Patent Number: 
Section: claims

1. 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 powerand,c. the parameter a0 is determined bya0=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 bya1=[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 byƒ(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, andg. said displayed calculated parameters prompting nuclear reactor operators in taking operational steps for nuclear reactor operations. 2. The method of claim 1 further comprising:a. nuclear reactor operations comprising power maneuvering of a nuclear reactor or designing a core loading pattern of a nuclear reactor or diagnosing of a suspected PCI cladding failure in a nuclear reactor. 3. The method of claim 2, further comprising:a. the parameters a0 and a1 and function ƒ(σ) are in tabulated form or fitted by curve fitting to a combination of experimental measurements of pellet and clad material properties and a result of large scale fuel performance codes. 4. The method of claim 3, when used for power maneuvering of a nuclear reactor, further comprising:a. tracking a maximum allowable linear power compared with a operating linear power in order to ensure that a safety margin to PCI fuel failure exists; projecting a safe power trajectory in order to plan maneuvers such as start up and control rod sequence exchange;b. dividing the projected safe power trajectory into a safe power jump followed by a continuous safe power ramp;c. predicting a safe power jump by determining the gap and hoop stress when a safe power trajectory projection is requested; calculating the safe power jump as the power corresponding to thermal expansion that consumes the present gap plus clad elastic strain corresponding to a given safe stress setpoint;d. predicting an instantaneous power ramp rate where the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, the power ramp rate being limited toR=(ƒ(σ*)−a0P)/a1  (12)e. predicting a continuous power ramp by limiting the power increase rate such that the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, wherewith the linear power is calculated from                                                        a              1                        ⁢                                          ⅆ                P                                            ⅆ                t                                              +                                    a              0                        ⁢            P                          =                  f          ⁡                      (                          σ              *                        )                                              (        11        )            which is integrated numerically by the reduced order model computer program whenever a safe power projection over a period of time is requested by the reactor operator;f. displaying the predicted safe power jump and power ramp for power maneuvering for nuclear reactor operators and;g. said predicted safe power jump and power ramp for prompting nuclear reactor operators in nuclear reactor operations. 5. The method of claim 3 further comprising:a. said calculated parameters are for use in core loading pattern design, power maneuvering of a nuclear reactor, or diagnostics of suspected PCI cladding failure in either a Boiling Water Reactor or a Pressurized Water Reactor;b. displaying the calculated parameters, σ and δ, for prompting of nuclear reactor operators in power maneuvering of a nuclear reactor. 6. The method of claim 3 using reduced order models for fast calculation of cladding stress in the majority of the fuel rods at different elevations of a reactor core for the purpose of calculating margin to and protecting against PCI failures comprising:a. 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;b. calculating, by the computer program, the gap size and clad stress when the gap is closed;c. outputting, to a display, the stress response to a planned power increase, and; said output display prompting core loading pattern design, operator power maneuvering or diagnostics of suspected PCI cladding failure for nuclear reactor operation. 7. The method of claim 6 further comprising:a. calculating the linear heat generation rate maximum allowable limit based on the calculated stress and outputting same for operational guidance;b. calculating the margin to reaching the maximum allowable linear heat generation rate and outputting same for operational guidance;c. calculating the margin in terms of total reactor thermal power to any rod segment reaching the maximum allowable linear heat generation rate and outputting same for design, diagnostics or operational guidance. 8. The method of claim 7 further comprising:a. the core instrumentation measurements and monitoring software are from a Boiling Water Reactor or from a Pressurized Water Reactor. 9. 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        )            andb. 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 powerand,c. the parameter a0 is determined bya0=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 bya1=[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 byƒ(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. 10. The method of claim 9 further comprising:a. the algorithms presented in the computer code and inputted into a computer are                                                                        1                E                            ⁢                                                ⅆ                  σ                                                  ⅆ                  t                                                      +                          f              ⁡                              (                σ                )                                              =                                                    a                0                            ⁢              P                        +                                          a                1                            ⁢                                                ⅆ                  P                                                  ⅆ                  t                                                                    ⁢                                  ⁢        and                            (        3        )                                                      ⅆ            δ                                ⅆ            t                          =                              f            ⁡                          (              σ              )                                -                                    a              0                        ⁢            P                    -                                    a              1                        ⁢                                          ⅆ                P                                            ⅆ                t                                                                        (        4        )            where,t is time, δ is the 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 the 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 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 initial gap size, δ0, is a manufacturing parameter, the coefficients a0 and a1 are not necessarily constant, and the creep rate function, ƒ, is not solely dependent on stress but also on the creep itself and the fast neutron fluence, and,b. nuclear reactor operations comprising power maneuvering of a nuclear reactor or designing a core loading pattern of a nuclear reactor or diagnosing of a suspected PCI cladding failure in a nuclear reactor. 11. The method of claim 10 further comprising:a. the parameter a0 is determined bya0=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;b. the coefficient, a1 is determined bya1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5  (6)and is generally dependent on burnup and linear power level;c. and the stress relaxation in the clad is directly linked to the inelastic strain rate given by the functionƒ(σ)=cT sin h(sσ)+cIRRσvΦw  (7)and is generally dependent on stress, σ, and neutron flux, Φ, with the coefficient set cT, s, cIRR, v, w being specific to clad material and heat treatment. 12. The method of claim 11 further comprising:a. the core instrumentation measurements and monitoring software are from a Boiling Water Reactor or from a Pressurized Water Reactor. 13. 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 powerande. the parameter a0 is determined bya0=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 bya1=[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 byƒ(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. 14. 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 bya0=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 bya1=[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. 15. 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        )            andcalculating 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 powerand,the parameter a0 is determined bya0=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 bya1=[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 byƒ(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. 16. 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 andthe parameter a0 is determined bya0=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 bya1=[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 byƒ(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. 17. The method of claim 16 further comprising:a. performing a stress survey and related functions offline using stored data obtained from reactor instrumentation and monitoring systems;b. performing the stress survey and related functions for the purpose of identifying a failed fuel assembly and further identifying the location of a failed fuel rod in the fuel assembly;c. performing the stress survey and related functions using design data for the purpose of optimizing the loading pattern of different fuel assemblies in a reactor core, and the associated planned control rod sequences for boiling water reactors, and reactivity management with soluble boron and control rod motion for pressurized water reactors;d. displaying the results of the stress survey for designers consideration and benefit to identify and distinguish design patterns with respect to their respective propensity to PCI related fuel rod failures. 18. 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 bya0=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 bya1=[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 thatP≦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. 19. A method of reactor maneuvering of claim 18 further comprising:a. obtaining stress driving the conditioning state calculation from a solution of the computer program system equations. 20. A method of reactor maneuvering of claim 19 further comprising:a. entering the computer program algorithms for calculating the conditioning state in a computer program connected to the reactor instrumentation and monitoring software;b. displaying the linear power margin, Π+ΔP−P, calculated by the program for guiding power maneuvering of a boiling water reactor. 21. A method of reactor maneuvering of claim 20 further comprising:a. displaying the core thermal power required for a plurality of fuel rod segments reaching the maximum linear power, Pmax=Π+ΔP, for guiding the power maneuvering of a pressurized water reactor.