JPH047960B2 - - Google Patents
Info
- Publication number
- JPH047960B2 JPH047960B2 JP60121514A JP12151485A JPH047960B2 JP H047960 B2 JPH047960 B2 JP H047960B2 JP 60121514 A JP60121514 A JP 60121514A JP 12151485 A JP12151485 A JP 12151485A JP H047960 B2 JPH047960 B2 JP H047960B2
- Authority
- JP
- Japan
- Prior art keywords
- primary coolant
- reference temperature
- temperature
- control
- reactor
- 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
Links
- 239000002826 coolant Substances 0.000 claims description 98
- 230000008859 change Effects 0.000 claims description 34
- KGBXLFKZBHKPEV-UHFFFAOYSA-N boric acid Chemical compound OB(O)O KGBXLFKZBHKPEV-UHFFFAOYSA-N 0.000 claims description 29
- 239000004327 boric acid Substances 0.000 claims description 29
- 238000000034 method Methods 0.000 claims description 12
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 10
- 238000001816 cooling Methods 0.000 claims description 3
- 239000000463 material Substances 0.000 claims description 2
- 230000009257 reactivity Effects 0.000 description 16
- 230000007423 decrease Effects 0.000 description 9
- 229910052724 xenon Inorganic materials 0.000 description 8
- FHNFHKCVQCLJFQ-UHFFFAOYSA-N xenon atom Chemical compound [Xe] FHNFHKCVQCLJFQ-UHFFFAOYSA-N 0.000 description 8
- 238000009825 accumulation Methods 0.000 description 5
- 238000010586 diagram Methods 0.000 description 3
- 230000000694 effects Effects 0.000 description 2
- 230000001052 transient effect Effects 0.000 description 2
- 238000010521 absorption reaction Methods 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 230000004907 flux Effects 0.000 description 1
- 230000008569 process Effects 0.000 description 1
- 230000004044 response Effects 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21D—NUCLEAR POWER PLANT
- G21D3/00—Control of nuclear power plant
- G21D3/08—Regulation of any parameters in the plant
- G21D3/12—Regulation of any parameters in the plant by adjustment of the reactor in response only to changes in engine demand
-
- 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
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Description
【発明の詳細な説明】
〔産業上の利用分野〕
本発明は加圧水型原子炉に関し、特に一次冷却
材の温度を調節して同原子炉の負荷追従運転性能
改善を計る方法に関するものである。DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to a pressurized water nuclear reactor, and more particularly to a method for improving the load following operational performance of the reactor by adjusting the temperature of the primary coolant.
従来の加圧水型原子炉の原子炉制御系では、一
次冷却材基準温度の制御には、定常運転時におい
ても負荷追従運転時においても、タービン出力の
関数として定めた基準温度プログラムが採用され
ている。その一例が第６図に示されており、この
図では、一次冷却材基準温度はタービン出力の一
意的関数として設定されている。従つて、出力変
更を伴う負荷追従運転を行つた場合、一次冷却材
の温度はタービン出力により定まる基準温度に制
御される。ところが一次冷却材基準温度をこのよ
うに制御すると、負荷追従運転時に冷却材の反応
度温度係数のため反応度の変化が大きくなるの
で、制御棒の位置調整と原子炉冷却材中のほう酸
の濃度調整の負担が大きくなる。そねため、負荷
追従能力は制御棒位置調整能力及びほう酸の濃度
調整能力により制限される場合があつた。特に、
原子炉の炉心末期においては、ほう酸の濃度調整
能力が低下するため、負荷追従能力が制限されて
いた。
In the reactor control system of conventional pressurized water reactors, a reference temperature program determined as a function of turbine output is used to control the primary coolant reference temperature, both during steady operation and during load following operation. . An example is shown in FIG. 6, where the primary coolant reference temperature is set as a unique function of turbine power. Therefore, when performing a load following operation that involves a change in output, the temperature of the primary coolant is controlled to the reference temperature determined by the turbine output. However, if the primary coolant reference temperature is controlled in this way, the change in reactivity becomes large due to the reactivity temperature coefficient of the coolant during load following operation, so it is necessary to adjust the position of the control rods and the concentration of boric acid in the reactor coolant. The burden of adjustment becomes heavier. As a result, the load following ability was sometimes limited by the control rod position adjustment ability and the boric acid concentration adjustment ability. especially,
In the final stage of a nuclear reactor core, the ability to adjust the concentration of boric acid declines, limiting the ability to follow the load.
従つて、従来の技術には、特に炉心末期におい
て負荷追従能力が高くない問題点があつた。本発
明はかかる目的を速やかに解決する加圧水型原子
炉の冷却材温度制御方法を提供することを目的と
するものである。
Therefore, the conventional technology has the problem that the load following ability is not high, especially at the end of the core. It is an object of the present invention to provide a method for controlling the coolant temperature of a pressurized water reactor, which promptly solves the above object.
この目的を達成するため、第１の発明は、原子
炉の負荷追従運転時の制御のために、一次冷却材
の基準温度を計算するプログラム演算回路と、こ
の基準温度に一次冷却材温度を制御すべく、原子
炉出力分布を制御するための制御棒を制御する制
御棒制御回路と、一次冷却材中のほう酸濃度を制
御するほう酸濃度制御回路とを備え、制御棒の位
置調整及びほう酸の濃度調整によつて負荷追従制
御する加圧水型原子炉の冷却材温度制御系におい
て、タービン出力の関数とした定常運転時の一次
冷却材基準温度と、タービン出力の関数とした、
負荷追従運転時の一次冷却材基準温度の変化範囲
を制限する上限温度及び下限温度とを定め、負荷
追従運転時の一次冷却材基準温度をその上下限温
度範囲内で、該負荷追従運転時の一次冷却材基準
温度と定常運転時の一次冷却材基準温度との間の
偏差の関数として、定常運転時の一次冷却材基準
温度に接近するよう制御することを特徴とするも
のである。
In order to achieve this object, the first invention provides a program calculation circuit that calculates a reference temperature of a primary coolant for control during load following operation of a nuclear reactor, and controls the temperature of the primary coolant to this reference temperature. In order to achieve this, the control rod control circuit controls the control rods to control the reactor power distribution, and the boric acid concentration control circuit controls the boric acid concentration in the primary coolant. In a coolant temperature control system for a pressurized water reactor that performs load following control through adjustment, the primary coolant reference temperature during steady operation is a function of turbine output, and the primary coolant temperature is a function of turbine output.
An upper limit temperature and a lower limit temperature that limit the change range of the primary coolant reference temperature during load following operation are determined, and the primary coolant reference temperature during load following operation is set within the upper and lower limit temperature range during the load following operation. It is characterized in that the primary coolant reference temperature is controlled to approach the primary coolant reference temperature during steady operation as a function of the deviation between the primary coolant reference temperature and the primary coolant reference temperature during steady operation.
また、上記目的を達成するため、第２の発明に
よれば、上述した冷却材温度制御系において、タ
ービン出力の関数とした定常運転時の一次冷却材
基準温度と、タービン出力の関数とした、負荷追
従運転の一次冷却材基準温度の変化範囲を制限す
る上限温度及び下限温度とを定め、負荷追従運転
時の一次冷却材基準温度をその上下限温度範囲内
で、上記定常運転時の一次冷却材基準温度に遅れ
をもつて追従する関数により、上記定常運転時の
一次冷却材基準温度に近付くように制御すること
を特徴とするものである。 Further, in order to achieve the above object, according to the second invention, in the above-mentioned coolant temperature control system, the primary coolant reference temperature during steady operation is set as a function of turbine output, and the primary coolant reference temperature is set as a function of turbine output. An upper limit temperature and a lower limit temperature are determined to limit the change range of the primary coolant reference temperature during load following operation, and the primary coolant reference temperature during load following operation is set within the upper and lower limit temperature ranges, and the primary cooling temperature during the steady operation described above is set. This is characterized in that the primary coolant reference temperature is controlled to approach the primary coolant reference temperature during steady operation using a function that follows the material reference temperature with a delay.
上述した第１、第２の発明のように負荷追従運
転時の一次冷却材基準温度を可変にすると、所要
の反応度変化に対する制御棒位置調整及びほう酸
濃度調整の負担が軽減し、その分だけ負荷追従性
能が向上する。 By making the primary coolant reference temperature variable during load following operation as in the first and second inventions described above, the burden of adjusting the control rod position and boric acid concentration in response to the required change in reactivity is reduced, and the burden of adjusting the boric acid concentration is reduced accordingly. Improves load following performance.
実施例 １
次に、本発明による冷却材温度制御方法の好適
な実施例について添付図面を参照して詳細に説明
する。Embodiment 1 Next, a preferred embodiment of the coolant temperature control method according to the present invention will be described in detail with reference to the accompanying drawings.
本発明によれば、一次冷却材の基準温度は第６
図に示した従来のように一意的でなく、第１図に
例示するように、斜線部の領域内において可変で
ある。基準温度を目標に制御される一次冷却材の
温度が高くなると、炉心の安全性の余裕が減少す
るため、安全性の余裕が失われないように可変温
度の上限が設定される。また、一次冷却材の温度
が低下すると、原子炉の一次冷却系に接続された
蒸気発生器の蒸気圧力が低下するので、必要な蒸
気発生器圧力を維持するため可変温度領域の下限
が設定されている。 According to the present invention, the reference temperature of the primary coolant is the sixth
It is not unique as in the conventional case shown in the figure, but is variable within the shaded area as illustrated in FIG. As the temperature of the primary coolant, which is controlled with reference temperature as the target, increases, the safety margin of the core decreases, so an upper limit of the variable temperature is set so as not to lose the safety margin. Additionally, when the temperature of the primary coolant decreases, the steam pressure in the steam generator connected to the reactor primary cooling system decreases, so the lower limit of the variable temperature range is set to maintain the required steam generator pressure. ing.
第２図は、第１図の基準温度の可変領域内に一
次冷却材の基準温度を設定し原子炉の反応度を制
御する本発明に係る可変温度プログラム方式を実
現する制御系の一例を示している。第２図におい
て、原子炉１は、原子炉出力Ｑ、原子炉の一次冷
却材温度Tr、及び原子炉中性子束分布を表す指
標ΔI（原子炉炉心上半分の出力から下半分の出力
を引いたもの）で運転されているものとする。 FIG. 2 shows an example of a control system that implements the variable temperature program method according to the present invention, which sets the reference temperature of the primary coolant within the variable range of the reference temperature shown in FIG. 1 and controls the reactivity of the reactor. ing. In Figure 2, reactor 1 has a reactor output Q, a reactor primary coolant temperature Tr, and an index ΔI representing the reactor neutron flux distribution (the output of the lower half of the reactor core is subtracted from the output of the upper half of the reactor core). It is assumed that the vehicle is being driven by a vehicle.
タービン出力Qtは、定常運転時の基準温度演
算回路２に入力されている。演算回路２において
は、タービン出力Qtに対する第６図に示すよう
な関数の定常時基準となる一次冷却材の定常時基
準温度Trefが計算されている。タービン出力Qt、
一次冷却材温度Tr、制御棒位置Ｒ、原子炉出力
分布信号ΔIがプログラム演算回路３へ送られ、
一次冷却材の負荷追従運転時の基準温度Tprogが
演算される。一次冷却材温度Trと計算された基
準温度Tprogとは加減算回路４へ入力され、この
加減算回路４でこれ等の温度が比較され温度差信
号Te＝Tprog−Trが計算される。また、原子炉
出力Ｑとタービン出力Qtは加減算回路５で偏差
が計算され、この出力不一致信号と前記の温度差
信号Teとは、制御棒制御回路６へ送られる。制
御回路６の出力は制御棒駆動装置７へ送られ、該
駆動装置７による制御棒の位置調整により原子炉
の反応度を調整する。 The turbine output Qt is input to the reference temperature calculation circuit 2 during steady operation. In the arithmetic circuit 2, a steady state reference temperature Tref of the primary coolant is calculated as a steady state reference for the function shown in FIG. 6 for the turbine output Qt. Turbine output Qt,
The primary coolant temperature Tr, control rod position R, and reactor power distribution signal ΔI are sent to the program calculation circuit 3.
A reference temperature Tprog of the primary coolant during load following operation is calculated. The primary coolant temperature Tr and the calculated reference temperature Tprog are input to an addition/subtraction circuit 4, which compares these temperatures and calculates a temperature difference signal Te=Tprog-Tr. Further, the difference between the reactor output Q and the turbine output Qt is calculated by an adder/subtractor circuit 5, and this output mismatch signal and the temperature difference signal Te are sent to the control rod control circuit 6. The output of the control circuit 6 is sent to a control rod drive device 7, and the drive device 7 adjusts the position of the control rods to adjust the reactivity of the nuclear reactor.
一方、原子炉出力Ｑは、出力分布分布基準演算
回路８に送られ、出力分布制御の目標である基準
出力分布信号ΔIrefが原子炉出力Ｑの関数として
演算される。加減回路９は、出力分布信号ΔIと
基準出力分布信号ΔIrefとの差を計算する。ほう
酸濃度制御回路１０はほう酸濃度制御信号Bcを
演算し、化学体積制御設備１１のほう酸濃度調整
部を制御することにより、原子炉のほう酸濃度の
調整を行う。 On the other hand, the reactor power Q is sent to the power distribution distribution reference calculating circuit 8, and a reference power distribution signal ΔIref, which is the target of power distribution control, is calculated as a function of the reactor power Q. The adding/subtracting circuit 9 calculates the difference between the output distribution signal ΔI and the reference output distribution signal ΔIref. The boric acid concentration control circuit 10 calculates the boric acid concentration control signal Bc and controls the boric acid concentration adjustment section of the chemical volume control equipment 11 to adjust the boric acid concentration in the reactor.
炉心反応度ρの変化分Δρ（ｔ）は主に次の要因
からなる。 The change Δρ(t) in the core reactivity ρ mainly consists of the following factors.
Δρ（ｔ）＝αq・ΔQ（ｔ）＋αm・ΔT（ｔ）＋αx
・ΔXe（ｔ）＋Δρr（ｔ）＋αb・ΔB（ｔ）……(1)
ここで、
αq＝∂ρ／∂Q＝反応度の出力係数（負の値）
αm＝∂ρ／∂T＝反応度の冷却材温度係数（負の値）
αx＝∂ρ／∂Xe＝キセノンによる反応度係数（負の
値）
αb＝∂ρ／∂B＝ほう酸の反応度係数（負の値）
Δρr（ｔ）＝制御棒位置調整による時間関数の反
応度変化量
ΔQ（ｔ）＝時間関数の出力変化量
ΔT（ｔ）＝時間関数の一次冷却材の温度変化
ΔXe（ｔ）＝時間関数のキセノン蓄積量変化量
ΔB（ｔ）＝時間関数の炉心ほう酸濃度変化量
キセノンは核分裂生成物の崩壊過程で生じるも
ので、その炉心内蓄積量は原子炉出力の変動の履
歴から決まる。即ち、
ΔXe（ｔ）＝fx（Ｑ） ……(2)
ここで、
fx（Ｑ）＝出力変化の履歴により決まるキセノン
蓄積量の関数
キセノンは熱中性子の吸収断面積が大きく、蓄
積すると原子炉の反応度を低下させる。原子炉が
日負荷追従運転を行つている場合は、原子炉出力
の変化は急速ではないので、炉心反応度は略一定
とし、下記と近似していると考えてよい。Δρ(t) = αq・ΔQ(t)+αm・ΔT(t)+αx ・ΔXe(t)+Δρr(t)+αb・ΔB(t)……(1) Here, αq=∂ρ/∂Q=reaction αm = ∂ρ/∂T = Coolant temperature coefficient of reactivity (negative value) αx = ∂ρ/∂Xe = Reactivity coefficient due to xenon (negative value) αb = ∂ ρ/∂B = Reactivity coefficient of boric acid (negative value) Δρr (t) = Reactivity change in time function due to control rod position adjustment ΔQ (t) = Output change in time function ΔT (t) = Time function Temperature change of primary coolant ΔXe(t) = Change in xenon accumulation as a function of time ΔB(t) = Change in core boric acid concentration as a function of time Xenon is produced during the decay process of fission products, and its accumulation in the core The amount is determined by the history of reactor power fluctuations. That is, ΔXe (t) = fx (Q) ... (2) where, fx (Q) = function of xenon accumulation determined by the history of output changes Xenon has a large absorption cross section for thermal neutrons, and when accumulated, it decreases the reactivity of When the reactor is in daily load following operation, the reactor output does not change rapidly, so the core reactivity can be considered to be approximately constant and approximated as shown below.
Δρ（ｔ）＝０ ……(3)
(1)、(2)及び(3)式から、
αq・ΔQ（ｔ）＋αm・ΔT（ｔ）＋αx・ｘ（Ｑ）
＋Δρr（ｔ）＋αb・ΔB（ｔ）＝０
或は、
Δρr（ｔ）＋αb・ΔB（ｔ）
＝−〔αx・fx（Ｑ）＋αq・ΔQ（ｔ）
＋αm・ΔT（ｔ）〕 ……(4)
(4)式において、ΔT（ｔ）は一次冷却材温度の
変化であり、準定常的な負荷追従運転では一次冷
却材基準温度に略等しい。即ち、ΔT（ｔ）は、
基準温度Tprogの変化量となる。ここで、式(4)の
左辺Δρr（ｔ）＋αb・ΔB（ｔ）は制御棒の位置調整
及びほう酸の濃度調整量で、従来の制御方式にお
ける制御の操作量に該当するものである。従つ
て、式(4)の右辺において、ΔT（ｔ）、即ちTprog
の変化を、左辺に変化量及び変化の速さが小さく
なるように、αx・fx（Ｑ）＋αq・ΔQ（ｔ）の変化
を考慮して定めれば、左辺値、即ち制御棒とほう
酸濃度制御の負担が軽減され、負荷追従性能が向
上できる。 Δρ(t)=0 ...(3) From equations (1), (2) and (3), αq・ΔQ(t)+αm・ΔT(t)+αx・x(Q) +Δρr(t)+αb・ΔB (t) = 0 Or, Δρr(t) + αb・ΔB(t) =−[αx・fx(Q)+αq・ΔQ(t) +αm・ΔT(t)] ...(4) In equation (4) , ΔT(t) is a change in the primary coolant temperature, which is approximately equal to the primary coolant reference temperature in quasi-steady load following operation. That is, ΔT(t) is
This is the amount of change in the reference temperature Tprog. Here, Δρr(t)+αb·ΔB(t) on the left side of equation (4) is the amount of control rod position adjustment and boric acid concentration adjustment, which corresponds to the control operation amount in the conventional control system. Therefore, on the right side of equation (4), ΔT(t), that is, Tprog
If we determine the change in αx fx (Q) + αq ΔQ (t) so that the amount of change and speed of change become smaller on the left side, then the value on the left side, that is, the control rod and boric acid concentration The control burden is reduced and load following performance can be improved.
炉心の設計によつては、炉心の出力分布を適切
な値に維持するために、出力分布に大きな影響を
与える制御棒を望ましい位置に調整する必要が生
じる場合がある。炉心の出力分布は、制御棒位置
の他に、原子炉出力、キセノン蓄積の炉内分布等
によつて変わる。従つて、制御棒位置調整による
反応度変化であるΔρr（ｔ）は、原子炉の特性に
基づき、原子炉出力Ｑとその履歴から決まる制約
を受ける。即ち、
Δρr（ｔ）∈fr（Ｑ） ……(5)
ここで、
fr（Ｑ）＝原子炉出力Ｑとその履歴から決まる関
数で、出力分布の観点で定まる制御
棒による反応度変化許容範囲
である。(5)式の制約がある場合は、一次冷却材基
準温度は、(5)式を考慮し、(4)式のΔρr（ｔ）とΔB
（ｔ）の変化巾及び変化速度が小さくなるように、
定めることにより、負荷変化時の反応度調整が容
易になり、負荷追従能力を向上することができ
る。 Depending on the design of the reactor core, in order to maintain the power distribution of the core at an appropriate value, it may be necessary to adjust the control rods, which have a large effect on the power distribution, to a desired position. The power distribution of the reactor core varies depending on the control rod position, the reactor power, the distribution of xenon accumulation in the reactor, etc. Therefore, Δρr(t), which is the reactivity change due to control rod position adjustment, is subject to constraints determined by the reactor output Q and its history based on the characteristics of the reactor. That is, Δρr(t)∈fr(Q) ...(5) Here, fr(Q) = a function determined from the reactor output Q and its history, and the allowable range of reactivity change due to the control rod determined from the viewpoint of power distribution. It is. If there is a constraint in equation (5), the primary coolant reference temperature is determined by considering equation (5) and Δρr(t) and ΔB in equation (4).
(t) so that the range of change and the speed of change are small.
By determining this, it becomes easy to adjust the reactivity when the load changes, and the load following ability can be improved.
一次冷却材基準温度Tprogは、以上の検討から
分かるように、αx・fx（Ｑ）＋αq・ΔQ（ｔ）とfr
（Ｑ）とを考慮して決めることが望ましく、出力
変化と時間との関数が望ましいことを知ることが
できる。 As can be seen from the above considerations, the primary coolant reference temperature Tprog is calculated by αx・fx(Q)+αq・ΔQ(t) and fr
It can be seen that it is desirable to decide by taking (Q) into consideration, and that a function of output change and time is desirable.
昼間は高出力運転を行い、夜間は出力を下げて
運転を行う日負荷追従運転において、第３図で示
すような出力変化を行つた場合に、炉心出力分布
の歪みがなく、式(5)を完全に満足する形で制御棒
を動かした場合に、ほう酸濃度調整が０となるよ
うな一次冷却材の温度変化を原子炉特性に基づい
て計算した例を第４図に示す。第４図は、望まし
い一次冷却材温度の変化を原子炉出力の関数とし
て示したものである。この図から、ほう酸濃度調
整量が少なく、出力分布の歪みの小さい制御のた
めに有利な一次冷却材基準温度は、従来の直線的
温度変化と異なることが分かる。即ち、出力下降
時は従来の出力下降時ほど下降せず、また、50％
出力に達してから後、キセノンの蓄積量の過渡的
増加に伴つて、ゆつくり低下していくような形と
なる。50％出力から100％出力へ復帰する場合も
同様な傾向を示している。 In daily load following operation, in which high output operation is performed during the day and operation is performed with reduced output at night, when the output changes as shown in Figure 3, there is no distortion in the core power distribution, and Equation (5) Figure 4 shows an example of calculating the temperature change of the primary coolant that would cause the boric acid concentration adjustment to be 0 when the control rods are moved in a manner that completely satisfies the following, based on the reactor characteristics. FIG. 4 shows the desired change in primary coolant temperature as a function of reactor power. From this figure, it can be seen that the primary coolant reference temperature, which requires a small amount of boric acid concentration adjustment and is advantageous for controlling output distribution with small distortion, is different from the conventional linear temperature change. In other words, when the output decreases, it does not decrease as much as the conventional output decrease, and the output decreases by 50%.
After reaching the output, the output gradually decreases as the accumulated amount of xenon increases transiently. A similar trend is shown when returning from 50% output to 100% output.
このことから、日負荷追従運転では、一次冷却
材基準温度は、(4)式、(5)式及び原子炉特性から許
容される可変の範囲内で決定するか、或は第４図
に示された形に近似した形に、一次冷却材基準温
度を決定することが負荷追従運転上有利である。 Therefore, in daily load following operation, the primary coolant standard temperature should be determined within the variable range allowed by equations (4) and (5) and the reactor characteristics, or it should be determined as shown in Figure 4. It is advantageous for load following operation to determine the primary coolant reference temperature in a form that approximates the given form.
このような望ましい一次冷却材基準温度は、第
２図に示した制御系を使用し、本発明に従つて次
のような様々な態様で実現することができる。 Such a desired primary coolant reference temperature can be achieved in the following various ways according to the present invention using the control system shown in FIG.
次に、第２図のプログラム演算回路３に於いて
一次冷却材基準温度を定める方法について述べ
る。 Next, a method for determining the primary coolant reference temperature in the program calculation circuit 3 shown in FIG. 2 will be described.
原子炉の通常運転時には、設計上及び運転管理
上望ましいように、タービン出力の一意関数とし
て、第６図のように定常運転時基準温度を定め
る。しかし、原子炉の負荷追従運転の過渡時には
次のような関数で、一次冷却材基準温度を定め
る。 During normal operation of a nuclear reactor, a reference temperature during steady operation is determined as a unique function of turbine output, as shown in FIG. 6, as desired in terms of design and operational management. However, during transient load following operation of a nuclear reactor, the primary coolant reference temperature is determined by the following function.
Tprog＝〔∫f1（Tref−Tprog）dt〕 ……(6)
ここで、
Tprog＝一次冷却材基準温度
Tref＝一次冷却材の定常時基準温度
f1＝（Tref−Tprog）の関数
〔 〕＝基準温度の上限、下限で制限すること
を示す
∫dt＝時間に関する積分
(6)式は、負荷変動でTrefが変化すると、
TprogがTprogとTrefの偏差の関数、例えば
Tprog−Trefに比例した速さでTrefへ復帰する
ことを示す。 Tprog=[∫f 1 (Tref−Tprog)dt] …(6) Where, Tprog=primary coolant reference temperature Tref=steady state reference temperature of primary coolant f 1 =function of (Tref−Tprog) [ ] = Indicates that it is limited by the upper and lower limits of the reference temperature ∫dt = Integral with respect to time Equation (6) shows that when Tref changes due to load fluctuation,
Tprog is a function of the deviation of Tprog and Tref, e.g.
It shows that the return to Tref occurs at a speed proportional to Tprog−Tref.
第３図に示す100％出力と50％出力間の日負荷
追従運転を行つた時の(6)式に基づく一次冷却材基
準温度の変化は第５図に示すようになり、第４図
に示す望ましい変化に近似したものとなる。従つ
て、この方式においては、タービン出力の関数と
した定常運転時の一次冷却材基準温度と、タービ
ン出力の関数とした、負荷追従運転時の一次冷却
材基準温度の変化範囲を制限する上限温度及び下
限温度とを定め、負荷追従運転時の一次冷却材基
準温度をその上下限温度範囲内で、負荷追従運転
時の該一次冷却材基準温度と定常運転時の一次冷
却材基準温度との間の偏差の関数として、定常運
転時の一次冷却材基準温度に接近するよう制御す
ることにより、負荷追従能力が向上する。 The change in the primary coolant standard temperature based on equation (6) when performing daily load follow-up operation between 100% output and 50% output shown in Figure 3 is as shown in Figure 5. This approximates the desired change shown. Therefore, in this method, there is a primary coolant reference temperature during steady operation as a function of turbine output, and an upper limit temperature that limits the range of change in the primary coolant reference temperature during load following operation as a function of turbine output. and lower limit temperature, and set the primary coolant reference temperature during load following operation within the upper and lower limit temperature range, and between the primary coolant reference temperature during load following operation and the primary coolant reference temperature during steady operation. The load following ability is improved by controlling the temperature to approach the primary coolant reference temperature during steady operation as a function of the deviation of .
実施例 ２
次に、前記実施例１で実現可能とした第５図の
一次冷却材基準温度の変化は次式でも実現するこ
とができる。Embodiment 2 Next, the change in the primary coolant reference temperature shown in FIG. 5, which can be realized in the first embodiment, can also be realized by the following equation.
Tprog＝１／１＋τs・Tref
ここで、
ｓ＝ラプラス演算子
τ＝時定数
即ち、タービン出力の関数とした定常運転時の
一次冷却材基準温度と、タービン出力の関数とし
た、負荷追従運転時の一次冷却材基準温度の変化
範囲を制限する上限温度及び下限温度とを定め、
負荷追従運転時の一次冷却材基準温度をその上下
限温度範囲内で、上記定常運転時の一次冷却材基
準温度に遅れをもつて追従する関数により、上記
定常運転時の一次冷却材基準温度に近付くように
制御する。ここで、時定数τの値は、第６図のよ
うな一次冷却材基準温度の変化を与え、第４図に
示す変化に近いように定めるもので、一般には、
キセノン蓄積量の過渡的変化に見合つたものとし
て、１時間乃至数時間の値となる。 Tprog = 1/1 + τs・Tref where, s = Laplace operator τ = time constant In other words, the primary coolant reference temperature during steady operation as a function of turbine output, and the primary coolant reference temperature during load following operation as a function of turbine output. Defining an upper limit temperature and a lower limit temperature that limit the range of change in the primary coolant reference temperature,
The primary coolant reference temperature during load following operation is adjusted to the primary coolant reference temperature during steady operation by a function that follows the primary coolant reference temperature during steady operation with a delay within its upper and lower limit temperature ranges. Control it to get closer. Here, the value of the time constant τ is determined so as to give a change in the primary coolant reference temperature as shown in FIG. 6, and to be close to the change shown in FIG. 4, and generally,
The values range from one hour to several hours, commensurate with the transient changes in xenon accumulation.
以上の説明から明らかなように、加圧水型原子
炉の一次冷却材基準温度の制御方式において、一
次冷却材基準温度を可変とし、第１、第２の発明
（実施例１、実施例２）にかかる方法により一次
冷却材基準温度を決定する制御方式を用いた場
合、制御棒の位置調整及びほう酸濃度調整の負担
が軽減され、制御性が向上し、反応度制御能力の
増大より、原子炉の負荷追従能力の向上が実現で
きる。更に、制御棒の位置調整及びほう酸濃度調
整の負担軽減に付随して、制御棒の機械的寿命の
延長、ほう酸濃度制御系の容量低減、ほう酸水処
理量の低減が計られる利点がある。
As is clear from the above explanation, in the control method for the primary coolant reference temperature of a pressurized water reactor, the primary coolant reference temperature is made variable, and the first and second inventions (Example 1 and Example 2) When using a control method that determines the primary coolant reference temperature using this method, the burden of adjusting the control rod position and boric acid concentration is reduced, controllability is improved, and the reactor is improved by increasing the reactivity control ability. It is possible to improve the load following ability. Furthermore, in addition to reducing the burden of adjusting the position of the control rods and adjusting the boric acid concentration, there are also advantages in extending the mechanical life of the control rods, reducing the capacity of the boric acid concentration control system, and reducing the amount of boric acid water throughput.
第１図は本発明に従つて制御される一次冷却材
基準温度と出力変化との関数を示すグラフ図、第
２図は本発明に従つて冷却材温度を制御するため
の原子炉制御系のブロツク図、第３図は日負荷追
従における原子炉出力と時間との関係を示すグラ
フ図、第４図は本発明に従つてほう酸濃度調整が
０となるように制御した場合の一次冷却材基準温
度と原子炉出力との関係を示すグラフ図、第５図
は本発明による別の方式に従つて制御したときの
一次冷却材基準温度と原子炉出力との関係を示す
グラフ図、第６図は従来の一次冷却材基準温度の
制御態様を示すグラフ図である。
１……原子炉、２……基準温度演算回路、３…
…プログラム演算回路、６……制御棒制御回路、
１０……ほう酸濃度制御回路。
FIG. 1 is a graphical diagram showing a function of the primary coolant reference temperature and power change controlled according to the present invention, and FIG. 2 is a graph of the reactor control system for controlling the coolant temperature according to the present invention. Block diagram, Figure 3 is a graph showing the relationship between reactor output and time in daily load tracking, and Figure 4 is the primary coolant standard when boric acid concentration adjustment is controlled to 0 according to the present invention. FIG. 5 is a graph showing the relationship between temperature and reactor power; FIG. 5 is a graph showing the relationship between primary coolant reference temperature and reactor power when controlled according to another method according to the present invention; FIG. FIG. 2 is a graph diagram showing a conventional control mode of the primary coolant reference temperature. 1... Nuclear reactor, 2... Reference temperature calculation circuit, 3...
...Program calculation circuit, 6...Control rod control circuit,
10...boric acid concentration control circuit.
Claims (1)
次冷却材の基準温度を計算するプログラム演算回
路と、この基準温度に一次冷却材温度を制御すべ
く、原子炉出力分布を制御するための制御棒を制
御する制御棒制御回路と、一次冷却材中のほう酸
濃度を制御するほう酸濃度制御回路とを備え、制
御棒の位置調整及びほう酸の濃度調整によつて負
荷追従制御する加圧水型原子炉の冷却材温度制御
系において、タービン出力の関数とした定常運転
時の一次冷却材基準温度と、タービン出力の関数
とした、負荷追従運転時の一次冷却材基準温度の
変化範囲を制限する上限温度及び下限温度とを定
め、負荷追従運転時の一次冷却材基準温度をその
上下限温度範囲内で、該負荷追従運転時の一次冷
却材基準温度と定常運転時の一次冷却材基準温度
との間の偏差の関数として、定常運転時の一次冷
却材基準温度に接近するよう制御することを特徴
とする、加圧水型原子炉の負荷追従運転用冷却材
温度制御方法。 ２ 原子炉の負荷追従運転時の制御のために、一
次冷却材の基準温度を計算するプログラム演算回
路と、この基準温度に一次冷却材温度を制御すべ
く、原子炉出力分布を制御するための制御棒を制
御する制御棒制御回路と、一次冷却材中のほう酸
濃度を制御するほう酸濃度制御回路とを備え、制
御棒の位置調整及びほう酸の濃度調整によつて負
荷追従制御する加圧水型原子炉の冷却材温度制御
系において、タービン出力の関数とした定常運転
時の一次冷却材基準温度と、タービン出力の関数
とした、負荷追従運転時の一次冷却材基準温度の
変化範囲を制限する上限温度及び下限温度とを定
め、負荷追従運転時の一次冷却材基準温度をその
上下限温度範囲内で、上記定常運転時の一次冷却
材基準温度に遅れをもつて追従する関数により、
上記定常運転時の一次冷却材基準温度に近付くよ
うに制御することを特徴とする加圧水型原子炉の
負荷追従運転用冷却材温度制御方法。[Claims] 1. A program calculation circuit that calculates a reference temperature of a primary coolant for control during load following operation of a nuclear reactor, and a program calculation circuit that calculates a reference temperature of a primary coolant to control the reactor output to this reference temperature. Equipped with a control rod control circuit that controls the control rods to control distribution and a boric acid concentration control circuit that controls the concentration of boric acid in the primary coolant, the load can be followed by adjusting the position of the control rods and the concentration of boric acid. In the coolant temperature control system of a pressurized water reactor to be controlled, changes in the primary coolant reference temperature during steady operation as a function of turbine output and the primary coolant reference temperature during load following operation as a function of turbine output. An upper limit temperature and a lower limit temperature that limit the range are determined, and the primary coolant reference temperature during load following operation is set within the upper and lower limit temperature range, and the primary coolant reference temperature during load following operation and the primary cooling temperature during steady operation are set. A method for controlling a coolant temperature for load following operation of a pressurized water reactor, characterized in that the coolant temperature is controlled to approach the primary coolant reference temperature during steady operation as a function of the deviation from the material reference temperature. 2. A program calculation circuit for calculating the reference temperature of the primary coolant for control during load following operation of the reactor, and a program calculation circuit for controlling the reactor power distribution in order to control the primary coolant temperature to this reference temperature. A pressurized water reactor that is equipped with a control rod control circuit that controls control rods and a boric acid concentration control circuit that controls boric acid concentration in the primary coolant, and that performs load follow-up control by adjusting the control rod position and boric acid concentration. In the coolant temperature control system, the upper limit temperature limits the range of change in the primary coolant reference temperature during steady operation as a function of turbine output and the primary coolant reference temperature during load following operation as a function of turbine output. and lower limit temperature, and by a function that follows the primary coolant reference temperature during steady operation with a delay within the upper and lower limit temperature range of the primary coolant reference temperature during load following operation,
A method for controlling a coolant temperature for load following operation of a pressurized water reactor, characterized in that the temperature is controlled so as to approach the primary coolant reference temperature during steady operation.
Priority Applications (3)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60121514A JPS61280600A (en) | 1985-06-06 | 1985-06-06 | Control coolant temperature for loading followup operation of pressurized water type nuclear reactor |
| FR868608134A FR2583207B1 (en) | 1985-06-06 | 1986-06-05 | PROCESS FOR CONTROLLING THE TEMPERATURE OF THE COOLING FLUID OF A REACTOR FOR OPERATION AFTER THE LOAD OF A NUCLEAR POWER PLANT |
| DE19863618871 DE3618871A1 (en) | 1985-06-06 | 1986-06-05 | METHOD FOR CONTROLLING THE REACTOR COOLANT TEMPERATURE FOR A LIVE TRACKING OPERATION OF A NUCLEAR POWER PLANT |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60121514A JPS61280600A (en) | 1985-06-06 | 1985-06-06 | Control coolant temperature for loading followup operation of pressurized water type nuclear reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS61280600A JPS61280600A (en) | 1986-12-11 |
| JPH047960B2 true JPH047960B2 (en) | 1992-02-13 |
Family
ID=14813088
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP60121514A Granted JPS61280600A (en) | 1985-06-06 | 1985-06-06 | Control coolant temperature for loading followup operation of pressurized water type nuclear reactor |
Country Status (3)
| Country | Link |
|---|---|
| JP (1) | JPS61280600A (en) |
| DE (1) | DE3618871A1 (en) |
| FR (1) | FR2583207B1 (en) |
Families Citing this family (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP1374006B1 (en) * | 2001-03-29 | 2006-03-01 | Pebble Bed Modular Reactor (Proprietary) Limited | A method of and control system for controlling a nuclear reactor outlet temperature |
| FR2985363B1 (en) | 2011-12-29 | 2015-01-30 | Areva Np | METHOD FOR CONTROLLING A PRESSURIZED WATER NUCLEAR REACTOR |
| KR102160064B1 (en) * | 2018-09-06 | 2020-09-25 | 한국수력원자력 주식회사 | System for load following operation including adjusting concentration of Boron and operating system using the same |
| JP7245112B2 (en) * | 2019-05-20 | 2023-03-23 | 三菱重工業株式会社 | Reactor control device, nuclear power plant and method of controlling nuclear reactor |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5930097A (en) * | 1982-08-13 | 1984-02-17 | 三菱原子力工業株式会社 | Coolant temperature control system of pwr type reactor |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3423285A (en) * | 1966-01-27 | 1969-01-21 | Westinghouse Electric Corp | Temperature control for a nuclear reactor |
| JPS56163497A (en) * | 1980-05-21 | 1981-12-16 | Tokyo Shibaura Electric Co | Method and device for operating follow-up atomic power plant |
| GB2122409B (en) * | 1982-06-17 | 1985-10-16 | Westinghouse Electric Corp | Method for controlling a nuclear fueled electric power generating unit and interfacing the same with a load dispatching system |
-
1985
- 1985-06-06 JP JP60121514A patent/JPS61280600A/en active Granted
-
1986
- 1986-06-05 DE DE19863618871 patent/DE3618871A1/en not_active Ceased
- 1986-06-05 FR FR868608134A patent/FR2583207B1/en not_active Expired - Lifetime
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5930097A (en) * | 1982-08-13 | 1984-02-17 | 三菱原子力工業株式会社 | Coolant temperature control system of pwr type reactor |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS61280600A (en) | 1986-12-11 |
| FR2583207B1 (en) | 1991-10-18 |
| DE3618871A1 (en) | 1986-12-11 |
| FR2583207A1 (en) | 1986-12-12 |
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