システムの二相抵抗を変えれ
ば、自然循環特性の安定点、すなわち運転点を制御可能
となる。That is, when the output increases from the maximum point, the number of voids increases, the two-phase flow resistance increases, and the natural circulation flow rate decreases. The output is suppressed by this increase in voids. On the other hand, when the output decreases from the maximum point, the natural circulation flow rate decreases, the voids decrease, and the output recovers. Therefore, it can be seen that the output in the natural circulation state is a feed back of the void reactivity and is stable near the maximum point of the natural circulation characteristic. On the other hand, since the stable point of the natural circulation characteristic corresponds to the minimum value of the sum of the water head and the two-phase resistance in the core shell, changing the two-phase resistance of the system controls the stable point of the natural circulation characteristic, that is, the operating point. It will be possible.
本発明は、ダウンカマ水位を変化させ、多孔スタンドパ
イプの冷却水の流路面積を変えられるため、シユラウド
内の二相抵抗を変更可能となる。従つて、自然循環流量
特性の極大点を、ダウンカマ水位の変更のみで変えら
れ、負荷に追従する沸騰水型原子炉が得られる。According to the present invention, since the downcomer water level can be changed and the flow passage area of the cooling water of the perforated stand pipe can be changed, it is possible to change the two-phase resistance in the shell. Therefore, the maximum point of the natural circulation flow rate characteristic can be changed only by changing the downcomer water level, and a boiling water reactor that follows the load can be obtained.
次に、本発明による沸騰水型原子炉の一実施例を第１図
により説明する。図において、原子炉圧力容器１内に装
荷された炉心２は、炉心シユラウド３に囲まれている。
炉心シユラウド３の上部には、縦方向に複数の孔を有す
るスタンドパイプ４がある。炉心２で発生した蒸気は、
スタンドパイプ４と加減弁17とを介し、タービン10に導
かれる。タービン10で凝縮した蒸気は、給水ポンプ９か
ら給水スパージヤ５を介し、圧力容器１内に戻る。ター
ビン10は、発電機11を駆動し、負荷12に電気を供給して
いる。Next, an embodiment of the boiling water reactor according to the present invention will be described with reference to FIG. In the drawing, a core 2 loaded in a reactor pressure vessel 1 is surrounded by a core shell 3.
Above the core shell 3 is a stand pipe 4 having a plurality of holes in the vertical direction. The steam generated in the core 2
It is guided to the turbine 10 via the stand pipe 4 and the regulator valve 17. The steam condensed in the turbine 10 returns from the water supply pump 9 to the inside of the pressure vessel 1 via the water supply sparger 5. The turbine 10 drives a generator 11 and supplies electricity to a load 12.
圧力容器１では、炉心シユラウド３内外の差圧に伴う自
然循環によつて除熱するとともに、その出力を制御す
る。炉心シユラウド３内の自然循環流は、スタンドパイ
プ４に設けた複数の孔から炉心シユラウド３の外側へオ
ーバフローする。これらスタンドパイプ４のオーバフロ
ー面積は、炉心シユラウド３の外側の水位に依存するた
めに、水位計および圧力計を設け、制御用のパラメータ
を得ている。In the pressure vessel 1, heat is removed by natural circulation associated with the pressure difference inside and outside the core shroud 3, and the output thereof is controlled. The natural circulation flow in the core shroud 3 overflows from the plurality of holes provided in the stand pipe 4 to the outside of the core shroud 3. The overflow area of these stand pipes 4 depends on the water level outside the core shroud 3, so a water level gauge and a pressure gauge are provided to obtain control parameters.
主蒸気および給水系により炉心シユラウド３外側の水位
を制御する例を以下に述べる。負荷12の増加に伴い要求
される蒸気量の増加は、蒸気加減弁17の開度を大きくし
て対処するが、同時に圧力容器１内の冷却水量が減少す
る方向となる。そこで、給水制御弁18の開度を大きく
し、主蒸気流量に対し給水流量が大きい条件を作る。こ
れによつて、炉心シユラウド３外側の水位が増加し、ス
タンドパイプオーバフロー面積が増加するとともに、自
然循環流量の増大に伴う出力の上昇が得られる。必要な
出力条件が達成されたら、給水系流量が主蒸気流量と同
じとなるように給水制御弁18の開度を減らすとともに、
水位計６で監視している炉心シユラウド３外側水位を一
定に維持する。An example of controlling the water level outside the core shell 3 by the main steam and the water supply system will be described below. The increase in the amount of steam required as the load 12 increases is dealt with by increasing the opening degree of the steam control valve 17, but at the same time, the amount of cooling water in the pressure vessel 1 tends to decrease. Therefore, the opening of the water supply control valve 18 is increased to create a condition in which the water supply flow rate is larger than the main steam flow rate. As a result, the water level outside the core shell 3 increases, the standpipe overflow area increases, and the output rises with the increase of the natural circulation flow rate. When the required output condition is achieved, reduce the opening of the feedwater control valve 18 so that the feedwater flow rate is the same as the main steam flow rate, and
The water level outside the core shell 3 monitored by the water level gauge 6 is kept constant.
逆に、出力を低下させる場合には、給水制御弁18の開度
を減少させ、必要な出力条件を達成する。On the contrary, when the output is reduced, the opening degree of the water supply control valve 18 is reduced to achieve the required output condition.
自然循環流量を制御して炉出力を変える本発明の原理を
第３図に示す。The principle of the present invention for controlling the natural circulation flow rate to change the furnace output is shown in FIG.
自然循環流量は、圧力容器１内に設けられた炉心シユラ
ウド３外の圧力（ΔPDC）と内の圧力（ΔPSH）の差から
求められる。ΔPDCは、炉心２の出力には依存せず、炉
心シユラウド３外側の水位に依存する。一方、ΔP
SHは、水頭（ΔPh）と二相摩擦損失（ΔPf）の和とな
る。ΔPhは、炉心２の出力増加に伴い指数関数的に減少
するのに対し、ΔPfは、二乗で増加する傾向を持つ。Δ
Pfは、クオリテイの高い部分で決定されることが知られ
ており、スタンドパイプ４での損失が最も支配的であ
る。そこで、自然循環流量がオーバフローする面積を変
えると、ΔPfはL1またはL2のごとく変化する。同時に、
ボイド反応度で制御されている炉心２の出力は、Q1また
はQ2で安定する。本実施例は、縦方向に複数の孔を有す
るスタンドパイプ４の自然循環流路面積を炉心シユラウ
ド３外側水位で制御するものである。The natural circulation flow rate is obtained from the difference between the pressure outside the core shell 3 (ΔP DC ) provided inside the pressure vessel 1 and the internal pressure (ΔP SH ). ΔP DC does not depend on the output of the core 2, but depends on the water level outside the core shell 3. On the other hand, ΔP
SH is the sum of hydrohead (ΔP h ) and two-phase friction loss (ΔP f ). ΔP h decreases exponentially with an increase in the power output of the core 2, while ΔP f tends to increase as the square. Δ
It is known that P f is determined in the high quality part, and the loss in the stand pipe 4 is the most dominant. Therefore, if the area where the natural circulation flow rate overflows is changed, ΔP f changes like L 1 or L 2 . at the same time,
The power of core 2 controlled by void reactivity stabilizes at Q 1 or Q 2 . In this embodiment, the natural circulation flow passage area of the stand pipe 4 having a plurality of holes in the vertical direction is controlled by the water level outside the core shell 3.
本発明のスタンドパイプの孔は、プラントの負荷変動範
囲などから任意に設計できるが、制御性が確保される条
件として、次の式を満足しなければならない。The hole of the stand pipe of the present invention can be arbitrarily designed from the load variation range of the plant, etc., but the following formula must be satisfied as a condition for ensuring controllability.
ここで、 A:スタンドパイプの縦孔合計面積〔m2〕 ρ：冷却材密度〔kg/m3〕 ΔP:シユラウド内外差圧〔kg/m2〕 W:設計炉心流量〔m3/s〕 g:重力加速度:9.8〔m/s2〕 ξ：設計データ:0.5〜1.0〔−〕 これは、自然循環流量が、スタンドパイプ上端よりオー
バフローすることを阻止する条件を示す。 Where: A: Total area of vertical holes in the stand pipe [m 2 ] ρ: Coolant density [kg / m 3 ] ΔP: Internal and external pressure differential of the shroud [kg / m 2 ] W: Design core flow rate [m 3 / s] g: Gravitational acceleration: 9.8 [m / s 2 ] ξ: Design data: 0.5 to 1.0 [−] This indicates the condition that the natural circulation flow rate is prevented from overflowing from the upper end of the stand pipe.
自然循環型原子炉は、一般に高出力部に不安定領域を持
つことが判つている。Natural circulation reactors are generally known to have an unstable region in the high power section.
第４図に既存炉の実験値を示す。自然循環型原子炉で
は、任意の圧力に対し、高出力密度状態で不安定となる
ため、図中の安定領域で運転する必要がある。これか
ら、要求電力に対する炉圧の概略仕様が得られ、炉圧の
維持を考慮して制御系のパラメータを決める。Figure 4 shows the experimental values of the existing furnace. Since the natural circulation reactor becomes unstable in a high power density state with respect to an arbitrary pressure, it is necessary to operate in a stable region in the figure. From this, a rough specification of the furnace pressure for the required power can be obtained, and the parameters of the control system are determined in consideration of maintaining the furnace pressure.
本発明によれば、大型の熱交換器などの付帯設備を設け
ることなく、負荷変動に即応できる自然循環型原子炉が
得られる。According to the present invention, it is possible to obtain a natural circulation reactor that can quickly respond to load fluctuations without providing auxiliary equipment such as a large heat exchanger.
付帯設備が少ないことから、発電炉を対象とする場合、
1/2以下のスペースで従来と同出力を得ることができ、
例えば、都市近郊の建設ニーズにも適合する。Since there are few incidental facilities, when targeting a power reactor,
The same output as before can be obtained in a space of 1/2 or less,
For example, it meets the construction needs in the suburbs.
また、負荷追従の面では、圧力容器内に可動部を設ける
ことなく、シユラウド外側の水位のみで対応可能である
ため、システムの信頼性が高く、操作も容易である。従
つて、運転員数を減らしまたは集中操作するのに好適
で、合理的運営ができる。Further, in terms of load follow-up, since the water level outside the shroud can be used without providing a movable part in the pressure vessel, the system has high reliability and is easy to operate. Therefore, it is suitable for reducing the number of operators or performing centralized operation, and rational operation can be performed.
第１図は本発明によるBWRの一実施例の原理的構成を示
す図、第２図は本発明によるBWRの出力制御の原理を示
す図、第３図は本発明の自然循環型流量制御により炉出
力を制御する方式を示す図、第４図は原子炉の安定領域
を示す図、第５図は再循環系を用いた従来の原子炉の一
例を示す図、第６図は熱交換器を外付けした従来の原子
炉の一例を示す図である。 １…圧力容器、２…炉心、３…炉心シユラウド、４…ス
タンドパイプ、５…給水スパージヤ、６…水位計、７…
圧力計、８…制御弁、９…給水ポンプ、10…タービン、
11…発電機、12…負荷、13…１次系ライン、14…熱交換
器、15…２次系ポンプ、16…再循環系、17…蒸気加減
弁、18…給水制御弁。FIG. 1 is a diagram showing a principle configuration of an embodiment of a BWR according to the present invention, FIG. 2 is a diagram showing a principle of output control of a BWR according to the present invention, and FIG. 3 is a diagram showing a natural circulation type flow control according to the present invention. FIG. 4 is a diagram showing a system for controlling reactor power, FIG. 4 is a diagram showing a stable region of the reactor, FIG. 5 is a diagram showing an example of a conventional reactor using a recirculation system, and FIG. 6 is a heat exchanger. It is a figure which shows an example of the conventional nuclear reactor which attached externally. 1 ... Pressure vessel, 2 ... Core, 3 ... Core shroud, 4 ... Stand pipe, 5 ... Water supply sparger, 6 ... Water level gauge, 7 ...
Pressure gauge, 8 ... Control valve, 9 ... Water supply pump, 10 ... Turbine,
11 ... Generator, 12 ... Load, 13 ... Primary system line, 14 ... Heat exchanger, 15 ... Secondary system pump, 16 ... Recirculation system, 17 ... Steam control valve, 18 ... Water supply control valve.
フロントページの続き (51)Int.Cl.6 識別記号 庁内整理番号 ＦＩ 技術表示箇所 Ｇ２１Ｃ 15/16 Ｇ２１Ｄ 3/08 Ｅ 9117−2Ｇ 3/14 9117−2Ｇ (72)発明者 松本 雅喜 茨城県日立市幸町３丁目１番１号 株式会 社日立製作所日立工場内 (72)発明者 橋本 光司 茨城県日立市幸町３丁目２番１号 日立エ ンジニアリング株式会社内 (56)参考文献 特開 昭60−169793（ＪＰ，Ａ)Continuation of front page (51) Int.Cl. 6 Identification number Office reference number FI Technical display location G21C 15/16 G21D 3/08 E 9117-2G 3/14 9117-2G (72) Inventor Masayoshi Matsumoto Hitachi, Ibaraki Prefecture 3-1-1, Saicho-cho, Hitachi, Ltd. Hitachi factory, Hitachi Ltd. (72) Inventor, Koji Hashimoto 3-2-1, Saiwai-cho, Hitachi, Ibaraki Hitachi Engineering Co., Ltd. (56) References 60-169793 (JP, A)
Claims (3)
ウドの内外差圧を駆動源とした自然循環流体で炉心を冷
却する沸騰水型原子炉において、縦方向に複数個の孔を
有するスタンドパイプを上記炉心シユラウド上部に少な
くとも１本設けるとともに、上記圧力容器内の上記炉心
シユラウド外側水位を変える手段を設けたことを特徴と
する沸騰水型原子炉。1. A boiling water reactor in which a natural circulating fluid is used as a drive source to cool the core in a reactor pressure vessel by using a differential pressure inside and outside a core shell that surrounds the core, and a stand having a plurality of holes in a vertical direction. A boiling water reactor characterized in that at least one pipe is provided above the reactor core shroud, and means for changing the water level outside the reactor core shroud in the pressure vessel is provided.
ンドパイプ縦方向に設けた複数の孔の合計面積が次式に
従うことを特徴とする沸騰水型原子炉。 ここに、 A:スタンドパイプの縦孔合計面積〔m2〕 ρ：冷却材密度〔kg/m3〕 ΔP:シユラウド内外差圧〔kg/m2〕 W:設計炉心流量〔m3/s〕 g:重力加速度:9.8〔m/s2〕 ξ：設計データ:0.5〜1.0〔−〕2. A boiling water nuclear reactor according to claim 1, wherein the total area of a plurality of holes provided in the vertical direction of the stand pipe complies with the following equation. Where: A: Total area of vertical holes in the stand pipe [m 2 ] ρ: Coolant density [kg / m 3 ] ΔP: Internal and external pressure difference of the shroud [kg / m 2 ] W: Design core flow rate [m 3 / s] g: Gravity acceleration: 9.8 (m / s 2 ) ξ: Design data: 0.5 to 1.0 (-)
て、前記炉心シユラウド外側水位を変える手段が、炉心
シユラウド外側水位計と炉心シユラウド外側圧力計とそ
れらの検出信号により前記原子炉圧力容器への給水制御
弁の開度を制御する手段とからなることを特徴とする沸
騰水型原子炉。3. The reactor pressure vessel according to claim 1 or 2, wherein the means for changing the outer water level of the core shell is based on the outer water level gauge of the core shroud, the outer pressure gauge of the core shroud and their detection signals. And a means for controlling the opening of a water supply control valve to the boiling water reactor.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61196645A JPH0752232B2 (en) | 1986-08-22 | 1986-08-22 | Boiling water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61196645A JPH0752232B2 (en) | 1986-08-22 | 1986-08-22 | Boiling water reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS6352097A JPS6352097A (en) | 1988-03-05 |
| JPH0752232B2 true JPH0752232B2 (en) | 1995-06-05 |
Family
ID=16361218
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61196645A Expired - Lifetime JPH0752232B2 (en) | 1986-08-22 | 1986-08-22 | Boiling water reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0752232B2 (en) |
Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP4197696B2 (en) | 2005-08-11 | 2008-12-17 | 株式会社東芝 | Natural circulation boiling water reactor |
| JP4556883B2 (en) * | 2006-02-27 | 2010-10-06 | 株式会社日立製作所 | Reactor power controller |
-
1986
- 1986-08-22 JP JP61196645A patent/JPH0752232B2/en not_active Expired - Lifetime
Also Published As
| Publication number | Publication date |
|---|---|
| JPS6352097A (en) | 1988-03-05 |
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