ML17286A629

Proposed Tech Spec Changes,Modifying Safety Limits,Thermal Power,High Pressure & High Flow

ML17286A629
Person / Time
Site:	Columbia
Issue date:	02/27/1991
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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ML17286A627	List: ML17286A626 ML17286A628 ML17286A629
References
NUDOCS 9103070009
Download: ML17286A629 (26)

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~ I L ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES MODIFIED SAFETY LIMIT THERMAL POWER, HIGH PRESSURE AND HIGH FLOW 9103070009 910228 PDR ADOCK 05000397 P PDR

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SUMMARY

JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES TECHNICAL SPECIFICATION NO. PAGE NO. JUSTIFICATION 2.1.2 2-1 New safety limit values to reflect cycle specific safety analysis, employing new methodology.

B2.0'2-1 Editorial reflect change to bases change to safety to limit 2. 1.2 discussed above.

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2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25" of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than lOX of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and ".

ACTION:

With THERMAL POWER exceeding 25K of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10~

of rated flow, be in at the requirements of Speci l~t HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with ion 6.7. 1.

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2.1.2 The MINIMUM CRITICAL POWE 1;0 (MCPR) shall not be less than with two recirculation loop operation ll not be less than with single recirculation loop operation withahois th ctor vessel steam dome pressure greater than 785 psig and core flow greater t 10~ of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 2.

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+hocTA Al500 ITAL o TTn~ +o Roc, AQp MCP ess than ws two recirculation loop operation or less than with single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than lOX of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

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2. 1. 3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6. 7. 1.

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CONTROLlED COPY 2.0 SAFET IMITS and LIMITIHG SAFETY SYSTEM SETTINGS gP To 4.Sg() lQUJ 9 //T'~ CqC.L,F EY, POSVRE, &No C.PCLE EgPesuQ,E, Cq&aATSR, TA,AH 4600 RAUCID/Mlle To EQC BASES INTRODUCTIOH The fuel cladding, reac'tor pressure vessel and primary system piping are the principal barriers to the release of .radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations. and anticipated. transients. The fuel cladding,ntegrity Safety Limit is set such that no fuel damage is calculated to occur if'the limit is not violated. Because fuel damage is not directly observable, a step-back a roach is used to establish a Safety Limit such that the MCPR is not les a 3RK- for two recirculation loop operation and or s ng e recircula-tion loo operation for all nuclear fuel in WNP-2. MCPR greater than . for wo recsrcu a ion oop operation an for single recirculation loop opera-tion represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuous'ly measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reac-tor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold be'yond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding integrity Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity safety limit assures that during normal operation and during anticipated operational occur Oo) rences, at least 99.9 percent of the fu rods 1n the core do not ex err transition boiling (

Reference:

XH-HF-52 A), ev. ; B om epor UK90-126; GEll Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project Ho. 2, Reload 5, Cycle 6). The latter two references support application of the above establishe fety limit to GE11 and SVEA-9 LFA fuel in WNP-2 SAFETY IMITS )sop VP To 4 80 VJD/lAVQ C'/O.E. EK SOPH.

l. I'8 FOR. CQCtic K)cppbIskE CqLs~Tgg. ~psi 4ropp mush J lo 2.1.1. THERMAL POWER Low r re or Low O.lO ~ F S For certain conditions of pressure and flow, the correlation is not valid for all critical power calculations The correlation is not valid pe for bundle mass velocities less than . x 10 lbs/hr-ft or pressures less goQ% than . Therefore, the fuel cladding integrity Safety Limit is estab-

)she by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10~ lbs/h (approximately a mass velocity of XS x 10s lbs/hr-fte), bundle pressure drop is nearly independent ofpundle power WASHIHGTOH NUCLEAR - UNIT 2 B 2-1 Amendment No. 84

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SAFETY LIMITS BASES THERMAL POWER Low Pressure or Low Flow (Continued) and has a value of 3.5 psi. Thus, the bundle flow'with a 4.5 psi driving head will be greater than 28 x 10'bs/h. Full scale ATLAS test data taken at pres-sures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50K of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below QS-@sic is. conservative.

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'2. 1.2 THERMAL POWER Hi h Pressure and Hi h Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated 'to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor opera-tion, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, tha critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9X of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR .is determinyg)using the AHF Critical Power, Methodology for boiling water reactors'hich is a statistical model that combines all.of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence;,of boil-ing transition is determined using the ANF nuclear critical heat fluxenthalpy pgF9 ~correlation. Thethe correlation is valid over the range of conditions used in the tests of da$ a used to deveIop the correlation.

The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1.

The bases for the uncertainties in the core parameters AQUA are given in XN-HF-524(A), Rev.

2a g and the basip ford the uncertainty in the phlF 9 correla-pgp ~gaycpgcr g +~A Ks rrcepdeuw Qg > ~

tion is given in . The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.

a. Y Nuclear"Critical Power Methodology for Boiling Mater Reactors, XN-HF-524(A), Rev. ~g ~ ~

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BASES TABLE B2.1.2"1 UNCERTAINTIES CONSIDERED IN THE MCPR SAFETY LIMIT STANDARD Parameter DEVIATION" Feedwater Flow Rate .0176 Feedwater Temperature .0076 Core Pressure .0050 Total Core Flow Rate .0250

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WASHINGTON NUCLEAR " UNIT 2 B 2-3

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Amendment No. 28

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ATTACHMENT II ANF-91-01 Page A-1 APPENDIX A MCPR SAFETY LIMIT A.1 INTRODUCTION Bundle power limits in a boiling water reactor (BWR) are determined through evaluation of critical heat flux phenomena. The basic criterion used in establishing critical power ratio (CPR) limits is that at least 99.9% of the fuel rods in the core will be expected to avoid boiling transition (critical heat flux) during normal operation and anticipated operational occurrences. Operating margins are defined by establishing a minimum margin to the onset of boiling transition condition for steady state operation and calculating a transient effects allowance, thereby assuring that the steady state limit is protected during anticipated off-normal conditions. This appendix addresses the calculation of the minimum margin to the steady state boiling transition condition, which is implemented as the MCPR safety limit in the plant technical specifications. The transient effects allowance, or the limiting transient change in CPR (i.e., hCPR), is treated in the body of this report.

The MCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associated with the thermal hydraulic state of the reactor using design basis radial, axial, and local power distributions and considering fuel assembly channel bow.

Reference 5 describes ANF MCPR safety limit methodology and the incorporation of channel bow effects. Some of the calculational uncertainties, incfuding those introduced by the critical power correlation, power peaking, and core coolant distribution, are fuel related. When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the MCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core. Similarly, when an ANF-fabricated reload batch is used to replace a group of dissimilar fuel assemblies, the core average fuel dependent parameters change because of the difference in the relative number of each type of bundle in the core, and the MCPR safety limit is again reevaluated.

ANF-91-01 Page A-2 The design basis power distribution is made up of components corresponding to representative radial, axial, and local peaking factors. Where such data are appropriately available from the previous cycle, these factors are determined through examination of operating data for the previous cycle and predictions of operating conditions during the cycle being evaluated for the MCPR safety limit, Available operating data for WNP-2 and the predicted operating conditions for Cycle 7 were evaluated to identify the design basis power distributions for use in the Cycle 7 MCPR safety limit analysis. A high neutron flux trip of 120% was used in the safety limit analysis.

ANF-91-01 Page A-2 The design basis power distribution is made up of components corresponding to representative radial, axial, and local peaking factors. Where such data are appropriately available from the previous cycle, these factors are determined through examination of operating data for the previous cycle and predictions of operating conditions during the cycle being evaluated for the MCPR safety limit. Available operating data for WNP-2 and the predicted operating conditions for Cycle 7 were evaluated to identify the design basis power distributions for use in the Cycle 7 MCPR safety limit analysis. A high neutron flux trip of 120% was used in the safety limit analysis.

ANF-91-01 Page A-3 A.2 ASSUMPTiONS A,2.1 Desi n Basis Power Distribution The local and radial power distributions which were determined to be conservative for use in the safety limit analysis are shown in Figures A.1 through A.4.

A.2.2 H draulic Demand Curve Hydraulic demand curves based on calculations with XCOBRA were used in the safety limit analysis. The XCOBRA calculation is described in ANF topical reports XN-NF-79-59(A),

"Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," and ANF-1125(P)(A),

"ANFB critical Power Correlation."

A.2.3 S stem Uncertainties System measurement uncertainties are not fuel dependent. The values reported by the NSSS supplier for these parameters remain valid for the insertion of ANF fuel. 7he values used in the safety limit analysis are tabulated in the topical report XN-NF-524(P)(A), Revision 2 and Supplements, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."

A.2.4 Fuel Related Uncertainties Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty. The values used in the safety limit analysis are also tabulated in the topical report XN-NF-524(P)(A), Revision 2 and Supplements, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."

ANF-91-01 Page A-4 A.3 SAFETY LIMITCALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(P) (A), Revision 2 and Supplements, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." With 500 Monte Carlo trials it was determined that for a minimum CPR value of 1.07 at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95% for the design basis power distributions from BOC to a cycle average burnup of 4500 MWd/MTU.

Similarly for the design basis power distributions from 4500 MWd/MTU to EOC a minimum CPR value of 1.11 is required to provide the same protection.

The effects of channel bow are included in the 1.07 and 1 ~ 11 safety limit MCPR values.

Without channel bow, the SLMCPRs would have been reduced by about 0.03. The Supply System has reused some initial core channels on ANF 8x8 fuel assemblies in the WNP-2 Cycle 7 core. A maximum extent of channel bow including, the effects of reused channels adjacent to assemblies with exposed channels was input for the 8x8 fuel. The 9x9-9X fuel input was based on a maximum extent of channel bow assuming a new channel on the 9x9-9X fuel adjacent to assemblies with exposed channels. The core results including input for both fuel types show that the ANF 8x8 assemblies having reused channels in the WNP-2 Cycle 7 core have sufficient MCPR margin that they do not contribute significantly to the number of rods in boiling transition even with the effects of increased channel bow due to reused channels. The maximum end of cycle exposure of the reused channels corresponds to a bundle average burnup of 48,000 MWd/MTU which is within the ANF channel bow data base to approximately 50,000 Mwd/MTU.

ANF-91-01 Page A-4 A,3 SAFETY UMIT CALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(P)(A), Revision 2 and Supplements, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." With 500 Monte Carlo trials it was determined that for a minimum CPR value of 1.07 at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95% for the design basis power distributions from BOC to a cycle average burnup of,4500 MWd/MTU.

Similarly for the design basis power distributions from 4500 MWd/MTU to EOC a minimum CPR value of 1.11 is required to provide the same protection.

The effects of.channel bow are included in the 1.07 and" 1.11 safety limit MCPR values.

Without channel bow, the SLMCPRs would have been reduced by about 0.03. The Supply System has reused some initial core channels on ANF 8x8 fuel assemblies in the WNP-2 Cycle 7 core. A maximum extent of channel bow including, the effects of reused channels adjacent to assemblies with exposed channels was input for the Gx8 fuel. The Gx9-9X fuel input was based on a maximum extent of channel bow assuming a new channel on the 9x9-9X fuel adjacent to assemblies with exposed channels. The core results including input for both fuel types show that the ANF 8x8 assemblies having reused channels in the WNP-2 Cycle 7 core have sufficient MCPR margin that they do not contribute significantly to the number of rods in boiling transition even with the effects of increased channel bow due to reused channels. The maximum end of cycle exposure of the revs'ed channels corresponds to a bundle average burnup of 48,000 MWd/MTU which is within the ANF channel bow data base to approximately 50,000 Mwd/MTU.

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ANF-91-01 Page A-5 1.026 1.063 1.062 1.023 1.125 1.012 1.036 1.020 0.966 1.063 0.959 1.043 1.021 0.870 1.009 1.016 0.919 0.996 1.062 1.043 1.021 1.080 1.109 1.067 0.995 0.996 0.991 1.023 1.021 1.080 0.000 0.000 0.000 1.052 0.976 0.954 1.125 0.870 1.109 0.000 0.000 0.000 1.081 0.834 1.048 1.012 1.009 . 1.067 0.000 0.000 0.000 1.042 0.967 0.944 1.036 1.016 0.995 1.052 1.081 1.042 0.975 0.875 0.968 1.020 0.919 0.996 0.976 0.834 0.967 0.875 0,884 0.955 0.966 0.996 0.991 0.954 1.048 0.944 0.968 0.955 0.907 FIGURE A.1 WNP-2 CYCLE 7 SAFETY LIMITLOCAL PEAKING FACTORS (ANF 9X9-9X FUEL WITH CHANNEL BOW)

ANF-91-01 Page A-6 1.047 1.054 1.096 1.081 1.068 1.085 1.100 1.075 1.054 1.045 0.940 1.060 1.042 0.934 0.950 1.060 1.096 0.940 1.037 1.009 0.977 0.979 0.906 1.016 1.081 1.060 . 1.009 0.000 0.883 0.953 0.987 0.978 1.068 1.042 0.977 0.883 0.000 0.963 0.982 0.971 1.085 0.934 0.979 0.953 0.963 0.967 0.855 0.991 1.100 0.950 0.906 0.987 0.982 0.855 0.906 1.007 1.075 1.060 1.016 0.978 0.971 0.991 1.007 '.993 FIGURE A.2 WNP-2 CYCLE 7 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF 8X8 FUEL WITH REUSED CHANNEL BOW)

ANF-91-01 Page A-6 1.047 1.054 1.096 1.081 1.068 1.085 1.100 1-.075 1.054 1.045 0.940 1.060 1.042 0.934 0.950 1.060 1.096 0.940 1.037 1.009 0.977 0.979 0.906 1.016 1.081 1.060 1.009 0.000 0;883 0.953 0.987 0.978 1.068 1.042 0.977 0.883 0.000 0.963 0.982 0.971 1.085 0.934 0.979 0.953 0.963 0.967 0.855 0.991 1.100 0.950 0.906 0.987 0.982 0.855 0;906 1.007 1.075 1.060 1.016 0.978 0.971 0.991 1.007 '.993 FIGURE A.2 WNP-2 CYCLE 7 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF 8X8 FUEL WITH REUSED CHANNEL BOW)

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. 0.2 0.4 06 08 1 1.2 1.6 RFlDIAL PONER PEFlKING FIGURE A.3 RADIAL POWER HISTOGRAM FOR FULL CORE SAFETY LIMIT MODEL BOL TO 4500 MWd/MTU CYCLE AVERAGE BURNUP

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CJ 75 50 25 0.2 0.4 0.6 0.8 1.2 1.6 RAOIAL POHER PEAKING DZ FIGURE A.4 RADIAL POWER HISTOGRAM FOR FULL CORE SAFETY LIMIT MODEL (0

4500 MWd/MTU CYCLE AVERAGE BURNUP TO EOC

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0 0.2 0.0 0.6 0.8 1.2 1.6 RADIAL POWER PEAKING FIGURE AA RADIAL POWER'HISTOGRAM FOR FULL CORE SAFETY LIMITMODEL 4500 MWd/MTU CYCLE AVERAGE BURNUP TO EOC