CN104200061A - Method for calculating power of third-generation nuclear power station pressurized water reactor core - Google Patents

Abstract

The invention provides a method for calculating power of a third-generation nuclear power station pressurized water reactor core. The method comprises the steps that according to signals measured on site, after the thermal power WSG of a steam generator, the thermal power WdeltaPr input by other thermal sources and the relative uncertainty uWR of a measuring meter, a collection board card, a calculation formula and the like are calculated, the power WR of the pressurized water reactor core is calculated according to the formula WR=WSG-WdeltaPr+ uWR. The method for calculating the power of the third-generation nuclear power station pressurized water reactor core can be used for checking the high-power level protection function and the reactor thermal power of a protection system of a nuclear power station, and calculation precision is further improved on the basis of simplifying the calculation formula.

Description

A kind of generation Ⅲ nuclear power station pressurized water reactor core power calculation method

Technical field

The invention belongs to nuclear power technology field, particularly a kind of PWR of Nuclear Power Station core power measuring technique.

Background technology

The calculating of third generation pressurized water reactor core power; be mainly used in the high power levels defencive function of protection system of nuclear power station and the check of heat output of reactor; to the safety of nuclear power plant reactor and economical operation; play very important effect; and the pressurized-water reactor nuclear power plant of the existing nuclear power technology of China is substantially all the second generation, the second generation half; the heat calculation method using; also can only be for the calculating of the core thermal power of the pressurized-water reactor nuclear power plant of the second generation, the second generation half nuclear power technology, cannot be for the calculating of generation Ⅲ nuclear power technology pressurized water reactor core thermal power.

In addition, in existing heat calculation method, the calculating to properties of water and steam, use be the formula of old standard, in computational accuracy, also have much room for improvement.

Therefore, need to be applied to the computing method of generation Ⅲ nuclear power technology pressurized water reactor core thermal power.

Summary of the invention

Can be for the computing method of generation Ⅲ nuclear power technology pressurized water reactor core thermal power in order to solve, can be in heat calculation method, need to calculate the parameters such as thermodynamic properties of the water and steam in thermal balance equation, and the parameters such as feedwater flow in calculating thermal balance equation, and then obtain reactor core thermal power.The result of calculation of core thermal power, also needs to consider that analog input card that sensor, transmitter, collection signal that measuring-signal is used use is, the thermal power of the uncertainty that the links such as computing formula of the use in thermal balance equation cause and the input of other thermals source.

Therefore; the invention provides a kind of high power levels defencive function of protection system of nuclear power station and generation Ⅲ nuclear power station pressurized water reactor core power calculation method of the check of heat output of reactor of can be used for; can, according to the signal of in-site measurement, calculate steam generator thermal power W _sG, the input of other thermals source thermal power W _{Δ Pr}, relative uncertainty uW _rafter, then according to formula W _r=W _sG-W _{Δ Pr}+ uW _rcalculate the power W of pressurized water reactor core _r, described relative uncertainty uW _rcomprise: the uncertainty of measurement instrument, analog input card, computing formula.

Wherein, described steam generator thermal power W _sGcan be by formula: W _sG=(H _v-H _e) Q _e-(H _v-H _p) Q _pcalculate, wherein, H _vfor enthalpy of wet steam, H _ebe main feedwater enthalpy, H _pfor blowdown enthalpy, Q _ebe main feedwater flow, Q _pfor blowdown flow, described blowdown flow Q _pfor the signal of in-site measurement, by steam generator blowdown system, APG measures, span 0-17kg/s.

1. the enthalpy of wet steam H in the present invention _vby formula: H _v=xH " _v+ (1-x) H' _vobtain, wherein: the span 0.95-1 that x is steam quality, H " _vfor saturated vapour enthalpy, H' _vfor saturation water enthalpy, wherein,

1.1 saturated vapour enthalpy H " _vby formula:

${H^{\″}}_v=\mathrm{\τRt}\×[\underset{i=1}{\overset{9}{\mathrm{\Σ}}}{n}_i^o{J}_i^o{\mathrm{\τ}}^{J_i^o-1}+\underset{i=1}{\overset{43}{\mathrm{\Σ}}}{n}_i{\mathrm{\π}}^{I_i}{J}_i{(\mathrm{\τ}-0.5)}^{J_i-1}]$ Obtain, wherein

π=P _vvp/ p ^*, τ=T ^*/ t, p ^*=1MPa, T ^*=540K, R=0.461526kJkg ^-1k ^-1, P _vvpfor main steam pressure, be the signal of in-site measurement, by main steam system, VVP measures, span 0-10MPa, t is saturation temperature value, is calculated kJkg by formula (1) ^-1k ^-1unit dimension, i, , for coefficient, come from standard I APWS-IF97;

1.2 saturation water enthalpy H' _vby formula

${H^{\′}}_v=\mathrm{\τRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}$ Obtain, wherein

π=P/p ^*, τ=T ^*/ t, p ^*=16.53MPa, T ^*=1386K, R=0.461526kJkg ^-1k ^-1, t is saturation temperature value, is calculated i, I by formula (1) _i, J _i, n _ifor coefficient, come from standard I APWS-IF97;

1.3 saturation temperature t are by formula:

$\frac{t}{T^{*}}=\frac{n_{10}+D-{\text{[}{(n_{10}+D)}^2\text{-4}(n_9+n_{10}D)\text{]}}^{0.5}}{2}---\left(1\right)$ Obtain, wherein

T ^*＝1K

$D=\frac{2G}{-F-{(F^2-4\mathrm{EG})}^{0.5}}$

E＝β ²+n ₃β+n ₆

F＝n ₁β ²+n ₄β+n ₇

G＝n ₂β ²+n ₅β+n ₈

β＝(P _vvp/p ^*) ^0.25

p ^*＝1MPa

I, n _ifor coefficient, come from standard I APWS-IF97.

2. blowdown enthalpy H in the present invention, _pwith saturation water enthalpy H' _vvalue is identical, i.e. H _p=H' _v.

3. main feedwater enthalpy H in the present invention, _eby formula

$H_e={\mathrm{\τRt}}^{\′}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}$ Obtain, wherein, π=P _are/ p ^*, τ=T*/T _are, p ^*=16.53MPa, T ^*=1386K, R=0.461526kJkg ^-1k ^-1, i, I _i, J _i, n _ifor coefficient, come from standard I APWS-IF97,

P _arebe main feed pressure, T _arebe main feed temperature, described main feed pressure P _arewith described main feed temperature T _are, be in-site measurement signal, by main feedwater flow system ARE, measured, wherein, described main feed pressure P _arespan 0-10MPa, described main feed temperature T _arespan 0-250 ℃.

4. in the present invention, main feedwater flow Q _eby formula

obtain, wherein, π=P _are/ p ^*, C is efflux coefficient, by formula (2), is calculated; ρ is density, by formula (3), is calculated; ε is fluid inflatable coefficient, ε=1 here,

D is orifice throttle bore dia, β is the ratio of orifice plate internal diameter and internal diameter of the pipeline, the ratio beta of described orifice throttle bore dia d and described orifice plate internal diameter and internal diameter of the pipeline is in-site measurement signal, take from the parameter of system equipment, wherein, the value of described orifice throttle bore dia d is 292mm, and the value of the ratio beta of described orifice plate internal diameter and internal diameter of the pipeline is 0.7

P _arebe main feed pressure, Δ P is main to differential water pressures, described main feed pressure P _arebe in-site measurement signal with described main feedwater pressure differential deltap P, by main feedwater flow system ARE, measured, wherein, described main feed pressure P _arespan 0-10MPa, the span 0-1500mbar of described main feedwater pressure differential deltap P;

4.1 efflux coefficient C are by formula

$\begin{array}{c}C=0.5961+0.0261{\mathrm{\β}}^2-0.216{\mathrm{\β}}^8+0.000521{\left(\frac{{10}^6\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.7}+(0.0188+0.0063A){\mathrm{\β}}^{3.5}{\left(\frac{{10}^6}{{\mathrm{Re}}_D}\right)}^{0.3}\\ +(0.043+0.080e^{-10L_1}-0.123e^{-7L_1})(1-0.11A)\frac{{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}-0.031({M^{\′}}_2-0.8{{M^{\′}}_2}^{1.1}){\mathrm{\β}}^{1.3}\end{array}$ (2) obtain, wherein:

Re _dfor the Reynolds number after iteration, by step (4), calculated;

L ₁(=l ₁/ D _f) for orifice plate upstream face arrives the distance of upstream pressure tappings divided by the value of pipe diameter,

L' ₂(=l' ₂/ D _f) for orifice plate downstream end face arrives the distance of downstream pressure tappings divided by the value of pipe diameter,

D _ffor orifice plate upstream pressure flange internal diameter, take from the parameter of system equipment, value is 419mm,

For corner pressure tapping: L ₁=L' ₂=25.4/D _f,

For D and D/2 pressure tappings: L ₁=1, L' ₂=0.47,

For flange pressure tappings: L ₁=L' ₂=25.4/D _f,

${M^{\′}}_2=\frac{2{L^{\′}}_2}{1-\mathrm{\β}},A={\left(\frac{19000\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.8};$

4.2 feedwater density p are by formula

$\mathrm{\ρ}=1/[\mathrm{\πRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}-n_i{I}_i{(7.1-\mathrm{\π})}^{I_i-1}{(\mathrm{\τ}-1.222)}^{J_i}/P]---\left(3\right)$ Obtain, wherein

π=P _are/ p ^*, τ=T ^*/ T _are, p ^*=16.53MPa, T ^*=1386K, R=0.461526kJkg ^-1k ^-1, i, I _i, J _i, n _ifor coefficient, come from standard I APWS-IF97;

Reynolds number Re after 4.3 iteration _dby formula

${\mathrm{Re}}_D=\frac{{4Q}_e}{\mathrm{\π}{\mathrm{\μ}}_1{D}_F}---\left(4\right)$ Obtain, wherein

μ ₁for feedwater kinetic viscosity, by formula (5), calculated;

4.4 feedwater kinetic viscosity μ ₁by formula

${\mathrm{\μ}}_1={\mathrm{\θ}}^{0.5}{\left[\underset{i=1}{\overset{4}{\mathrm{\Σ}}}{n}_i^o{\mathrm{\θ}}^{1-i}\right]}^{-1}\exp \left[\mathrm{\δ}\underset{i=1}{\overset{21}{\mathrm{\Σ}}}{n}_i{(\mathrm{\δ}-1)}^{I_i}{({\mathrm{\θ}}^{-1}-1)}^{J_i}\right]\×{\mathrm{\μ}}^{*}---\left(5\right)$ Obtain, wherein

δ=ρ/ρ ^*, τ=T ^*/ T _are, ρ ^*=ρ _c=322kgm ^-3, T ^*=T _c=647.096K, μ ^*=1 * 10 ^-6pas, ρ is density value, by formula (6.2), calculated, i, for coefficient, come from standard I APWS-IF97.

Preferably, steam generator has many can have many, and preferably 4, the power of pressurized water reactor core is: wherein, it is the general power of 4 steam generator thermal powers.

5. due to the existence of measuring error, measured true value is difficult to determine, measurement result is with uncertainty.Uncertainty is an important indicator of evaluation measurement result quality height.Uncertainty is less, and the quality of measurement result is higher.Uncertainty of measurement refers to not affirming of measurement result variation, is to characterize measured true value in an estimation of certain value scope, and be estimation and dispersed two parts of measuring measured value.Because the calculating parameter in algorithm comes from sensor, transmitter, analog input card, computing formula etc., therefore need to consider the uncertainty of these links.To obtain a result of calculation more accurately, so in invention, uncertainty uW _rby formula:

${\mathrm{uW}}_R={\left\{\begin{array}{c}{\left(\frac{W_{\mathrm{SG}1}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}1}}{W_{\mathrm{SG}1}}\right)}^2+{\left(\frac{W_{\mathrm{SG}2}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}2}}{W_{\mathrm{SG}2}}\right)}^2+{\left(\frac{W_{\mathrm{SG}3}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}3}}{W_{\mathrm{SG}3}}\right)}^2+\\ {\left(\frac{W_{\mathrm{SG}4}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}4}}{W_{\mathrm{SG}4}}\right)}^2{+}^{{\left[\frac{W_{\mathrm{\ΔPr}}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{\ΔPr}}}{W_{\mathrm{\ΔPr}}}\right]}^2}\end{array}\right\}}^{1/2}$ Obtain, wherein,

W _sG1be the 1st steam generator thermal power, W _sG2be the 2nd steam generator thermal power, W _sG3be the 3rd steam generator thermal power, W _sG4be the 4th separate unit steam generator thermal power, described every steam generator thermal power is by described formula W _sG=(H _v-H _e) Q _e-(H _v-H _p) Q _pobtain;

be the 1st the relative uncertainty of steam generator thermal power, be the 2nd the relative uncertainty of steam generator thermal power, be the 3rd the relative uncertainty of steam generator thermal power, be the 4th the relative uncertainty of steam generator thermal power, the relative uncertainty calculation method of thermal power of described every steam generator is identical, by formula

$\frac{\mathrm{\Δ}{W}_{\mathrm{SG}}}{W_{\mathrm{SG}}}={\left\{\begin{array}{c}{\left[\frac{H_v(Q_e-Q_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_v}{H_v}\right]}^2+{\left(\frac{H_e{Q}_e}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_e}{H_e}\right)}^2+{\left(\frac{H_p{Q}_p}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_p}{H_p}\right)}^2+\\ {\left[\frac{Q_e(H_v-H_e)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_e}{Q_e}\right]}^2+{\left[\frac{Q_p(H_v-H_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_p}{Q_{[}}\right]}^2\end{array}\right\}}^{1/2}$ Obtain, wherein:

for the relative uncertainty of steam generator outlet enthalpy, by the vapor pressure uncertainty of measurement of described main steam system VVP and water vapour thermodynamic properties, calculate uncertainty and obtain;

be the relative uncertainty of main Enthalpy of Feed Water, the uncertainty of being measured by the main feed temperature of described main feedwater flow control system ARE and water vapour thermodynamic properties calculate uncertainty and obtain;

for the relative uncertainty of blowdown enthalpy, equal the uncertainty of saturation water enthalpy, by the vapor pressure uncertainty of measurement of main steam system VVP and properties of water and steam, calculate uncertainty and obtain;

for the relative uncertainty of feedwater flow, the relative uncertainty of described feedwater flow is obtained by step (10.1);

for the relative uncertainty of dirty flow, from process design parameter, span 12.5%-25%;

for the thermal power uncertainty of other thermal source inputs, from process design parameter, span 12.5%-25%;

In the present invention, feedwater flow relative uncertainty is by formula:

$\frac{\mathrm{\δQe}}{\mathrm{Qe}}=\sqrt{{\left(\frac{\mathrm{\δC}}{C}\right)}^2+{\left(\frac{\mathrm{\δ\ϵ}}{\mathrm{\ϵ}}\right)}^2+{\left(\frac{2{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}\right)}^2+{\left(\frac{\mathrm{\δ}{D}_F}{D_F}\right)}^2+{\left(\frac{2}{1-{\mathrm{\β}}^4}\right)}^2{\left(\frac{\mathrm{\δd}}{d}\right)}^2+\frac{1}{4}{\left[\frac{\mathrm{\δ}\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}{\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}\right]}^2+\frac{1}{4}{\left(\frac{\mathrm{\δ}{\mathrm{\ρ}}_e}{{\mathrm{\ρ}}_e}\right)}^2}$ Obtain.

Generation Ⅲ nuclear power provided by the invention station pressurized water reactor core power calculation method, also comprises the thermal power W that other thermals source are inputted _{Δ Pr}, the thermal power W of other thermal source inputs _{Δ Pr}calculating comprise: the heat that the heat that the heat that the heat that main pump is brought into, replenishment pump are brought into, voltage stabilizer are brought into, non-regenerative heat exchanger are taken away, the heat taken away of sealing water-to-water heat exchanger, heat and the system thermal loss that reactor cooling system is taken away, under normal circumstances, the thermal power W of described other thermal source inputs _{Δ Pr}numerical value change little, preferred, described W _{Δ Pr}=26MW.

In the present invention, to the isoparametric calculating of the thermodynamic properties of water and steam, used the computing formula in current industrial new standard IAPWS-IF97.Calculating to feedwater flow parameter, has been used the computing formula in new standard ISO5167-2003.To uncertain calculating, applied in theory of errors and standard the computing formula about uncertainty.

The invention has the beneficial effects as follows, the calculating of third generation pressurized water reactor core power, be mainly used in the high power levels defencive function of protection system of nuclear power station and the check of heat output of reactor, to the safety of nuclear power plant reactor and economical operation, play very important effect, and the pressurized-water reactor nuclear power plant of the existing nuclear power technology of China is substantially all the second generation, the second generation half, the heat calculation method using, also can only be for the second generation, the calculating of the core thermal power of the pressurized-water reactor nuclear power plant of the second generation half nuclear power technology, cannot be for the calculating of generation Ⅲ nuclear power technology pressurized water reactor core thermal power.

In addition, use the computing method of generation Ⅲ nuclear power technology pressurized water reactor core thermal power provided by the invention simplifying on the basis of computing formula, further improved computational accuracy.

Accompanying drawing explanation

Below in conjunction with accompanying drawing, generation Ⅲ nuclear power of the present invention station pressurized water reactor core power calculation method is specifically described.

Fig. 1 is schematic flow sheet of the present invention;

Fig. 2 is secondary circuit heat balance principle signal of the present invention.

Embodiment

1, generation Ⅲ nuclear power station pressurized water reactor core power calculation side ratio juris

By the measured value of the main feed pressure of main feedwater flow control system (ARE), main feed temperature, main feedwater flow differential pressure, the sewer flow measurements of the vapor pressure measured value of main steam system (VVP) and steam generator blowdown system (APG), the heat that calculates a circuit cools agent medium is passed to secondary circuit to the enthalpy liter of water generates by steam generator (SG), calculate again the thermal power of a circuit cools agent system, and then deduct the thermal power that other thermals source are inputted to a circuit cools agent system except reactor core, and then obtain reactor core thermal power.In conjunction with the uncertainty of correlation formula, evaluate the quality height of the result of calculation of core thermal power again.

2, the step of computing method

2.1 according to the principle of computing method, can obtain the computing formula of reactor core thermal power

$W_R=\underset{i=1}{\overset{4}{\mathrm{\Σ}}}{W}_{\mathrm{SGi}}-W_{\mathrm{\ΔPr}}---\left(1\right)$

In formula:

-W _sGifor separate unit steam generator thermal power, computing method are shown in 2.2 formula (2);

-W _{Δ Pr}for the thermal power of other thermal source inputs, W _{Δ Pr}calculating comprise: main pump is studied in greatly volume heat, heat that replenishment pump is brought into, the heat that voltage stabilizer is brought into, the heat that non-regenerative heat exchanger is taken away, the heat that sealing water-to-water heat exchanger is taken away, heat and the system thermal loss that reactor cooling system is taken away, under normal circumstances, the thermal power W of described other thermal source inputs _{Δ Pr}numerical value change little, preferred, described W _{Δ Pr}=26MW.

2.2 separate unit steam generator thermal power W _sGcomputing formula

W _SG＝(H _v-H _e)Q _e-(H _v-H _p)Q _p (2)

In formula:

-H _vfor enthalpy of wet steam, computing method are shown in 2.3 formula (3);

-H _ebe main feedwater enthalpy, computing method are shown in 2.7 formula (7);

-H _pfor blowdown enthalpy, computing method are shown in 2.8 formula (8);

-Q _ebe main feedwater flow, computing method are shown in 2.9 formula (9);

-Q _pfor blowdown flow, the sewer flow measurement by steam generator blowdown system (APG) obtains.

2.3 enthalpy of wet steam H _vcomputing formula

H _v＝xH" _v+(1-x)H' _v (3)

In formula:

-x is steam quality, is the important technical indicator of nuclear power station steam generator system.Refer to the mass dryness fraction of steam here, its span 0.95-1.

-H " _vfor saturated vapour enthalpy, computing method are shown in 2.4 formula (4);

-H' _vfor saturation water enthalpy, computing method are shown in 2.6 formula (6).

2.4 saturated vapour enthalpy H " _vcomputing formula

${H^{\″}}_v=\mathrm{\τRt}\×[\underset{i=1}{\overset{9}{\mathrm{\Σ}}}{n}_i^o{J}_i^o{\mathrm{\τ}}^{J_i^o-1}+\underset{i=1}{\overset{43}{\mathrm{\Σ}}}{n}_i{\mathrm{\π}}^{I_i}{J}_i{(\mathrm{\τ}-0.5)}^{J_i-1}]---\left(4\right)$

In formula:

π＝P _vvp/p ^*，τ＝T ^*/t，p ^*＝1MPa，T ^*＝540K，R＝0.461526kJkg ^-1K ^-1

-P _vvpfor saturation pressure value, the force value that the pressure transducer by main steam system (VVP) measures;

-t is saturation temperature value, and computing method are shown in 2.5 formula (5).

-kJkg ^-1k ^-1it is unit dimension

In formula, coefficient comes from standard I APWS-IF97, specific as follows:

The computing formula of 2.5 saturation temperature t

$\frac{t}{T^{*}}=\frac{n_{10}+D-{\text{[}{(n_{10}+D)}^2\text{-4}(n_9+n_{10}D)\text{]}}^{0.5}}{2}---\left(5\right)$

In formula:

T ^*=1K; K is unit dimension, Kelvin temperature

$D=\frac{2G}{-F-{(F^2-4\mathrm{EG})}^{0.5}}$

E＝β ²+n ₃β+n ₆

F＝n ₁β ²+n ₄β+n ₇

G＝n ₂β ²+n ₅β+n ₈

β＝(P _vvp/p ^*) ^0.25

p ^*＝1MPa

-P _vvpfor saturation pressure value, the force value that the pressure transducer by main steam system (VVP) measures, in formula, coefficient is as follows, comes from standard I APWS-IF97:

2.6 saturation water enthalpy H' _vcomputing formula

${H^{\′}}_v=\mathrm{\τRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}---\left(6\right)$

In formula:

π＝P _vvp/p ^*，τ＝T ^*/t，p ^*＝16.53MPa，T ^*＝1386K，R＝0.461526kJkg ^-1K ^-1

-P _vvpfor saturation pressure value, the force value that the pressure transducer by main steam system (VVP) measures.

-t is saturation temperature value, and computing method are shown in 2.5 formula (5).

In formula, coefficient is as follows, comes from standard I APWS-IF97:

2.7 main feedwater enthalpy H _ecomputing formula

$H_e={\mathrm{\τRt}}^{\′}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}---\left(7\right)$

In formula:

π＝P _are/p ^*，τ＝T ^*/T _are，p ^*＝16.53MPa，T ^*＝1386K，R＝0.461526kJkg ^-1K ^-1

-P _arefor feed pressure value, the pressure transducer by main feedwater flow control system (ARE) measures;

-T _arefor feed temperature value, the temperature sensor measurement by main feedwater flow control system (ARE) obtains, main feed temperature T _are200 degrees Celsius of left and right;

In formula, coefficient is as follows, comes from standard I APWS-IF97:

2.8 blowdown enthalpy H _pcomputing formula

H _p＝H' _v (8)

In formula:

-H' _vfor saturation water enthalpy, computing method are shown in 2.6 formula (6).

2.9 main feedwater flow Q _ecomputing formula

$Q_e=\frac{C}{\sqrt{1-{\mathrm{\β}}^4}}\mathrm{\ϵ}\frac{\mathrm{\π}}{4}{d}^2\sqrt{2\mathrm{\Δ}{P}_{\mathrm{\ρ}}}---\left(9\right)$

In formula:

-d is the orifice throttle bore dia under actual operating mode, from device parameter value;

-β is orifice plate internal diameter under actual operating mode and the ratio of internal diameter of the pipeline, from device parameter value;

-C is efflux coefficient, and computing method are shown in 2.10 formula (10);

-ρ is density, and computing method are shown in 2.13 formula (13);

-Δ P is differential pressure value, the measuring of the main feedwater flow differential pressure by main feedwater flow control system (ARE);

-ε is fluid inflatable coefficient, incompressible fluid ε=1.The fluid here refers to water.

The computing formula of 2.10 efflux coefficient C

$\begin{array}{c}C=0.5961+0.0261{\mathrm{\β}}^2-0.216{\mathrm{\β}}^8+0.000521{\left(\frac{{10}^6\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.7}+(0.0188+0.0063A){\mathrm{\β}}^{3.5}{\left(\frac{{10}^6}{{\mathrm{Re}}_D}\right)}^{0.3}\\ +(0.043+0.080e^{-10L_1}-0.123e^{-7L_1})(1-0.11A)\frac{{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}-0.031({M^{\′}}_2-0.8{{M^{\′}}_2}^{1.1}){\mathrm{\β}}^{1.3}\end{array}---\left(10\right)$

Reynolds number Re after-iteration _d, computing method are shown in 2.11 formula (11);

-β is orifice plate internal diameter under actual operating mode and the ratio of internal diameter of the pipeline, and from device parameter value, wherein parameter value is taken from standard ISO 5167-2003;

-L ₁(=l ₁/ D _f) for orifice plate upstream face arrives the distance of upstream pressure tappings divided by the value of pipe diameter, wherein parameter value is taken from standard ISO 5167-2003.

-L' ₂(=l' ₂/ D _f) for orifice plate downstream end face arrives the distance of downstream pressure tappings divided by the value of pipe diameter, wherein parameter value is taken from standard ISO 5167-2003;

For corner pressure tapping: L ₁=L' ₂=25.4/D _f;

For D and D/2 pressure tappings: L ₁=1, L' ₂=0.47;

For flange pressure tappings: L ₁=L' ₂=25.4/D _f.

-D _ffor the orifice plate upstream pressure flange internal diameter under actual operating mode, from device parameter value, wherein parameter value is taken from standard ISO 5167-2003.

In formula

${M^{\′}}_2=\frac{2{L^{\′}}_2}{1-\mathrm{\β}}---(10-1)$

$A={\left(\frac{19000\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.8}---(10-2)$

Reynolds number Re after 2.11 iteration _dcomputing formula, be the formula in ISO5167-2003 standard.

${\mathrm{Re}}_D=\frac{4Q_e}{\mathrm{\π}{\mathrm{\μ}}_1{D}_F}---\left(11\right)$

Under-supposed situation, use main feedwater flow Q _e, with main feedwater, by flow formula substitution hypothesis efflux coefficient, try to achieve here, be the iterative calculation method providing in ISO5167-2003 standard.

-feedwater kinetic viscosity μ ₁, computing method are shown in 2.12 formula (12);

-D _ffor the orifice plate upstream pressure flange internal diameter under actual operating mode, from device parameter value, wherein parameter value is taken from standard ISO 5167-2003.

2.12 feedwater kinetic viscosity μ ₁computing formula

${\mathrm{\μ}}_1={\mathrm{\θ}}^{0.5}{\left[\underset{i=1}{\overset{4}{\mathrm{\Σ}}}{n}_i^o{\mathrm{\θ}}^{1-i}\right]}^{-1}\exp \left[\mathrm{\δ}\underset{i=1}{\overset{21}{\mathrm{\Σ}}}{n}_i{(\mathrm{\δ}-1)}^{I_i}{({\mathrm{\θ}}^{-1}-1)}^{J_i}\right]\×{\mathrm{\μ}}^{*}---\left(12\right)$

In formula:

δ＝ρ/ρ ^*，θ＝T _are/T ^*，ρ ^*＝ρ _c＝322kgm ^-3，T ^*＝T _c＝647.096K，μ ^*＝1×10 ^-6Pa·s

-ρ is density value, and computing method are shown in 2.13 formula (13);

-T _arefor temperature value, the temperature sensor by main feedwater flow control system (ARE) measures, the number of degrees with actual measurement to numerical value be as the criterion, generally 200 degrees Celsius of left and right.

In formula, coefficient is as follows, comes from standard I APWS-IF97:

The computing formula of 2.13 feedwater density p

$\mathrm{\ρ}=1/[\mathrm{\πRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}-n_i{I}_i{(7.1-\mathrm{\π})}^{I_i-1}{(\mathrm{\τ}-1.222)}^{J_i}/P]---\left(13\right)$

In formula:

π＝P _are/p ^*，τ＝T ^*/T _are，p ^*＝16.53MPa，T ^*＝1386K，

R＝0.461526kJkg ^-1K ^-1，

-P _arebe main feed pressure, the pressure transducer by main feedwater flow control system (ARE) measures,

-T _arebe main feed temperature, the temperature sensor measurement by main feedwater flow control system (ARE) obtains,

In formula, coefficient is as follows, comes from standard I APWS-IF97:

The computing formula problem of the relative uncertainty of 2.14 separate unit steam generator thermal power:

$\frac{\mathrm{\Δ}{W}_{\mathrm{SG}}}{W_{\mathrm{SG}}}={\left\{\begin{array}{c}{\left[\frac{H_v(Q_e-Q_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_v}{H_v}\right]}^2+{\left(\frac{H_e{Q}_e}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_e}{H_e}\right)}^2+{\left(\frac{H_p{Q}_p}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_p}{H_p}\right)}^2+\\ {\left[\frac{Q_e(H_v-H_e)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_e}{Q_e}\right]}^2+{\left[\frac{Q_p(H_v-H_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_p}{Q_{[}}\right]}^2\end{array}\right\}}^{1/2}---\left(14\right)$

In formula:

-H _vfor enthalpy of wet steam, computing method are shown in 2.3 formula (3);

-H _ebe main feedwater enthalpy, computing method are shown in 2.7 formula (7);

-H _pfor blowdown enthalpy, computing method are shown in 2.8 formula (8);

-Q _ebe main feedwater flow, computing method are shown in 2.9 formula (9);

-Q _pfor blowdown flow, the sewer flow measurement by steam generator blowdown system (APG) obtains;

-W _sGfor separate unit steam generator thermal power, computing method are shown in 2.2 formula (2);

for the relative uncertainty of steam generator outlet enthalpy, by vapor pressure uncertainty of measurement and the properties of water and steam calculating uncertainty of main steam system (VVP), determined;

be the relative uncertainty of main Enthalpy of Feed Water, the uncertainty of being measured by the main feed temperature of main feedwater flow control system (ARE) and water vapour thermodynamic properties calculate uncertainty and determine;

for the relative uncertainty of blowdown enthalpy, equal the uncertainty of saturation water enthalpy, by vapor pressure uncertainty of measurement and the properties of water and steam calculating uncertainty of main steam system (VVP), determined;

for the relative uncertainty of feedwater flow, computing method are shown in 2.14 formula (14-1);

for the relative uncertainty of dirty flow, from process design parameter.

According in international standard ISO5167-2:2003 about the regulation of orifice plate uncertainty, the error calculation formula of mass rate is as follows:

$\frac{\mathrm{\δQe}}{\mathrm{Qe}}=\sqrt{{\left(\frac{\mathrm{\δC}}{C}\right)}^2+{\left(\frac{\mathrm{\δ\ϵ}}{\mathrm{\ϵ}}\right)}^2+{\left(\frac{2{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}\right)}^2+{\left(\frac{\mathrm{\δ}{D}_F}{D_F}\right)}^2+{\left(\frac{2}{1-{\mathrm{\β}}^4}\right)}^2{\left(\frac{\mathrm{\δd}}{d}\right)}^2+\frac{1}{4}{\left[\frac{\mathrm{\δ}\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}{\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}\right]}^2+\frac{1}{4}{\left(\frac{\mathrm{\δ}{\mathrm{\ρ}}_e}{{\mathrm{\ρ}}_e}\right)}^2}---(14-1)$

The computing formula of the relative uncertainty of thermal power that 2.15 steam generators are total

$\frac{\mathrm{\ΔW}}{W}={\left\{\begin{array}{c}{\left(\frac{W_{\mathrm{SG}1}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}1}}{W_{\mathrm{SG}1}}\right)}^2+{\left(\frac{W_{\mathrm{SG}2}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}2}}{W_{\mathrm{SG}2}}\right)}^2+{\left(\frac{W_{\mathrm{SG}3}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}3}}{W_{\mathrm{SG}3}}\right)}^2+\\ {\left(\frac{W_{\mathrm{SG}4}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}4}}{W_{\mathrm{SG}4}}\right)}^2+{\left[\frac{W_{\mathrm{\Δ}{P}_r}}{W}\frac{\mathrm{\Δ}\left(W_{\mathrm{\Δ}{P}_r}\right)}{W_{\mathrm{\Δ}{P}_r}}\right]}^2\end{array}\right\}}^{1/2}---\left(15\right)$

In formula:

-W _sG1, W _sG2, W _sG3, W _sG4for separate unit steam generator thermal power, computing method are shown in 2.2 formula (2);

- for the relative uncertainty of separate unit steam generator thermal power, computing method are shown in 2.14 formula (14);

-W is the total thermal power of steam generator,

-W _{Δ Pr}for the thermal power of other thermal source inputs, this numerical value is obtained by test;

for the thermal power uncertainty of other thermal source inputs, from process design parameter, be design load.

3. calculating other thermals source passes to the computing formula of the heat WPr of reactor cooling system and is: positive energy exchange=main pump is studied in greatly the heat that heat+voltage stabilizer that volume heat+replenishment pump brings into brings into-(heat+system thermal loss that heat+reactor cooling system of taking away of heat+sealing water-to-water heat exchanger of taking away of non-regenerative heat exchanger is taken away), under normal circumstances, the numerical value change of WPr is little, here get constant, the total thermal loss of reactor circuit is 26MW.

The above is only preferred embodiment of the present invention, not the present invention is done to any pro forma restriction, although the present invention discloses as above with preferred embodiment, yet not in order to limit the present invention, any technician who is familiar with this patent is not departing within the scope of technical solution of the present invention, when can utilizing the technology contents of above-mentioned prompting to make a little change or being modified to the equivalent embodiment of equivalent variations, in every case be the content that does not depart from technical solution of the present invention, any simple modification of above embodiment being done according to technical spirit of the present invention, equivalent variations and modification, all still belong in the present invention program's scope.

Claims (9)

1. a generation Ⅲ nuclear power station pressurized water reactor core power calculation method, for the high power levels defencive function of the protection system of nuclear power station and the check of heat output of reactor, is characterized in that, according to the signal of in-site measurement, calculates steam generator thermal power W _sGwith relative uncertainty uW _rafter, then according to formula W _r=W _sG+ uW _rcalculate the power W of pressurized water reactor core _r, described relative uncertainty uW _rcomprise: the uncertainty of measurement instrument, analog input card, computing formula.

2. a kind of generation Ⅲ nuclear power according to claim 1 station pressurized water reactor core power calculation method, is characterized in that, according to W _sG=(H _v-H _e) Q _e-(H _v-H _p) Q _pcalculate described steam generator thermal power W _sG, wherein, H _vfor enthalpy of wet steam, H _ebe main feedwater enthalpy, H _pfor blowdown enthalpy, Q _ebe main feedwater flow, Q _pfor blowdown flow,

Described enthalpy of wet steam H _vwith described blowdown enthalpy H _p, by the signal of in-site measurement: main steam pressure P _vvpthrough calculating;

Described main feedwater enthalpy H _e, by the signal of in-site measurement: main feed pressure P _arewith main feed temperature T _arethrough calculating;

Described Q _ebe main feedwater flow, by the signal of in-site measurement: main feed pressure P _are, main feedwater pressure differential deltap P and main feed temperature T _arethrough calculating;

Described blowdown flow Q _pfor the signal of in-site measurement, by steam generator blowdown system, APG measures.

3. a kind of generation Ⅲ nuclear power according to claim 2 station pressurized water reactor core power calculation method, is characterized in that, described enthalpy of wet steam H _vby formula: H _v=xH " _v+ (1-x) H' _vobtain, wherein: x is steam quality, span is 0.95-1, H " _vfor saturated vapour enthalpy, H' _vfor saturation water enthalpy, wherein,

(3.1) described saturated vapour enthalpy H " _vby formula:

${H^{\″}}_v=\mathrm{\τRt}\×[\underset{i=1}{\overset{9}{\mathrm{\Σ}}}{n}_i^o{J}_i^o{\mathrm{\τ}}^{J_i^o-1}+\underset{i=1}{\overset{43}{\mathrm{\Σ}}}{n}_i{\mathrm{\π}}^{I_i}{J}_i{(\mathrm{\τ}-0.5)}^{J_i-1}]$ Obtain, wherein

π=P _vvp/ p ^*, τ=T ^*/ t, p ^*=1MPa, T ^*=540K, R=0.461526kJkg ^-1k ^-1, described main steam pressure P _vvp, be the signal of in-site measurement, by main steam system, VVP measures, and t is saturation temperature value, is calculated kJkg by step (3.3) ^-1k ^-1unit dimension, i, , for coefficient, come from standard I APWS-IF97;

(3.2) described saturation water enthalpy H' _vby formula

${H^{\′}}_v=\mathrm{\τRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}$ Obtain, wherein

π＝P _vvp/p ^*，τ＝T ^*/t，p ^*＝16.53MPa，T ^*＝1386K，R＝0.461526kJkg ^-1K ^-1，

T is saturation temperature value, is calculated i, I by step (3.3) _i, J _i, n _ifor coefficient, come from standard I APWS-IF97;

(3.3) described saturation temperature t is by formula:

$\frac{t}{T^{*}}=\frac{n_{10}+D-{\text{[}{(n_{10}+D)}^2\text{-4}(n_9+n_{10}D)\text{]}}^{0.5}}{2}$ Obtain, wherein

T ^*＝1K

$D=\frac{2G}{-F-{(F^2-4\mathrm{EG})}^{0.5}}$

E＝β ²+n ₃β+n ₆

F＝n ₁β ²+n ₄β+n ₇

G＝n ₂β ²+n ₅β+n ₈

β＝(P _vvp/p ^*) ^0.25

p ^*＝1MPa

I, n _ifor coefficient, come from standard I APWS-IF97.

4. according to a kind of generation Ⅲ nuclear power station pressurized water reactor core power calculation method described in claim 2 or 3, it is characterized in that described blowdown enthalpy H _pwith described saturation water enthalpy H' _vvalue is identical, i.e. H _p=H' _v.

5. a kind of generation Ⅲ nuclear power according to claim 2 station pressurized water reactor core power calculation method, is characterized in that, described main feedwater enthalpy H _eby formula

$H_e={\mathrm{\τRt}}^{\′}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}{n}_i{(7.1-\mathrm{\π})}^{I_i}{J}_i{(\mathrm{\τ}-1.222)}^{J_i-1}$ Obtain, wherein, π=P _are/ p ^*, τ=T*/T _are, p ^*=16.53MPa, T ^*=1386K, R=0.461526kJkg ^-1k ^-1, i, I _i, J _i, n _ifor coefficient, come from standard I APWS-IF97,

P _arebe main feed pressure, T _arebe main feed temperature, described main feed pressure P _arewith described main feed temperature T _are, be in-site measurement signal, by main feedwater flow system ARE, measured.

6. a kind of generation Ⅲ nuclear power according to claim 2 station pressurized water reactor core power calculation method, is characterized in that, described main feedwater flow Q _eby formula

obtain, wherein, π=P _are/ p ^*, C is efflux coefficient, by step (6.1), is calculated; ρ is density, by step (6.2), is calculated; ε is fluid inflatable coefficient, ε=1 here,

D is orifice throttle bore dia, β is the ratio of orifice plate internal diameter and internal diameter of the pipeline, the ratio beta of described orifice throttle bore dia d and described orifice plate internal diameter and internal diameter of the pipeline is in-site measurement signal, takes from the parameter of system equipment, and described parameter value is taken from standard ISO 5167-2003.

P _arebe main feed pressure, Δ P is main to differential water pressures, described main feed pressure P _arebe in-site measurement signal with described main feedwater pressure differential deltap P, by main feedwater flow system ARE, measured;

(6.1) described efflux coefficient C is by formula

$\begin{array}{c}C=0.5961+0.0261{\mathrm{\β}}^2-0.216{\mathrm{\β}}^8+0.000521{\left(\frac{{10}^6\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.7}+(0.0188+0.0063A){\mathrm{\β}}^{3.5}{\left(\frac{{10}^6}{{\mathrm{Re}}_D}\right)}^{0.3}\\ +(0.043+0.080e^{-10L_1}-0.123e^{-7L_1})(1-0.11A)\frac{{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}-0.031({M^{\′}}_2-0.8{{M^{\′}}_2}^{1.1}){\mathrm{\β}}^{1.3}\end{array}$ Obtain, wherein:

Re _dfor the Reynolds number after iteration, by step (6.3), calculated;

L ₁(=l ₁/ D _f) for orifice plate upstream face arrives the distance of upstream pressure tappings divided by the value of pipe diameter,

L' ₂(=l' ₂/ D _f) for orifice plate downstream end face arrives the distance of downstream pressure tappings divided by the value of pipe diameter,

D _ffor orifice plate upstream pressure flange internal diameter, take from the parameter of system equipment, described parameter value is taken from standard ISO 5167-2003;

For corner pressure tapping: L ₁=L' ₂=25.4/D _f,

For D and D/2 pressure tappings: L ₁=1, L' ₂=0.47,

For flange pressure tappings: L ₁=L' ₂=25.4/D _f,

${M^{\′}}_2=\frac{2{L^{\′}}_2}{1-\mathrm{\β}},A={\left(\frac{19000\mathrm{\β}}{{\mathrm{Re}}_D}\right)}^{0.8};$

(6.2) described feedwater density p is by formula

$\mathrm{\ρ}=1/[\mathrm{\πRt}\×\underset{i=1}{\overset{34}{\mathrm{\Σ}}}-n_i{I}_i{(7.1-\mathrm{\π})}^{I_i-1}{(\mathrm{\τ}-1.222)}^{J_i}/P]$ Obtain, wherein

π=P _are/ p ^*, τ=T ^*/ T _are, p ^*=16.53MPa, T ^*=1386K, R=0.461526kJkg ^-1k ^-1, i, I _i, J _i, n _ifor coefficient, come from standard I APWS-IF97;

(6.3) reynolds number Re after described iteration _dby formula

${\mathrm{Re}}_D=\frac{{4Q}_e}{\mathrm{\π}{\mathrm{\μ}}_1{D}_F}$ Obtain, wherein

μ ₁for feedwater kinetic viscosity, by step (6.4), calculated;

(6.4) described feedwater kinetic viscosity μ ₁by formula

${\mathrm{\μ}}_1={\mathrm{\θ}}^{0.5}{\left[\underset{i=1}{\overset{4}{\mathrm{\Σ}}}{n}_i^o{\mathrm{\θ}}^{1-i}\right]}^{-1}\exp \left[\mathrm{\δ}\underset{i=1}{\overset{21}{\mathrm{\Σ}}}{n}_i{(\mathrm{\δ}-1)}^{I_i}{({\mathrm{\θ}}^{-1}-1)}^{J_i}\right]\×{\mathrm{\μ}}^{*}$ Obtain, wherein

δ=ρ/ρ ^*, θ=T _are/ T ^*, ρ ^*=ρ _c=322kgm ^-3, T ^*=T _c=647.096K, μ ^*=1 * 10 ^-6pas, ρ is density value, by formula (6.2), calculated, i, for coefficient, come from standard I APWS-IF97.

7. generation Ⅲ nuclear power according to claim 1 station pressurized water reactor core power calculation method, is characterized in that, described steam generator has many, and preferably 4, the power of described pressurized water reactor core is: wherein, it is the general power of 4 steam generator thermal powers.

8. according to the arbitrary generation Ⅲ nuclear power station pressurized water reactor core power calculation method described in claim 1-7, it is characterized in that described uncertainty uW _rby formula:

${\mathrm{uW}}_R={\left\{\begin{array}{c}{\left(\frac{W_{\mathrm{SG}1}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}1}}{W_{\mathrm{SG}1}}\right)}^2+{\left(\frac{W_{\mathrm{SG}2}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}2}}{W_{\mathrm{SG}2}}\right)}^2+{\left(\frac{W_{\mathrm{SG}3}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}3}}{W_{\mathrm{SG}3}}\right)}^2+\\ {\left(\frac{W_{\mathrm{SG}4}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{SG}4}}{W_{\mathrm{SG}4}}\right)}^2{+}^{{\left[\frac{W_{\mathrm{\ΔPr}}}{W}\frac{\mathrm{\Δ}{W}_{\mathrm{\ΔPr}}}{W_{\mathrm{\ΔPr}}}\right]}^2}\end{array}\right\}}^{1/2}$ Obtain, wherein,

W _sG1be the 1st steam generator thermal power, W _sG2be the 2nd steam generator thermal power, W _sG3be the 3rd steam generator thermal power, W _sG4be the 4th separate unit steam generator thermal power, described every steam generator thermal power is by described formula W _sG=(H _v-H _e) Q _e-(H _v-H _p) Q _pobtain;

be the 1st the relative uncertainty of steam generator thermal power, be the 2nd the relative uncertainty of steam generator thermal power, be the 3rd the relative uncertainty of steam generator thermal power, be the 4th the relative uncertainty of steam generator thermal power, the relative uncertainty calculation method of thermal power of described every steam generator is identical, by formula

$\frac{\mathrm{\Δ}{W}_{\mathrm{SG}}}{W_{\mathrm{SG}}}={\left\{\begin{array}{c}{\left[\frac{H_v(Q_e-Q_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_v}{H_v}\right]}^2+{\left(\frac{H_e{Q}_e}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_e}{H_e}\right)}^2+{\left(\frac{H_p{Q}_p}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{H}_p}{H_p}\right)}^2+\\ {\left[\frac{Q_e(H_v-H_e)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_e}{Q_e}\right]}^2+{\left[\frac{Q_p(H_v-H_p)}{W_{\mathrm{SG}}}\frac{\mathrm{\Δ}{Q}_p}{Q_{[}}\right]}^2\end{array}\right\}}^{1/2}$ Obtain, wherein:

for the relative uncertainty of steam generator outlet enthalpy, by the vapor pressure uncertainty of measurement of described main steam system VVP and water vapour thermodynamic properties, calculate uncertainty and obtain;

be the relative uncertainty of main Enthalpy of Feed Water, the uncertainty of being measured by the main feed temperature of described main feedwater flow control system ARE and water vapour thermodynamic properties calculate uncertainty and obtain;

for the relative uncertainty of blowdown enthalpy, equal the uncertainty of saturation water enthalpy, by the vapor pressure uncertainty of measurement of main steam system VVP and properties of water and steam, calculate uncertainty and obtain;

for the relative uncertainty of feedwater flow, the relative uncertainty of described feedwater flow is obtained by step (10.1);

for the relative uncertainty of dirty flow, from process design parameter;

for the thermal power uncertainty of other thermal source inputs, from process design parameter;

(8.1) the relative uncertainty of described feedwater flow is by formula:

$\frac{\mathrm{\δQe}}{\mathrm{Qe}}=\sqrt{{\left(\frac{\mathrm{\δC}}{C}\right)}^2+{\left(\frac{\mathrm{\δ\ϵ}}{\mathrm{\ϵ}}\right)}^2+{\left(\frac{2{\mathrm{\β}}^4}{1-{\mathrm{\β}}^4}\right)}^2+{\left(\frac{\mathrm{\δ}{D}_F}{D_F}\right)}^2+{\left(\frac{2}{1-{\mathrm{\β}}^4}\right)}^2{\left(\frac{\mathrm{\δd}}{d}\right)}^2+\frac{1}{4}{\left[\frac{\mathrm{\δ}\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}{\left({\mathrm{\ΔP}}_{\mathrm{eM}}\right)}\right]}^2+\frac{1}{4}{\left(\frac{\mathrm{\δ}{\mathrm{\ρ}}_e}{{\mathrm{\ρ}}_e}\right)}^2}$ Obtain.

9. generation Ⅲ nuclear power according to claim 1 station pressurized water reactor core power calculation method, is characterized in that, also comprises the thermal power W of other thermal source inputs _{Δ Pr}, the power W of described pressurized water reactor core _rcomputing formula be: W _r=W _sG-W _{Δ Pr}+ uW _r, wherein, the thermal power W of described other thermal source inputs _{Δ Pr}calculating comprise: the heat that the heat that the heat that the heat that main pump is brought into, replenishment pump are brought into, voltage stabilizer are brought into, non-regenerative heat exchanger are taken away, the heat taken away of sealing water-to-water heat exchanger, heat and the system thermal loss that reactor cooling system is taken away, under normal circumstances, the thermal power W of described other thermal source inputs _{Δ Pr}numerical value change little, preferred, described W _{Δ Pr}=26MW.