US 8,553,829 B2Granted

Reduced order stress model for online maneuvering, diagnostics of fuel failure and design of core loading patterns of light water reactors

AssigneeAreva Np

InventorYousef Farawila

Life of this patent

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Abstract

The invention is principally directed to a reduced order model, XEDOR, facilitating the prediction of and the diagnostics of pellet-clad interaction stress-corrosion-cracking failure of nuclear fuel rods. The invention more particularly relates to assessment of susceptibility to PCI failure for guidance in the design of fuel loading in nuclear reactors. The invention additionally relates to the protection against PCI failure by providing operational information to operators of a nuclear reactor during power maneuvering, including predictive calculations prior to executing power maneuvers. Additionally, the invention relates to the diagnostics of an event suggesting a possible PCI cladding failure.

Claims21

Dependent claims

A method for prediction of a plurality of variables characterizing fuel performance, utilizing a reduced order model computer program, applied to nuclear reactor operations, and comprising:

a. loading a computer program into a computer;
inputting, into the computer, reactor core instrumentation measurements;
using said measurements to provide input to core monitoring software;
accessing the monitoring software generated operational data of linear heat generation rate, neutron flux, burnup, fast neutron fluence, for all or the majority of fuel rods at several axial locations;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and,
b. calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power
and,
c. the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
d. the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
e. a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
f. displaying the calculated parameters, σ and/or δ, and
g. said displayed calculated parameters prompting nuclear reactor operators in taking operational steps for nuclear reactor operations.

71 more elements

A method for prediction of variables in fuel performance, as used in a nuclear reactor utilizing a reduced order model computer program for nuclear reactor operations consisting essentially of:

a. inputting into a computer a reduced order computer code wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations for use in nuclear reactor operations;
and;
said reduced order computer code consisting essentially of;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and
b. calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power
and,
c. the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
d. the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
e. a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
c. displaying said calculations, σ and/or δ, in physical operator displays for operator consideration where calculations are physically based for nuclear reactor operations safety relative to empirical maneuvering guides;
d. demonstrating, via the operator display, calculations which are predictive of the stress level and the associated linear heat generation rate margin to the maximum allowable limits, prompting, by knowing the definite representation of each fuel design parameters and material properties, nuclear reactor operators in the operation of nuclear reactors.

70 more elements

A method for prediction of variables in fuel performance, as used in a nuclear reactor, utilizing a core monitoring system where algorithms are encoded in a computer program for the purpose of calculating stress in fuel rod segments prior to performing nuclear reactor operations comprising:

a. inputting a computer program encoding the computer program algorithms into a computer;
b. measuring, via core instrumentation, core operational data required for core monitoring software to generate detailed operational parameters for each fuel rod in the core at a plurality of axial locations or segments in the form of computer data arrays;
the computer program algorithms accepting as input the data arrays representing the fuel rod segment data;
and;
the computer program consisting essentially of:
c. inputting into a computer a reduced order computer code wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations for use in nuclear reactor operations;
and;
said reduced order computer code consisting essentially of;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and,
d. calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power
and
e. the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
f. the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
g. a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
h. receiving calculations, from the computer program algorithms manipulation of the input data, and outputting the calculations for nuclear reactor operator operations of a nuclear reactor.

74 more elements

A method for prediction of, protection against, reduction in the likelihood of and the diagnostics of pellet-clad interaction failure of nuclear fuel rods utilizing a core monitoring system where computer program algorithms are encoded in a computer program and are loaded in a computer having a computer-readable storage medium having computer program logic stored thereon for enabling a processor to execute the computer program algorithms, upon receipt of operational data from a reactor core monitoring system, calculate margin to PCI failure, said calculations used for reactor operations for executing a safe power maneuver;

computer program algorithms comprising inputting into a computer a computer program implementing the algorithms for calculating the gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
,
(
4
)
and,
calculating the evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
,
(
3
)
where t is time, δ is the pellet-clad size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress for given fast neutron flux level and cumulative fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and;
where the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is the linear power in kW/m, pu is the fraction of unstable pores in the ceramic pellet material, and ps is the fraction of stable pores in the same material and,
where the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
and:
additionally, inputting, into the computer, core instrumentation measurements, using said measurements to provide input to core monitoring software, accessing the monitoring software generated operational data of linear heat generation rate, neutron flux, burnup, fast neutron fluence, for all or the majority of the fuel rods at several axial locations and calculating, by the computer program, the gap size and clad stress when the gap is closed;
thereafter, outputting, to an operators display, the stress response to a planned power increase, as operational guidance for power maneuvering.

66 more elements

A method for prediction of, protection against, prevention of and the diagnostics of pellet-clad interaction failure of nuclear fuel rods utilizing a core monitoring system where computer program algorithms are encoded in a computer program and are loaded in a computer performing the steps of:

receiving operational data from a reactor core monitoring system;
calculating stress level in cladding of a plurality of fuel rods and the associated linear power margin to the maximum allowable limits assigned to PCI failure mode, using definite representation of each fuel type design parameters and material properties and;
where said computer program algorithms consist essentially of:
inputting into a computer a reduced order computer code wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations for use in nuclear reactor operations;
and;
said reduced order computer code consisting essentially of;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and
calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power
and,
the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
displaying said calculations to nuclear reactor operators to execute a safe PCI failure free power maneuver such that the said limits are not exceeded.

72 more elements

A method for prediction of, protection against, reduction in the likelihood of and the diagnostics of pellet-clad interaction failure of nuclear fuel rods, during power maneuvering, using a computer program as coupled to a core monitoring system comprising:

a. using data provided online by a nuclear plant monitoring and instrumentation system;
calculating the stress distribution in many or all fuel rods in the reactor core at several axial locations;
b. calculating a margin parameter, for each fuel rod segment representing an axial location, to identify the margin to reaching an operator provided stress limit where such parameter can be the linear power required to reach the stress limit, or the linear power increase above the existing level at the time required to reach the stress limit, or the ratio between the maximum linear power associated with the stress limit and the existing linear power at the time, or other parameters which provide the operator with information regarding the operation margin to reaching the stress limit;
and;
where said computer program algorithms consist essentially of:
inputting into a computer a reduced order computer code wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations for use in nuclear reactor operations;
and;
said reduced order computer code consisting essentially of;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and,
calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and
the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
c. projecting, in a predictive manner, the time evolution of the linear power at each rod segment in the core designated for such calculation conditional on stress remaining at or below an operator provided value;
d. performing the above functions online while directly connected to the reactor instrumentation and monitoring software;
e. displaying the margin parameters to reaching an operator provided stress limit for nuclear reactor operators in operations of a nuclear reactor in power maneuvering.

75 more elements

A method for prediction of, protection against, reduce the likelihood of and provide diagnostics of pellet-clad interaction failure of nuclear fuel rods during power maneuvering guidance of nuclear reactors using a computer program-comprising:

a. computer program algorithms comprising inputting into a computer a computer program implementing the algorithms for calculating the gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
,
(
4
)
and,
calculating the evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
,
(
3
)
where t is time, δ is the pellet-clad size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress for given fast neutron flux level and cumulative fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and,
where the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is the linear power in kW/m, pu is the fraction of unstable pores in the ceramic pellet material, and ps is the fraction of stable pores in the same material and,
where the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6) and:
additionally, inputting, into the computer, core instrumentation measurements, using said measurements to provide input to core monitoring software, accessing the monitoring software generated operational data of linear heat generation rate, neutron flux, burnup, fast neutron fluence, for all or the majority of the fuel rods at several axial locations and calculating, by the computer program, the gap size and clad stress when the gap is closed, and;
b. calculating a fuel conditioning state, Π, as function of time, t, by
a
1
⁢
ⅆ
Π
ⅆ
t
+
a
0
⁢
Π
=
f
⁡
(
σ
)
(
10
)
where the function ƒ is the clad inelastic strain rate driven by the time-dependent hoop stress, σ, while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, and;
c. allowing the linear power at a fuel rod segment, P, to vary during a reactor power maneuver with the constraints that it does not exceed the corresponding conditioning state plus a prescribed margin, ΔP, such that
P≦Π+ΔP (9), and:
thereafter, outputting, to an operators display, the stress response to a planned power increase, as operational guidance for a nuclear reactor operator for power maneuvering.

90 more elements

Dependencies are inferred from the claim text.

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Abstract

A method for prediction of a plurality of variables characterizing fuel performance, utilizing a reduced order model computer program, applied to nuclear reactor operations, and comprising:

a. loading a computer program into a computer;
inputting, into the computer, reactor core instrumentation measurements;
using said measurements to provide input to core monitoring software;
accessing the monitoring software generated operational data of linear heat generation rate, neutron flux, burnup, fast neutron fluence, for all or the majority of fuel rods at several axial locations;
calculating a gap size evolution from its initial as-manufactured value by
ⅆ
δ
ⅆ
t
=
f
⁡
(
σ
)
-
a
0
⁢
P
-
a
1
⁢
ⅆ
P
ⅆ
t
(
4
)
and,
b. calculating an evolution of clad hoop stress with time upon gap size reaching zero by
1
E
⁢
ⅆ
σ
ⅆ
t
+
f
⁡
(
σ
)
=
a
0
⁢
P
+
a
1
⁢
ⅆ
P
ⅆ
t
(
3
)
where t is time, δ is pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is linear heat generation rate in pellets of a fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power
and,
c. the parameter a0 is determined by
a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)
where B is the burnup in MWd/kgU, P is linear power in kW/m, pu is fraction of unstable pores in the ceramic pellet material, and ps is a fraction of stable pores in the same material;
d. the parameter a1 being generally dependent on burnup and linear power level is determined by
a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)
e. a clad creep rate function ƒ(σ), being generally dependent on stress and fast neutron flux, Φ, is determined by
ƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)
where a coefficient set cT, s, cIRR, v, w is specific to clad material and heat treatment;
f. displaying the calculated parameters, σ and/or δ, and
g. said displayed calculated parameters prompting nuclear reactor operators in taking operational steps for nuclear reactor operations.

71 more elements