ML050320301
ML050320301 | |
Person / Time | |
---|---|
Site: | San Onofre ![]() |
Issue date: | 11/30/2004 |
From: | Laakso V, Paggen V Westinghouse |
To: | Office of Nuclear Reactor Regulation |
References | |
WCAP-16167-NP, Rev 00 | |
Download: ML050320301 (72) | |
Text
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,1 ENCLOSURE 5 San Onofre Unit 3 Pressure and Temperature Limits Report (PTLR)
Westinghouse Non-Proprietary Class 3 WCAP-16167-NP Rev 00 San Onofre Nuclear Generating Station Unit 3 November 2004 RCS Pressure and Temperature Limits Report PI 12J et Wi C
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Supplier Status Stamp Retereee: S0123-)OWMS.26 W23-723-W q7 itevY O Manufactuning merproce ed: O Yes O No
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so23 -92.s. r-1-7 Ize%'/ 0 Westinghouse Electric Company LLC 20 International Drive. P.O. Box 500 Windsor, Connecticut 06095-0500 S023-923-M97
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16167-NP, Rev 00 San Onofre Nuclear Generating Station Unit 3 RCS Pressure and Temperature Limits Report November 2004 Author V. Laakso V. Paggen Technical:
S. Byrne F. Ferraraccio Approved:
W. Turkowski oa23 q23s.Mqz Rteq/O PIM O 2004 Westinghouse Electric Company LLC 20 International Drive Windsor, Connecticut 06095 All Rights Reserved S023-923-M97
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TABLE OF CONTENTS LIST OF TABLES...............................................
iii LIST OF FIGURES..............................................
iii LIST OF ACRONYMS...............
iv ABSTRACT v
INTRODUCTION...............
1-1 1.0 NEUTRON FLUENCE VALUES 1-1 1.1 INPUT DATA................
1-2 1.1.1 Materials and Geometry.
1-2 1.1.2 Cross Sections...............
1-3 1.2 CORE NEUTRON SOURCE 1-3 1.2.1 Synthesized Three Dimensional Results..............
1-3 1.2.2 Calculated Activities and Measured Activities................
1-4 1.2.3 Measurement / Calculation Ratios..............
1-5 1.3. FLUENCE CALCLATION..................
1-5 1.A METHODOLOGY QUALIFICATION AND UNCERTAINTY ESTIMATES 1-6 1.4.1 Analytic UncertaintyAnalysis............................................................
1-6 1.4.2 Comparison with Benchmark and Plant Specific Measurements.
..... 1-8 1.4.3 Overall Bias and Uncertainty............................................................
1-8 2.0 REACTOR VESSEL SURVEILLANCE PROGRAM.............................................................. 2-1 2.1 TEST MATERIAL SELECTION............................................................
2-1 2.1.1 Plate Material Selection................................
2-1 2.1.2 Weld Material Selection................................
2-2 2.2 TEST SPECIMENS.................................
2-3 2.2.1 Type and Quantity................................
2-3 2.2.2 Unirradiated Specimens................................
2-3 2.2.3 Irradiated Specimens.
2-4 2.3 SPECIMEN IRRADIATION................................
2-4 2.3.1 Encapsulation of Specimens..................................
2-4 2.3.2 Flux and Temperature Measurement
.............................. 2-5 2.3.3 Irradiation Locations................................
2-6 2.3.4 Capsule Assembly Removal................................
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TABLE OF CONTENTS 3.0 LTOP SYSTEM LIMITS........................................................
3-1 3.1 LOWTEMPERATURE OVERPRESSURE PROTECTION SYSTEM................................ 3-1 3.2 BASIS FOR LTOP SYSTEM LIMITS 3-1 3.2.1 PeakTransient Pressures........................................................
3-2 3.2.2 Applicable P-T Limits........................................................
3-5 4.0 BELTLINB MATERIALADJUSTED REFERENCE TEMPERATURE
........................... 4-1
4.1 BACKGROUND
4-1 4.2 RESULTS..............................................................................................................................4-1 4.2.1 Assumptions.....................
4-2 4.2.2 Fluence Calculation.....................
4-2 4.2.3 Chemistry Factor Calculation 4-2 4.2.4 Calculation of ART for the Limiting Plates at 1/4T and 3/4T..................................... 4-3 4.3 ANALYSIS DETAILS.................................................
4-3 4.3.1 Selection of Representative and Limiting Cases....................................
4-3 4.3.2 Wetted Surface Fluence Calculation....................................
4-3 4.4 CALCULATION OF THE ADJUSTED REFERENCE TEMPERATURES AND SELECTION OF THE MAXIMA....................................
4-4 5.0 PRESSURE-TEMPERATURE LIMITS USING LIMITING ADJUSTED REFERENCE TEMPERATURE IN THE P-T CURVE CALCULATION.................................... 5-1 5.1 RCS TEMPERATURE RATE-OF-CHANGE LIMITS........................................
5-1 5.2 RCS PRESSURE-TEMPERATURE LIMITS 5-1 6.0 MINIMUM TEMPERATURE REQUIREMENTS IN THE PRESSURE-TEMPERATURE CURVES.................................
6-1 7.0 APPLICATION OF SURVEILLANCE DATA TO ADJUSTED REFERENCE TEMPERAATURE CALCULATIONS.......................................
7-1
8.0 REFERENCES
8-1 APPENDIX A TECHNICAL SPECIFICATION REFERENCES TO THE PTLL............................................
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LIST OF TABLES 1-1 Calculated Wetted Surface Neutron Fluence 1-5 1-2 Fluence and Fluence Factors at 1/4T and 314T........................................................
1-6 1-3 Neutron Fluence Uncertainty.....................................................
1-8 2-1 Type and Quantity of Specimens for Irradiation Exposure and Irradiated Tests.
2-8 2-2 SONGS Unit 3 Capsule Assembly Removal Schedule.....................................................
2-8 3-1 Low Temperature RCS Overpressure Protection Range 3-1 4-1 Calculated Adjusted Reference Temperatures at 20 and 32 EFPY.
4-1 4-2 Calculated Wetted Surface Fluence.....................................................
4-3 4-3 Fluence and Fluence Factors at l/4T and 314T.....................................................
4-4 4-4 Calculated Adjusted Reference Temperatures at 1/4T and 3/4T.
4-5 5-1 SONGS Unit 3 Heatup at 60'F/hr, RCS Pressure-Temperature Limits through 32 EFPY.5-2 5-2 SONGS Unit 3 Cooldown via Control Room Instrumentation, RCS Pressure-Temperature Limits through 32 EFPY.5-3 5-3 SONGS Unit 3 Cooldown via Remote Shutdown Panel Instrumentation, RCS Pressure-Temperature Limits through 32 EFPY....................................................
5-4 6-1 Minimum Temperature Requirements for SONGS Unit 3 through 32 EFPY.............................. 6-1 6-2 Limiting Indicated Minimum Pressure Requirements for SONGS Unit 3 through 32 EFPY....................................................
6-2 7-1 Calculation of Chernistry Factors for Surveillance Plate and Weld Material.................
............. 7-2 7-2 Credibility Test for Surveillance Plate and Weld Material...................................................
7-2 7-3 Credibility Test for Standard Reference Material....................................................
7-2 A-i Low Temperature RCS Overpressure Protection Range...
................................................... A-9 LIST OF FIGURES 1-1 Fluence Analysis Methodologies for SONGS Unit 3 Surveillance Capsule................................ 1-9 5-1 SONGS Unit 3 RCS Heatup Pressure-Temperature Limits through 32 EFPY-Normal Operation..
5-6 5-2 SONGS Unit 3 RCS Cooldown Pressure-Temperature Limits through 32 EFPY-Normal Operation.5-7 5-3 SONGS Unit 3 RCS Cooldown Pressure-Temperature Limits through 32 EFPY - Remote Shutdown Panel Operation........................................................
5-8 A-1 SONGS Unit 3 RCS Heatup P-T Limits through 32 EFPY - Normal Operation.............
........... A-5 A-2 SONGS Unit 3 RCS Cooldown P-T Limits through 32 EFPY - Normal Operation................... A-6 A-3 SONGS Unit 3 RCS Cooldown P-T Limits through 32 EFPY - Remote Shutdown Panel Operation........................................................... A-7 s93 23-qz3.-
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LIST OF ACRONYMS ART Adjusted Reference Temperature CF Chemistry Factor CFR Code of Federal Regulations EFPY Effective Full Power Years EOC End of Cycle EOL End of Life ff Fluence factor HAZ Heat Affected Zone HPSI High Pressure Safety Injection LCO (Technical Specification) Limiting Condition for Operation LTOP Low-Temperature Overpressure Protection MeV Million electron volts NDTT Nil Ductility Transition Temperature NRC Nuclear Regulatory Commission NSF Non-saturation correction factor P-T Pressure-Temperature PHTP Pre-service Hydrostatic Test Pressure PTLR Pressure Temperature Limits Report PTS Pressurized Thermal Shock RCP Reactor Coolant Pump RCS Reactor Coolant System RTN=p Reference Temperature perASMIE Code NB2300 RTprs Pressurized Thermal Shock Reference Temperature SCE Southern California Edison SDCS Shutdown Cooling Systcm SONGS San Onofre Nuclear Generating Station UFSAR Updated Final Safety Analysis Report SO23_L 23_M'7 go ocrxt WCAP-16167-NP, Rev 00 Page iv November 2004 S023-923-M97
ABSTRACT During the development of the improved standard technical specifications, the NRC staff agreed with the industry that the reactor coolant system Pressure-Temperature (P-T) and Low Temperature Overpressure Protection (LTOP) system curves and setpoints may be voluntarily relocated outside the technical specifications to a licensee-controlled document. This change, promulgated in Generic Letter 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, permits the licensee to maintain these limits efficiently and at a lower cost, provided that the parameters for constructing the curves and setpoints are derived using a methodology approved by the NRC.
Pressure and temperature limits for heatup and cooldown of the San Onofre Nuclear Generating Station (SONGS) Unit 3 reactor coolant system through 32 effective full power years (EFPY) of operation are developed in this report. These heatup and cooldown limits are designed to prevent potential brittle fracture of the reactor pressure vessel during the most restrictive low temperature overpressure event. SONGS Unit 3 Technical Specifications affected by Pressure-Temperature or LTOP limits are discussed in Sections 3 and 5 of this report. A summary of the SONGS Unit 3 Technical Specifications changes is shown in Appendix A.
The methodology in this document is applicable to both SONGS Units 2 & 3. However, only the Pressure and Temperature Limits affecting Unit 3 and corresponding Technical Specification information is contained in this report. The SONGS LTOP methodology described in this report is conservative relative to the current methods approved by the NRC and encompass the expected operating conditions for SONGS Unit 3 through 32 EFPY.
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INTRODUCTION This document contains the information needed to develop the reactor coolant system Pressure-Temperature (P-T) limit curves and low temperature overpressure protection (LTOP) set point values and curves for San Onofre Nuclear Generating Station (SONGS) Unit 3, extend these P-T limits through 32 effective full-power years (EFPY) of operation, and support the relocation of these P-T limits from the Technical Specifications into this PTLR. Methodology used in this report was approved by the NRC in CE NPSD-683-A, Rev 6, Reference 1, which in turn is based on the guidance contained in Generic Letter 96-03, Reference 2. Additional guidance on the information needed in the SONGS-specific PTLR in order to satisfy Generic Letter 96-03 criteria is contained in the Safety Evaluation issued for CE NPSD-683-A, Rev 6.
Consistent with Generic Letter 96-03, Sections I through 7 of this report develop the P-T limits, establish LTOP setpoints, calculate the Adjusted Reference Temperature, develop a reactor vessel surveillance program, and calculate the neutron fluence to support the SONGS Unit 3 PTLR. The methodology used is compatible with that currently approved by the NRC and is presently used in the design bases for SONGS Units 2 & 3. This PTLR incorporates the currently approved SONGS methods, clarifies the differences, and justifies the changes relative to the NRC-approved PTLR report, Reference 1.
The methodology in this document is applicable to both SONGS Units 2 and 3. However, only the P-T limits affecting Unit 3 and corresponding Technical Specification information are contained in this report The peak adjusted reference temperatures and P-T limit curves in this report are valid for SONGS Unit 3 through 32 EFPY.
1.0 NEUTRON FLUENCE VALUES The reactor vessel beltine neutron fluence has been calculated for the critical locations in accordance with Reference 3. The following discussion gives the results of the fluence calculation followed by the details of the calculational analysis for SONGS Unit 3.
The peak value of neutron fluence (E>l MeV) at the vessel wetted surface projected to 32 effective full power years (EFPY) is 4.1907 x 1019 neutrons per square centimeter (n/cm2) and corresponds to the intermediate shell plates. This value is used as input to the adjusted reference temperature (ART) calculations for SONGS Unit 3. The peak fluence for the lower shell after 32 EFPY is 4.0545 x 10'9 n/cm2. The fluence values have an associated two-sigma (2a) uncertainty of +22.8%, Reference 4.
SONGS Unit 3 capsule number W263 was located at 7.0 degrees off the major axis, References 5 and 6, for Cycles 1 through 11. The core power distribution during these eleven irradiation cycles was S023-q23-M 01l7 Rev O oPg c
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symmetric in both azimuthal and axial direction, i.e., the axial power shape was roughly the same for any azimuthal angle, and the azimuthal power shape was the same for any height. This means that the neutron flux at some point (R, 0, Z) can be considered to be a separable function of (R, 0) and (R, Z). Therefore, irradiation for Cycles 1 through 11 was modeled using the standard synthesis procedures of Reference 7.
Figure 1-1, Reference 4, depicts the analytical process that is used to determine the fluence accumulated over each irradiation period. As shown in this figure, the analysis is divided into seven tasks: (1) generation of the neutron source, (2) development of the DORT geometry models, (3) calculation of the macroscopic material cross sections, (4) synthesis of the results, and (5-7) estimation of the calculational bias, the calculational uncertainty, and the final fluence. Additional detail for these tasks is provided in the following sections.
1.1 INPUT DATA 1.1.1 Materials and Geometry The time-averaged space and energy-dependent neutron sources for Cycles 1-11 were calculated using the SORREL code, Reference 8. The effects of burnup on the spatial distribution of the neutron source were accounted for by calculating the cycle average fission spectrum for each fissile isotope on an assembly-by-assembly basis, and by determining the cycle-average specific neutron emission rate. This data was then used with the normalized time weighted average pin-by-pin relative power density distribution to determine the space and energy-dependent neutron source. The azimuthally averaged, time averaged axial power shape in the peripheral assemblies was used with the fission spectrum of the peripheral assemblies to determine the neutron source for the axial DORT run. These two neutron source distributions were input to DORT as indicated in Figure 1-1. Three separate sources (Cycles 1-8a, 8b-lOa, and 1Ob-l 1) were developed in order to account for changes in reactor coolant inlet temperature that occurred during Cycles 8 and 10. A power uprate that occurred in Cycle II was also accounted for in the synthesis procedure.
The system geometry models for the mid-plane (R, 0) DORT were developed using standard interval size and configuration guidelines. The (R, 0) model for the Cycles 1-8a, 8b-lOa, and lOb-l I analysis extended radially from the center of the core to the outer surface of the pressure vessel, and azimuthally from the major axis to 45°. The axial model extended from 35 cm below the active core region to 35 cm above the active core region. This geometric model either meets or exceeds all guidance criteria concerning interval size that are provided in US Regulatory Guide 1.190, Reference 3. Cold dimensions were used in all cases. The geometry models were input to the DORT code as indicated in Figure 1-1.
These models can be used for all subsequent fluence analyses for SONGS Unit 3.
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1.1.2 Cross Sections In accordance with Regulatory Guide 1.190, cross sections for SONGS Unit 3 were obtained from the BUGLE-93 cross section library, Reference 9. The GIP code, Reference 10, was used to calculate the macroscopic energy-dependent cross sections for all materials used in the analysis, i.e., from the core out through the pressure vessel and from core plate to core plate. The ENDFAB-VI dosimeter reaction cross sections were used to generate the response functions that were used for the DORT-calculated "saturated" specific activities.
1.2 CORE NEUTRON SOURCE The primary tool used in the determination of the flux and fluence exposure to the surveillance capsule dosimeters is the two-dimensional discrete ordinates transport code DORT, Reference 11.
The cross sections, geometry, and appropriate source were combined to create a set of DORT models (r, 6 and r, z) for the Cycles 1-8a, 8b-lOa, and lOb-1I analyses. Each DORT run utilized a cross section Legendre expansion of three (p3), seventy directions (sjo), with the appropriate boundary conditions. The r, z models used a cross section Legendre expansion of three (p3), seventy directions (slo), with the appropriate boundary conditions. A theta-weighted flux extrapolation model was used, and all other requirements of Regulatory Guide 1.190 that relate to the various DORT parameters, were either met or exceeded for all DORT runs.
1.2.1 Synthesized Three Dimensional Results DORT analyses produce two sets of two-dimensional flux distributions; one for a vertical cylinder and one for the radial plane for each set of dosimetry. The vertical cylinder, referred to as the R, Z plane, is defined as the plane bounded 35 cm above and 35 cm below the active core region, and radially by the center of the core and the outside surface of the reactor pressure vessel. The horizontal plane, referred to as the R, 0 plane, is defined as the radial plane bounded by the center of the core and the outside surface of the pressure vessel, and azimuthally by the major axis and the adjacent 45 degree radius. The vessel flux, however, varies significantly in all three cylindrical-coordinate directions (R, O, Z). This means that if a point of interest is outside the boundaries of both R, Z DORT and R, B DORT, then the true flux cannot be determined from either DORT run. Under the assumption that the three-dimensional flux is a separable function (Reference 7), both two-dimensional data sets were mathematically combined to estimate the flux at all three-dimensional points (R, 0, Z) of interest. The basis used for the flux-synthesis process is identical to the procedure outlined in Regulatory Guide 1.190.
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1.2.2 Calculated Activities and Measured Activities The calculated activities for each dosimeter type "d" for each irradiation period were determined using the following equation:
C,,
(r,)xRFd xBd xNSF (1-1) where:
Cd calculated specific activity for dosimeter "d" in piCi of product isotope per gram of target isotope Ip (rd) - three dimensional flux for dosimeter "d" at position rd for energy group "g"
= dosimeter response function for dosimeter "d" and energy group "g" Bd = bias correction factors for dosimeter "d" NSF = non-saturation correction factor Three separate sets of activities were calculated for this analysis, with a combination of the calculated activities performed using Equation 1-2. The end of Cycle 11 (EOC 11) total calculated activity, Cd II.
was computed using Equation 1-1 for dosimeter "d" as:
(Cd(cycles 1-1 )-
(Cd(cycles 1-8a)) + (Cd(cycle Bb-lOa)) + (Cd(cycle lOb-1l))
(1-2)
Each activity in Equation 1-2, Cd(cycle), is calculated using Equation 1-1, however each cycle-specific set of data, i.e., 1 -8a, 8b-I Oa, and I Ob-l 1, is calculated using a cycle-specific NSFcycie factor.
The bias correction factors, Bd, used in the above specific activity calculation depends on the type of dosimeter. A short half life bias is used for an activation type dosimeter whereas either a photo-fission or an impurities bias is employed for a fission type dosimeter. Aphoto-fission factor was applied to correct for the fact that some of the cesium-137 atoms present in the dosimeter were produced by 6'. f) reactions and were not accounted for in DORT analysis. Likewise, an impurity factor was included to account for U-235 content in the U-238 dosimetry. The short half life bias was insignificant and therefore was not applied.
1.2.3 Measurement / Calculation Ratios The following explanations define the meanings of the terms "measurements" and "calculations" as used in this analysis, Reference 7.
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- Measurements:
The term "measurements" as used here means the measurement of the physical quantity of the dosimeter (specific activity) that responded to the neutron fluence, not to the "measured fluence." For example, reference to an iron dosimeter measurement means the specific activity of 54Mn in pCi /g, which is the product isotope of the dosimeter reaction:
s'Fe + 'on -4 4Mn + p Calculations:
The calculational methodology produces two primary results, the calculated dosimeter activities and the neutron flux at all points of interest. The term "calculation" as used here means the calculated dosimeter activity. The calculated activities are determined in such a way that they are directly comparable to the measurement values, but without recourse to the measurements. That is, the calculated values determined by DORT are directly comparable to the measurement values.
ENDF/B-VI based dosimeter reaction cross sections, Reference 12, and response functions were used in determining the calculated values for each individual dosimeter. In summary, it should be stressed that the calculation values in this approach, Reference 7, are independent of the measurement values.
1.3 FLUENCE CALCULATION The following values were obtained from Reference 19:
End of Cycle 11 (EOC 11) = 14.925 EFPY Wetted surface cumulative fluence, as shown in Table 1-1.
The fluence values for 20 EFPY shown in Table 1-1 were calculated by linear interpolation, Reference 19. Fluence values for any time between 14.925 and 32 EFPY can be calculated by linear interpolation.
Table 1-1 Calculated Wetted Surface Neutron Fluence (in 1019 n/cm2)
Location 14.925 EFPY 20 EFPY 32 EFPY Intermediate shell 2.0111 2.659 4.1907 (Plates C-6802-1, 2, 3)
Lower shell 1.9496 2.575 4.0545 (Plates C-6802-4, 5, 6)
Reference 13 gives the following equation for the attenuation of fluence with distance into the plate:
f =f. (e014x)
(1-3)
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where: f
= fluence at the desired location, f,,rf = fluence at the wetted surface of the vessel, x
- distance, in inches, from the wetted surface of the vessel to the desired location.
For each location (intermediate and lower shells), the fluence at 1/4T and 314T were calculated by using Equation (1-3), with the following values of "x" from Reference 19:
x = 2.375 in. for 1/4T, x = 6.6875 in. for 3/4T.
The fluence factors were calculated by using Equation (1-4) from Reference 13:
ff f(0 28- 0.10 log f)
(1-4)
Table 1-2 Fluence (10"nlcm ) and Fluence Factors at 1/4T and 314T Location EFPY 1/4T f*
1/4T ff**
3/4T f*
314T ff**
Intermediate Shell 20 1.5037 1.1129 0.5342 0.8248 (Plates C-6802-1, 2, 3) 32 2.3699 1.2328 0.8419 0.9517 Lower Shell 20 1.4562 1.1042 0.5173 0.8159 (Plates C-6802-4, 5, 6) 32 2.2929 1.2244 0.8145 0.9424
- f -neutron fluence
- ff.-.fluence factor per Equation 1-4 1.4 METHODOLOGY QUALIFICATION AND UNCERTAINTY ESTIMATES The SONGS Unit 3 Cycles I through 11 fluence predictions were based on the methodology described in the Framnatome ANP "Fluence and Uncertainty Methodologies" topical report, Reference 7. Time-averaged fluxes and fluence values throughout the reactor and vessel were calculated with the DORT discrete ordinate computer code using three-dimensional synthesis methods. The basic theory for synthesis is described in Section 3.0 of Reference 7. DORT three-dimensional synthesis results are the bases for the fluence predictions using the Framatome ANP "Semi-Analytical" (calculational) methodology.
1.4.1 Analytic Uncertainty Analysis The uncertainties in the SONGS Unit 3 fluence values have been evaluated to ensure that the greater than 1.0 MeV calculated fluence values are accurate with no discernible bias, and have a mean standard deviation that is consistent with the Frarnatome ANP benchmark database of uncertainties. Consistency between the fluence uncertainties in the updated calculations for SONGS Unit 3 Cycles 1-11 and those in the Framatome ANP benchmark database ensures that the vessel fluence predictions are consistent with So2.3-q23-Mq7 e
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the 10 CFR 50.61, Pressurized Thermal Shock (PTS) screening criteria and Regulatory Guide 1.99, Reference 13, embrittlement evaluations.
The verification of the fluence uncertainty for the SONGS Unit 3 reactor includes:
Estimating the uncertainties in the Cycles 1 through 11 dosimetry measurements, Estimating the uncertainties in the Cycles I through 11 benchmark comparison of calculations to measurements, Estimating the uncertainties in the Cycles I through 11 pressure vessel fluence, and Determining if the specific measurement and benchmark uncertainties for Cycles 1-10 are consistent with the Framatorne ANP database of generic uncertainties in the measurements and calculations.
The embrittlement evaluations in Regulatory Guide 1.99 and those in 10 CFR 50.61 for the PTS screening criteria apply a margin term to the reference temperatures. The margin term includes the product of a confidence factor of 2.0 and the mean embrittlement standard deviation. The factor of 2.0 implies a very high level of confidence in the fluence uncertainty as well as the uncertainty in the other variables contributing to the embrittlement. The dosimeter measurements from the SONGS Unit 3 analysis would not directly support this high level of confidence. However, the dosimeter measurement uncertainties are consistent with the Framatome ANP database. Therefore, the calculational uncertainties in the updated fluence predictions for SONGS Unit 3 are supported by 728 additional dosimeter measurements and thirty-nine benchmark comparisons of calculations to measurements as shown in Appendix A of Reference 7. The calculational uncertainties are also supported by the fluence sensitivity evaluation of the uncertainties in the physical and operational parameters, which are included in the vessel fluence uncertainty, Reference 7. The dosimetry measurements and benchmarks, as well as the fluence sensitivity analyses in the topical report, are sufficient to support a 95 percent confidence level with a confidence factor of 42.0 in the fluence results from the "Semi-Analytical" methodology.
The Framatome ANP generic uncertainty in the dosimetry measurements has been determined to be unbiased and has an estimated standard deviation of 7.0 percent for the qualified set of dosimeters. The SONGS Unit 3 Cycles 1-11 dosimetry measurement uncertainties were evaluated to determine if any biases were evident and to estimate the standard deviation. The dosimetry measurements were found to be appropriately calibrated to standards traceable to the National Institute of Standards and Technology and are thereby unbiased by definition. The mean measurement uncertainty associated with Cycles 1-11 is as follows:
a=
6.22%
This value was determined from Equation 7.6 in Reference 7 and indicates that there is consistency with the Framatome ANP database. Consequently, when the database is updated, the SONGS Unit 3 Cycles 1-11 dosimetry measurement uncertainties may be combined with the other 728 dosimeter measurements.
Since Cycle 1-11 measurements are consistent with the database, it is estimated that the SONGS Unit 3 dosimeter measurement uncertainty may be represented by the database standard deviation of 7.0 percent, perAppendix D of Reference 4. Based on the database, there appears to be a 95 percent level of So23. - q23-M q7 Ad O Pot e 17 WCAP-16167-NP,RevOO Page 1-7 November 2004 S023-923-M97
confidence that 95 percent of the SONGS Unit 3 dosimetry measurements for fluence reactions above 1.0 MeV are within 4:142 percent of the true values.
1.4.2 Comparison with Benchmark and Plant Specific Measurements The Framatome ANP generic uncertainty for benchmark comparisons of dosimetry calculations relative to the measurements indicates that any benchmark bias in the greater than 1.0 MeV results is too small to be uniquely identified. The estimated standard deviation between the calculations and measurements is 9.9 percent. This implies that the root mean square deviation between the calculations of the SONGS Unit 3 dosimetry and the measurements should be approximately 9.9 percent in general and bounded by L2O.04 percent for a 95 percent confidence interval with thirty-nine independent benchmarks.
The weighted mean values of the ratio of calculated dosimeter activities to measurements (C/M) for Cycles -11I have been statistically evaluated using Equation 7.15 from Reference 7. The standard deviation in the benchmark comparisons is as follows:
crcim = 6.36%
This standard deviation indicates that the benchmark comparisons are consistent with the Framatome ANP database. Consequently, when the database is updated, the Cycles I-1I benchmark uncertainties may be included with the other thirty-nine benchmark uncertainties in Reference 7.
1.4.3 Overall Bias and Uncertainty The consistency between the Cycles 1-11 benchmark uncertainties and those in the database indicates that the SONGS Unit 3 fluence calculations for Cycles 1-11 have no discernible bias for fluence values greater than 1.0 MeV. In addition, this consistency indicates that the fluence values can be represented by the Framatome ANP reference set which includes a calculational standard deviation of 7.0 percent at dosimetry locations. That is, the uncertainty in the calculated neutron fluence values is as shown in Table 1-3:
Table 1-3 Neutron Fluence Uncertainty Uncertainty (%)
Type of Calculation Standard Deviation (a) 95% / 95% Confidence (- E2a)
Capsule 7.0 14.2 Pressure Vcssel (maximum 10.0 20.0 location)
Pressure Vessel (extrapolation) 11.4 22.8 So23-S -023-M e17 Revs. o acoe
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Figure 1-1 Fluence Analysis Methodology for SONGS Unit 3 Surveillance Capsule S023-f2 MC17 Rev. 0 WCAP-16167-NP, RevOO November 2004 Page 1-9 S023-923-M97
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2.0 REACTOR VESSEL SURVEILLANCE PROGRAM The reactor vessel surveillance program for SONGS Units 2 and 3 is being conducted to monitor the neutron-irradiation induced changes in mechanical properties of the reactor vessel materials. The reactor vessel surveillance program and the surveillance capsule withdrawal schedule are described below and in References 5 and 6. The reports describing the pre-irradiation and post-irradiation evaluations of the surveillance materials are contained in References 4, 14 & 15.
The surveillance program for SONGS Units 2 and 3 was designed in accordance with ASTM El 85-70, "Standard Recommended Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Reactor Vessel Material. " ASTM El 85-70 presents criteria for monitoring changes in the fracture toughness properties of reactor vessel beltline materials. The reactor vessel surveillance program for SONGS Units 2 and 3 adheres to all ASTM E185-73 requirements and to 10 CFR 50 Appendix H, with the exception of the method of attachment of the holders for the six surveillance capsules in each unit. At SONGS, the capsule holders are attached directly to the cladding on the inside of the vessel in the beltline region. The current requirements of 10 CFR 50 Appendix H (IB.2) do not treat the method of attachment of the capsule holders as a compliance issue, since it states:
"...Ithe capsule holders are attached to the vessel wall or to the vessel cladding, construction and inservice inspection of the attachments and attachment wvelds must be done according to the requirements for permanent structural attachments to reactor vessels given in Sections KII and Xl of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The design and location of the capsule holders must permit insertion of replacement capsules..."
The capsule holder attachment method for SONGS Units 2 and 3 meets the design and inspection requirements of the ASME Code, Sections m and XI. Therefore, there are no deviations or exceptions needed from the current requirements of 10 CFR 50 Appendix IL 2.1 TEST MATERIAL SELECTION Three metallurgically-different materials representative of the reactor vessel are used for test specimens in accordance with the general guidelines ofASTM E185-73. These materials include base metal, weld metal, and heat affected zone (HAZ) materials.
2.1.1 Plate Material Selection Intermediate and lower shell plate materials are nearest to the reactor core; hence they will sustain the greatest neutron exposure. Each of the six plates which make up the intermediate and lower shell courses were evaluated, Reference 6, in terms of initial RTNDT, copper content, and its effects on the NDTT shift o023 -cZ3
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at the end-of-life (EOL). Adjusted RTimT calculations showed that base metal plate C-6802-1 had the highest value (limiting plate) and was selected to be the base plate material for the San Onofre Unit 3 surveillance program.
Base metal test materials for SONGS Unit 3 were manufactured from sections of intermediate shell plate C-6802-1, Reference 6. The section of shell plate used was adjacent to the test material used for ASME Code Section Ell tests and was at a distance of at least one plate thickness from any water-quenched edge.
This material was heat-treated to a metallurgical condition representative of the final metallurgical condition of the base metal in the completed reactor vessel.
In addition to the base materials from sections of the reactor vessel shell plate, material from a standard heat ofASTM A533 B I steel, made available through the NRC-sponsored Heavy Section Steel Technology program, is also included. This reference material has been fully processed and has been characterized as to the sensitivity of its mechanical and fracture toughness properties to neutron radiation embrittlement. Correlation monitors provide an independent check on the measurement of the estimated irradiation conditions for the surveillance materials. Compilation of data generated from post-irradiation tests of the correlation monitors has been carried out in the Heavy Section Steel Technology Program.
A summary of the materials included in the six surveillance capsules is presented in Table 2-1 (from Reference 6).
2.1.2 Weld Material Selection The weld material for the SONGS Unit 3 surveillance weld was selected to duplicate weld seam 9-203, Reference 6. The "equivalency approach" for selecting weld material was not used.
Weld metal and HAZ material specimens for SONGS Unit 3 were produced by welding together sections from plates C-6802-2 and C-6802-3. The surveillance weldment for SONGS Unit 3 was fabricated using 3116 inch diameter bare wire of Type Mil B4, heat number 90069 and Linde Type 124 flux. HAZ test material was manufactured from a section of the same shell plate used for the base metal test material.
The section of shell plate used for weld metal and HAZ test material are adjacent to the test material used for ASME Code Section I(( tests and are at a distance of at least one plate thickness from any water-quenched edge. Heat-treatment of the surveillance weld materials was equivalent to the heat treatment accorded the reactor vessel. A summary of the weld and HAZ materials used in the surveillance capsules is presented in Table 2-1 (cf., Reference 6).
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2.2 TEST SPECIMENS 2.2.1 Type and Quantity The magnitude of the neutron-induced property changes of the reactor vessel materials is determined by comparing the results of tests using irradiated impact and tensile specimens to the results of similar tests using unirradiated specimens. Changes in RTNDr of the vessel materials were determined by adding to the reference temperature (RTNDT) the amount of the temperature shift in the Charpy test curves between the unirradiated material and the irradiated material, measured at 30 ft-lb. The new reference temperature values are known as the "adjusted reference temperatures" (ART).
Drop weight, Charpy impact, and tensile test specimens were provided for unirradiated tests. Drop weight tests were conducted in accordance with ASTM E208. Charpy impact tests were conducted in accordance with ASTM E23. Tensile tests were conducted in accordance with ASTM E8 and E21.
Correlation of drop weight and Charpy impact tests to establish reference temperature were made in accordance with NB-2300 of the ASME Code, Section EII. Charpy impact and tensile test specimens are provided for post-irradiation tests.
The total quantity of specimens furnished for carrying out the overall requirements of this program is presented in Reference 6. A sufficient amount of base metal, weld metal, and HAZ test material to provide two additional sets of test specimens has been obtained with full documentation and identification for future evaluation should the need arise. Each of the test materials was chemically analyzed for approximately 21 elements, including all those listed in Paragraph 4.1.3 of ASTM El 85-73.
2.2.2 Unirradiated Specimens The type and quantity of test specimens provided for establishing the properties of the unirradiated reactor vessel materials are presented in Reference 6. The data from tests of these specimens provide the basis for determining the neutron-induced property changes of the reactor vessel materials.
Drop Weight Test Specimens: Twelve drop weight test specimens, each of the base metal (longitudinal and transverse), weld metal, and HAZ material are provided for establishing the nil ductility transition temperature (NDTI) of the unirradiated surveillance materials. These data form the basis for RTNDT determination. RTNDT is the reference temperature from which subsequent neutron-induced changes are determined.
Charpy Impact Test Specimens: Thirty test specimens, each of base metal (longitudinal and transverse), weld metal, and HAZ material are provided for impact testing. This quantity exceeds the minimum number of test specimens recommended byASTM El 85 for developing a Charpy impact energy transition curve and is intended to provide a sufficient number of data points for establishing 023-.q3
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accurate Charpy impact energy transition temperatures for these materials. These data, together with the drop weight NDTT, are used to establish an RTNDT for each material.
Unlaxlal Tension Test Specimens: Eighteen tensile test specimens, each of base metal (longitudinal and transverse), weld metal, and HAZ materials are provided for tension testing. This quantity also exceeds the mninimum, number of test specimens recommended byASTM El 85 and is intended to permit a sufficient number of tests to accurately establish the tensile properties for these materials at a minimum of three test temperatures; e.g., ambient, operating, and design.
2.2.3 Irradiated Specimens Both tensile and impact test specimens are used for determining changes in the static and dynamic properties of the materials due to neutron irradiation. A total of 288 Charpy impact and 54 tensile test specimens are provided. The type and quantity of test specimens provided for establishing the properties of the irradiated materials over the lifetime of the vessel are presented in Table 2-1 (cf., Reference 6).
2.3 SPECIMEN IRRADIATION 2.3.1 Encapsulation of Specimens The test specimens are housed within corrosion-resistant capsule assemblies in order to:
- Prevent corrosion of the carbon steel test specimens by the primary coolant during irradiation,
- Physically locate the test specimens in selected locations within the reactor, and
- Facilitate the removal of a desired quantity of test specimens from the reactor when a specified fluence has been attained.
A typical SONGSS Unit 3 capsule assembly (cf., References 5 and 6) consists of a series of seven specimen compartments, connected by wedge couplings, and a lock assembly. Each compartment enclosure of the capsule assembly is internally supported by the surveillance specimens and is externally prcssure tested during final fabrication. The wedge couplings also serve as end caps for the specimen compartments and position the compartments within the capsule holders, which are attached to the reactor vessel. The lock assemblies fix the locations of the capsules within the holders by exerting axial forces on the wedge coupling assemblies which cause these assemblies to exert horizontal forces against the sides of the holders preventing relative motion. The lock assemblies also serve as a point of attachment for the tooling used to remove the capsules from the reactor.
Each capsule assembly is made up of four Charpy impact test specimens (Charpy impact) and three tensile test specimen-flux/temperature monitor (tensile-monitor) compartments. Each capsule compartment is assigned a unique identification so that a complete record of test specimen location within each compartment can be maintained.
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2.3.1.1 Charpy Impact Compartments Each Charpy impact compartment contains 12 Charpy impact specimens. This quantity of specimens provides an adequate number of data points for establishing an impact energy transition curve for a given irradiated material. Comparison of the unirradiated and irradiated Charpy impact energy transition curves permits determination of the RTNDT changes due to irradiation for the various materials.
The specimens are arranged vertically in four I x 3 arrays and are oriented with the notch toward the core.
The temperature differential between the specimens and the reactor coolant is minimized by using spacers between the specimens and the compartment and by sealing the entire assembly in an atmosphere of helium.
2.3.1.2 Tensile-Monitor Compartments Each tensile-monitor compartment contains three tensile test specimens, a set of neutron flux monitors, and a set of temperature monitors for estimating the maximum temperature to which the specimens have been exposed. The entire tensile-monitor compartment is sealed within an atmosphere of helium. The tensile specimens are placed in a housing machined to fit inside the compartment. Split spacers are placed around the gage length of the specimens to minimize the temperature differential between the specimen gage length and the coolant.
2.3.2 Flux and Temperature Measurement The changes in the RTNMT of the reactor vessel materials are derived from specimens irradiated to various fluence levels and in different neutron energy spectra. In order to permit accurate predictions of the RTNDr of the vessel materials, complete information on the neutron flux, neutron energy spectra, and the irradiation temperature of the surveillance specimens must be available.
2.3.2.1 Flux Measurements Neutron flux measurements are obtained from detectors located in each of the six irradiation capsules.
Such detectors are particularly suited for the proposed application because their effective threshold energies lid in the low MeV range. (See References 5 and 6 for a list of detectors used.) Selection of threshold detectors is based on the recommendations ofASTM E261, "Method ofMeasuring Neutron Flux by Radioactive Techniques."
Neutron threshold detectors can be used to monitor the thermal and fast neutron spectra incident on the test specimens. These detectors possess reasonably long half-lives and activation cross-sections covering the desired neutron energy range. One set of neutron flux spectrum monitors is included in WCAP-16167-NP, Rev O D3 S-I MQ7 ev. o B e 25 Page2-5 S023-923-M97
each tensile-monitor compartment. Each detector is placed inside a sheath which identifies the material and facilitates handling. Cadmium covers are used for those materials; e.g., uranium, nickel, copper and cobalt, which have competing neutron capture activities. The flux monitors are placed in holes drilled in stainless steel housings at three axial locations in each capsule assembly to provide an axial fluence profile for each set of test specimens.
In addition to these detectors, the program also includes correlation monitors (i.e., Charpy impact test specimens made from reference heat ASTM A53 3 E I) which are irradiated along with the specimens made from reactor vessel materials. The changes in impact properties of the reference material provide a cross-check on the dosimetry in any given surveillance program. These changes also provide data for correlating the result from this surveillance program with the results from experimental irradiations and other reactor surveillance programs using the same reference material.
2.3.2.2 Temperature Estimates Because the changes in mechanical and impact properties of irradiated specimens are highly dependent on the irradiation temperature, it is necessary to have knowledge of the temperature of the specimens as well as that of the pressure vessel. During irradiation, instrumented capsules are not practical for a surveillance program extending over the design lifetime of a power reactor. The maximum temperature of the irradiated specimens can be estimated with reasonable accuracy by including in the capsule assemblies small pieces of low melting point alloys or pure metals. The compositions of candidate materials with melting points in the operating range of power reactors are listed in References 5 and 6. The monitors are selected to bracket the operating temperature of the reactor vessel.
The temperature monitors consist of a helix of low melting alloy wire inside a sealed quartz tube. A stainless steel weight is provided to destroy the integrity of the wire when the melting point of the alloy is reached. The compositions and therefore the melting temperatures of the temperature monitors are differentiated by the physical lengths of the quartz tubes which contain the alloy wires.
A set of temperature monitors is included in each tensile-monitor compartment. The temperature monitors are placed in holes drilled in stainless steel housings and are also placed at three axial locations in each capsule assembly to provide an axial profile of the maximum temperature to which the specimens were exposed.
2.3.3 Irradiation Locations The encapsulated test specimens are irradiated at approximately identical radial positions about the midplane of the core. The test specimens are enclosed within six capsule assemblies at axial positions that are bisected by the midplane of the core.
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The test specimens contained in the capsule assemblies are used to monitor the irradiation induced property changes of the reactor vessel materials. These capsules, therefore, are positioned near the inside wall of the reactor vessel so that the irradiation conditions (fluence, flux spectrum, temperature) of the test specimens resemble, as closely as possible, the irradiation condition of the reactor vessel. The neutron fluence of the test specimens is expected to be within 15% of that seen by the adjacent vessel wall. The RTNDT changes resulting from the irradiation of these specimens closely approximate the RTNDT changes in the materials of the reactor vessel.
The capsule assemblies are placed in capsule holders positioned circumferentially about the inside of the reactor vessel. Table 2-2 (Reference 4) presents the exposure locations for the capsule assemblies. All capsule assemblies were inserted into their respective capsule holders during the final reactor assembly operation.
23.4 Capsule Assembly Removal Surveillance capsule assemblies are withdrawn during an appropriate refueling outage when the test specimens have attained the desired fluence. The target or actual neutron fluence for removal of each capsule assembly is presented in Table 2-2 (Reference 4).
The target fluence levels for the surveillance capsules were determined for each azimuthal location and for the time intervals indicated in the withdrawal schedule in 10 CFR 50 Appendix H (1133.3). The Unit 3 capsule assembly located in the 97-degree position was withdrawn as described in Reference 15. The Unit 3 capsule assembly located in the 263-degree position was withdrawn as described in Reference 4.
Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdowns most closely approaching the withdrawal schedule.
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Table 2-1 Type and Quantity of Specimens for Irradiation Exposure and Irradiated Tests (Ref. 6)
Quantity of Specimens Type of Specimen Orientation Base Weld HAZ SRM l
Tota Metal Metal HZ Si t
oa Longitudinal 48 24 72 h
Transverse 72 72 72 216 Uniaxial Tensile Lgit Transverse 18 I18 54 Total 138 90 90 24 342
- Standard Reference Material (SRM) characterized by Heavy Section Steel Technology program; specimens are provided only for correlation with characterization tests.
Table 2-2 SONGS Unit 3 Capsule Assembly Removal Schedule (Ref. 4)
Capsule Azimuthal Removal Time (EFPY)
Fluence ( s Io,, n/cm2)
Number Location (degrees) 1 83*
25 4.03 2
97 4.33 0.8 3
104 Standby 4
284 Standby 5
263 14.9 2.471 6
2770 Standby
- Either the 830 or the 2770 capsule can be withdrawn after 25 EFPY, while the other capsule remains on standby.
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3.0 LTOP SYSTEM LIMITS 3.1 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM The Low Temperature Overpressure Protection (LTOP) system protects the reactor coolant system (RCS) pressure boundary integrity by ensuring that the RCS pressure remains below the applicable P-T limits of 10 CFR 50, Appendix G. particularly at low temperatures when the RCS is water-solid.
SONGS Unit 3 Tech Spec LCO 3.4.12.1 protects the design basis assumptions for the LTOP system that no more than two IPSI pumps can be operable and that the safety injection tanks must be isolated or depressurized to less than the limit specified in LCO 3.4.3. In the event that the RCS is below the enable temperature, the Shutdown Cooling System (SDCS) relief valve shall be operable or the RCS vented with an area greater than 5.6 sq. in.
LCO 3.4.12.2 specifies that the SDCS relief valve or the pressurizer code safety valves must be operable above the enable temperature specified in LCO 3.4.3. The following sections describe the process for developing the LTOP system limits and demonstrating adequate performance at SONGS Unit 3.
3.2 BASIS FOR LTOP SYSTEM LIMITS The design basis for the LTOP System for SONGS Unit 2 and 3 is described in Reference 17. The LTOP system limitations consist of a SDCS relief valve satpoint aligned whenever the RCS temperature is below an enable temperature along with controls on the RCS heatup and cooldown rates. The relief valve setpoint and capacity have been selected such that the peak transient pressures in the postulated overpressure events do not exceed the applicable RCS P-T limits presented in LCO 3.4.3. The development of the setpoint follows the plant-specific methods described in Reference 17 and the results bound the NRC approved methodology contained in Section 3.0 of Reference 1.
Calculated limiting temperatures for LTOP heatup and cooldown protection are given in Table 3-1.
Table 3-1 Low Temperature RCS Overpressure Protection Range Operating Period Indicated Cold Leg Temperature During Heatup During Cooldown Through 32 EFPY (Normal and Remote
- 239.5°F S 214.30F Shutdown Panel Operation)
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3.2.1 Peak Transient Pressures Analyses of the worst-case energy addition (RCP start) and mass addition (HPSI start) overpressure events were performed to calculate the peak transient pressures. The design basis for SONGS Units 2 and 3 is contained in Reference 17, which describes the models and assumptions used to produce a conservative LTOP model and analysis for these overpressure transients. These analytic models present a bounding design for LTOP that allows the plant to operate with discretionary conservatism relative to the NRC-approved analytic models identified in Reference 1. It should be noted that the SDCS relief valve installed at SONGS Unit 3 (PSV-9349) has excess capacity relative to relieving requirements, thus any LTOP pressure transient is quickly terminated upon valve actuation. The design basis transients displayed in UFSAR Figures 5.2-6 and 5.2-7 clearly show the relative severity for several potential overpressure transients and the rapid termination of such transients.
These design basis peak pressure analyses incorporate the following assumptions:
- The pressurizer is initially water-solid, i.e., no steam space,
- The RCS pressure boundary is rigid, i.e., no expansion due to pressure or thermal effects,
- No heat is transferred to or from the RCS,
- The RCS letdown flow is isolated,
- All pumps attain rated speed instantaneously,
- Only one relief valve (PSV-9349) is used in the transient mitigation,
- No operator action is required, and
- Conservative energy addition sources are used for both energy addition and mass addition transient scenarios, including:
o Full heat output from all pressurizer heaters (1500 Kw) is assumed for the duration of the transient in order to maximize the energy input into the RCS, and o Decay heat, increased by 10% for conservatism, is assumed constant throughout the transient at a value consistent with the earliest time after shutdown that the transient can occur.
The following additional assumptions are made to assure a conservative analysis:
The SDCS is assumed isolated at the start of the transient in order to minimize the total volume absorbing the heatlmass addition and to isolate any heat removal from the RCS, The SDCS relief valve opening profile is conservative relative to the ASME model described in Section 3.2.1.1. This results in a delayed response to the relief valve lift and a resulting delay in providing the relief capability,
- No RCP seal leakage or controlled bleed-off is assumed,
- The RCS is isothermal and is not cooled or heated by any mass addition, and S3301-O23-Ml 7 g
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- The initial conditions are chosen to maximize the pressure transients in order to develop the greatest rate of pressure rise. The RCS pressure at the initiation of the transients is selected to be 376 psia, which is the highest pressure for Shutdown Cooling System operation and is also the highest pressure for the SDCS relief valve being in service.
The following sections discuss the details for the transient analysis used in the determination of the SONGS Unit 3 LTOP design basis.
3.2.1.1 ReliefValve Overpressure Protection LTOP system overpressure protection at SONGS Unit 3 is performed by a spring actuated relief valve located in the SDCS suction line. This valve is placed in service at the LTOP enable temperature in order to protect the reactor vessel from brittle fracture in the event of a low-temperature overpressure event.
The SDCS relief valve passes subcooled water due to its location in the SDCS piping. Valve opening and discharge characteristics are consistent and conservative relative to the ASME Code requirements for spring loaded safety valves and/or the manufacturer's recommendations, whichever is more conservative.
This SDCS valve has a relieving capacity substantially greater than that needed to mitigate the design basis mass addition and energy addition transients affecting the reactor coolant system. The valve setpoint is sufficiently below the limiting reactor coolant system pressure established by 10 CFR 50 Appendix G and referenced in the Technical Specifications, thus assuring LTOP protection of the reactor vessel.
The SDCS relief valve design parameters are a lift pressure of 417 psia allowing 3089 gpm to pass when fully open at 10% pressure accumulation. These relief valve parameters were selected for protection of the SDCS and as shown in the LTOP analyses are quite conservative and require no changes to provide the necessary LTOP function. The analytical model for the valve opening and its associated capacity prior to the 10% accumulation is an important characteristic for the limiting pressure transient scenario.
The NRC-accepted LTOP model in Reference I follows the ASME Code model with an initial opening at 3% accumulation, while the SONGS Unit 3 design basis evaluation was conducted with a model that delayed opening until 7% accumulation. This conservatism resulted in the relief valve opening at a 4%
accumulation delay (7% versus 3%) that increased the peak pressure transient before the pressure excursion was terminated. It should be noted that the pressure rise is terminated in both limiting transients before the relief valve reaches full open.
3.2.1.2 Mass Addition Overpressure Events The design basis mass addition transient was identified as an inadvertent actuation of two HPSI pumps while all three charging pumps are operating at their design flowrate. This event was analyzed by o.23- °123-t1-47 R.
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determining incremental inputs for HPSI mass additions, charging pump mass addition, and the equivalent-mass additions that result frorn energy additions. The RCS, assumed to be a single node at a uniform temperature and pressure, remains at the initial temperature and volume while the mass addition (Reference 17) results in a pressure rise. The additional mass is added to the original mass and divided by the system volume to calculate an updated specific volume. A new RCS system pressure is then determined using the initial temperature and the updated specific volume. Finally, the updated pressure was assumed to be the "back pressure" for the HPSI pump delivery curve in order to determine the revised HPSI delivery for the next time step.
The HPSI mass addition was obtained from curves developed for ECCS calculations using two-pump delivery with no losses, which is the maximum volumetric delivery at any pressure difference between the reactor coolant system and the refueling water tank. A conservative low temperature is assumed for the supply water to establish the largest mass addition rate.
The overpressure transient results given in SONGS Unit 3 UFSAR Figure 5.2-6 show that the values for the HPSI mass addition are significantly greater than the pressurizer heater effects. The incremental effect for relief valve discharge flow was combined within each time increment to develop the transient curves shown in UFSAR Figures 5.2-7. The time steps were small since the RCS was water solid and the pressure rise was rapid. The transient was quickly mitigated due to the large capacity discharge through the relief valve, making the cumulative effect of the decay heat and the pressurizer heaters inconsequential. The peak pressure is less than 450 psia as shown in UFSAR Figure 5.2.7.
3.2.13 Energy Addition Overpressure Events The SONGS LTOP design basis results for transient energy addition events were determined with OVERP, a computer code that simulates the pressure increase to a solid RCS due to reverse heat transfer from relatively hot steam generators when an idle RCP is started. A detailed description of the OVERP computer code is provided in Reference 18. OVERP, used extensively in LTOP analyses performed by Westinghouse, simulates the discharge from a relief device and determines pressure during the relieving action. An earlier (mnainframe) version of the OVERP computer code was used in the original design basis analysis for SONGS; the current model runs on a personal computer platform. The SDCS relief valve opening characteristics in the current SONGS design basis analysis assumes valve opening at 7%° accumulation rather than 3% accumulation listed in Reference 1. Although this results in the relief valve opening later and the RCS pressure transient peaking at a slightly higher value, the relieving capacity is sufficient to protect the RCS. The OVERP model used in the SONGS design bases evaluations complies with the NRC requirements in a conservative manner.
The following paragraphs discuss the energy addition model and input parameters that further illustrate the conservative nature of the earlier design basis calculation contained in the design report, Reference 17.
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The following assumptions are included in the initiation of this design basis energy addition transient:
The steam generator secondary temperature is assumed to be 1000F hotter than the primary coolant temperature.
One RCP is assumed to start and instantaneously reach rated speed to initiate the transient. The model assumes a constant heat input for the duration of the analysis.
The analytical model assumption of constant specific volume results from the RCS fluid mass remaining constant by assuming zero RCP seal leakage and no charging flow.
With the relatively large capacity SDCS relief valve, the energy addition pressure transient is mitigated immediately upon the valve opening at 7% accumulation. The peak pressure is less than 450 psia as shown in UFSAR Figure 5.2.7.
3.2.2 Applicable P-T Limits The P-T limits for the SONGS Unit 3 LTOP system setpoints were developed using the methodology described in Section 5.0. These heatup and cooldown P-T limits are listed in Tables 5-1 through 5-3 and are shown in Figures 5-1 through 5-3 to permit comparison of the P-T values with the peak transient pressures given in Section 3.2.1. Applicable P-T limits are established based on the method described in Reference 1, which performs a comparative evaluation of the P-T limitations developed per Section 5.0 and the peak pressurizer transient evaluation of Section 32.1.
Applicable limiting heatup and cooldown rates for SONGS Unit 3 are presented in Appendix A.
Figures A-1 through A-3 presented in Appendix A are identical to Figures 5-1 through 5-3, respectively, but simplified for clarity and case of use by plant operators.
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4.0 BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE The calculation of the adjusted reference temperature (ART), Reference 19, for the beltline region,was performed using the NRC-accepted methodologies as described below. The limiting ART values in the beltline region for the SONGS Unit 3 reactor vessel corresponding to 32 EFPY for the 1/4T and 3/4T locations are as follows:
Location ART (E F
Limiting Material 1/4T 145.80F Intermediate Shell Plate C-6802-1 3/4T 125.50F Intermediate Shell Plate C-6802-l The RTpTs value for the SONGS Unit 3 reactor vessel was calculated in accordance with 10 CFR 50.61 using the neutron fluence at the clad-base metal interface after 32 EFPY. The highest predicted value is 154.60F and corresponds to Intermediate Shell Plate C-6802-1. The RTpTs determination is defined in Section 4.4.
4.1 BACKGROUND
Given below is the determination of the adjusted reference temperature (ART) for the SONGS Unit 3 reactor vessel beltline materials for 20 and 32 effective full power years. These results are consistent with the measurements obtained from the 263-degree surveillance capsule described in Reference 4, and the 97-degree surveillance capsule described in Reference 15.
4.2 RESULTS The results of the adjusted reference temperature calculations are summarized in Table 4-1:
Table 4-1 Calculated Adjusted Reference Temperatures at 20 and 32 EFPY Location I
FY Adjusted Rererence Temperature 1/4T 3/4T Plate C-6802-1 20 137.1°F 116.40F Plate C-6802-1 32 145.80F 125.50F The calculated ART values reflect the added confidence from using the measured properties of the vessel, for example the Charpy transition temperature shifts as a result of measured surveillance capsule evaluations. These data were used tojustify reducing the standard deviation for transition temperature shift from 340F to 170F in conjunction with a chemistry factor of 720F. (See Section 7.0.)
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4.2.1 Assumptions The input and assumptions from Reference 19 were used when calculating the adjusted reference temperatures applicable to beltline materials and limiting plates and welds.
4.2.2 Fluence Calculation Fluence is a function of time and location. Reference 4 provides the fluence values at the vessel clad interface for the end of Cycle 11 (14.925 EFPY) and for EOL (32 EFPY). Fluence can be calculated by linear interpolation for any intermediate time.
The following equation from Reference 13 gives the attenuation of fluence with distance into the plate:
f (e0 24x)
(4-1) where:
f fluence at the desired location, frf = fluence at the wetted surface of the vessel, x
= distance, in inches, from the wetted surface of the vessel to the desired location.
4.2.3 Chemistry Factor Calculation The Chemistry Factor (CF) is determined using Tables I and 2 of Reference 13. Table I is used for weld materials; for plate materials, Table 2 is used. The Chemistry Factor values determined in this fashion are reported in Reference 19.
Surveillance data, when available, are used to determine a Chemistry Factor according to the following equations from Reference 13:
ARTNDT = (CF)
- ff (4-2) where: f... f(f28 - 0.10 1og f)
(4-3) and:
CF = I(ARTNDT
- fM / Iff 2 (44)
A test of the validity of the estimated Chemistry Factor (CF) consists of calculating ARTNDT for a given fluence and comparing it with the measured ARTNDT for that fluence. The measured ARTNDT must fall within +/-a1a of the calculated GRTNDT where Ca = 17F for base metal and a = 28TF for welds (Reference 13).
WCAP-16167-NP,RevOO 023 qz Rt-ket/- o AREY e-6 Page4-2 November 2004 s
S023-923-M97
4.2.4 Calculation of ART for the Limiting Plates at 1/4T and 3/4T Adjusted reference temperatures are calculated using the following equation from Reference 13:
ART = Initial RTNDT + ARTNDT + Margin (4-5) where all temperatures are in degrees F.
4.3 ANALYSIS DETAILS 4.3.1 Selection of Representative and Limiting Cases All beltline plates and welds were evaluated for the purpose of determining the maximum ART values.
The material with the highest ART was the intermediate shell plate C-6802-1.
4.3.2 Wetted Surface Fluence Calculation The following values were obtained from Reference 19:
End of Cycle 11 (EOC 11) = 14.925 EFPY Wetted surface cumulative fluence, as shown in Table 4-2.
The fluence values for 20 EFPY, shown in Table 4-2, were calculated by linear interpolation.
Table 4-2 Calculated Wetted Surface Fluence (10"1 ncm t )
Location 14.925 EFPY 20 EFPY 32 EFPY Intemnediate shell2011269497 (Plates C-6802-1, 2, 3) 2.0111 2.659 4.1907 Lower shell 1.9496 2.575 4.0545 (Plates C-6802-4, 5, 6) 1.46255404 For each location (intermediate and lower shells), the fluence at l/4T and 3/4T are calculated by using Equation (4-1), with the following values of depth (x) from Reference 4:
x = 2.375 inches for 1/4T, and x = 6.6875 inches for 314T.
The fluence factors listed in Table 4-3 were calculated by using Equation 4-3.
Rev 00 3_ 12 3_ tCt 7
/tv November 2004 S
~
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Pero e-3 7 Page 4-3 S023-923-M97
Table 4-3 Fluence (10" n/cm) and Fluence Factors at 1/4T and 314T Location EFPY 114T fT 114T ff**
314T f l
3/4T fl**
Intermediate Shell 20 1.5037 1.1129 0.5342 0.8248 (Plates C-6802-1, 2, 3) 32 2.3699 1.2328 0.8419 0.9517 Lower Shell 20 1.4562 1.1042 0.5173 0.8159 (Plates C-68024, 5, 6) 32 2.2929 1.2244 0.8145 0.9424
- f = neutron fluence
- ff - fluence factor per Equation 4-3 4.4 CALCULATION OF THE ADJUSTED REFERENCE TEMPERATURES AND SELECTION OF THE MAXIMA The values of CF, f, aj and the initial RTNmr are known for all the materials of interest and are provided in Reference 19. The values of CF and ca for lower shell plate C-6802-1 were determined based on surveillance program data as described in Section 7.0. The adjusted reference temperatures are calculated using Equation (4-5). According to Reference 13:
Margin = 2 (c;2 + a,2)12 (4-6)
However, c,& need not exceed one-half of the predicted shift (Reference 13). According to Reference 16, the standard deviation of the initial RTNwr, Gj, is = 0, and Margin = 2*a.
The calculated values of ART appear in Table 4-4, with the highest values listed in Table 4-1. The adjusted reference temperature values developed here are used to define the P-T limits for SONGS Unit 3 for 32 EFPY (EOL).
Similar calculations were performed for the limiting plate and weld at end of life following I0 CFR 50.61. RTm was calculated for each of the beltline plates and welds using the values of CF and initial RT1,,DT, shown in Table 4-4 and the margin derived as described above. Calculations were based on a peak fluence at the vessel clad base metal interface after 32 EFPY, 3.847 x 10'9 n/crm2 in the lower shell and 3.976 x 10" n/cm2 in the intermediate shell. The highest predicted RTm is 154.67F and corresponds to intermediate shell plate C-6802-1.
o2 92 Mr-q7 Pte". 0 Po6 e33 WCAP-161 67-NM, Rev 00 November 2004 Page 44 S023-923-M97
N l'
0 l
c To Table 4-4 Calculated Adjusted Reference Temperatures at 1/4T and 314T Initial Wet. Surr.
1/4 T Position 3/4 T Position Plate/I EFPY RT,.T CF Margin (IF)
Fluence Fluence
~
T AT Fluence ATD R
Weld No.
(
w 11/4,3M4 (EA1 9 (EA19 ff ART AR (EA19 f
ARTlm AR)T 20 40 72 17.0,17.0 2.659 1.504 1.113 80.1 137.1 0.534 0.825 59.4 116.4 C-6802-1_
32 40 72 17.0,17.0 4.191 2.370 1.233 88.8 145.8 0.842 0.952 68.5 125.5 20 10 26 28.9,21.4 2.659 1.504 1.113 28.9 67.9 0.534 0.825 21.4 52.9 C-6802-2 32 10 26 32.1, 24.7 4.191 2.370 1.233 32.1 74.1 0.842 0.952 24.7 59.5 C-6802-3 20 20 37 34.0,30.5 2.659 1.504 1.113 41.2 95.2 0.534 0.825 30.5 81.0 32 20 37 34.0,34.0 4.191 2.370 1.233 45.6 99.6 0.842 0.952 35.2 89.2 20 10 31 34.0,25.3 2.575 1.456 1.104 34.2 78.2 0.517 0.816 25.3 60.6 C-6802-4 32 10 31 34.0,29.2 4.055 2.293 1.224 38.0 82.0 0.815 0.942 29.2 68.4 20 10 26 28.7,21.2 2.575 1.456 1.104 28.7 67.4 0.517 0.816 21.2 52.4 C-6802-5 32 10 26 31.8,24.5 4.055 2.293 1.224 31.8 73.7 0.815 0.942 24.5 59.0 20 20 37 34.0,30.2 2.575 1.456 1.104 40.9 94.9 0.517 0.816 30.2 80.4 C-6802-6
___7_
34_0_4_0_4_055_1_24 45_3_99_3 32 20 37 34.0,34.0 4.055 2.293 1.224 45.3 99.3 0.8 15 0.942 34.9 88.9 WCAP-16167-NP, Rev 00 November 2004 Page 4-5
w I
0o Table 44 (Continued)
Calculated Adjusted Reference Temperatures at 114T and 3/4T Wet. surr.
1/4 T Position 314 T Position Plate/
IFY RniJtial Margin (IF)
Fluence FuneFune~
R Weld No.
EFPY RTT 11/4,3/41 nkm)
Fiuenme 2
ARTNDT
- 2)
ART 2-203A 20
-40 40 44.5,33.0 2.659 1.504 1.113 44.5 49.0 0.534 0.825 33.0 26.0 32
-40 40 49.3,38.1 4.191 2.370 1.233 49.3 58.6 0.842 0.952 38.1 36.1 2-203B 20
-40 50 55.6,41.2 2.659 1.504 1.113 55.6 71.3 0.534 0.825 41.2 42.5 32
-40 50 56.0,47.6 4.191 2.370 1.233 61.6 77.6 0.842 0.952 47.6 55.2 2-203C 20
-40 32 35.6,26.4 2.659 1.504 1.113 35.6 31.2 0.534 0.825 26.4 12.8 32
-40 32 39.4,30.5 4.191 2.370 1.233 39.4 38.9 0.842 0.952 30.5 20.9 3-203A 20
-70 44 48.6,35.9 2.575 1.456 1.104 48.6 27.2 0.517 0.816 35.9 1.8 32
-70 44 53.9,41.5 4.055 2.293 1.224 53.9 37.7 0.815 0.942 41.5 12.9 3-203B 20
-70 42 46.4,34.3 2.575 1.456 1.104 46.4 22.8 0.517 0.816 34.3
-1.5 32
-70 42 51.4,39.6 4.055 2.293 1.224 51.4 32.8 0.815 0.942 39.6 9.2 3-203C 20
-70 44 48.6,35.9 2.575 1.456 1.104 48.6 27.2 0.517 0.816 35.9 1.8 32
-70 44 53.9,41.5 4.055 2.293 1.224 53.9 37.7 0.815 0.942 41.5 12.9 9-203 20
-50 34 37.5, 27.7 2.575 1.456 1.104 37.5 25.1 0.517 0.816 27.7 5.5 32
-50 34 41.6,32.0 4.055 2.293 1.224 41.6 33.3 0.815 0.942 32.0 14.1 WCAP-16167-NP, Rev 00 Page 4-6 WCAP-16167-NP, Rev 00 November 2004 Page 4-6
5.0 PRESSURE-TEMPERATURE LIMTS USING LIMITING ADJUSTED REFERENCE TEMPERATURE IN THE P-T CURVE CALCULATION Analytical methodology approved by the NRC and described in Reference 1 is used to develop the beltilne RCS pressure-temperature limits for SONGS Unit 3. These pressure-temperature limits will be removed from Technical Specification LCO 3A.3 and updated limits presented in Appendix A of this PTLR. The PTLR methodology is also documented in the SONGS Unit 3 Technical Specifications Bases.
RCS pressure-temperature limits established for non-beltline locations do not change significantly due to the lower exposure to neutron flux unless they are updated through regulation or more recent advances in technology. However, it is appropriate to consider these non-beltline locations, as necessary, in the updating of pressure-ternperature limits throughout plant life, as they are currently part of the plant design basis. Therefore, the pressure-temperature limits for the beltline region are combined with the non-beltline regions, as appropriate, to develop the set of composite curves for specific modes of operation in this PTLR. The lower bound of these composite curves defines the pressure-temperature limit for a plant at a specific mode of operation. The pressure-temperature limits for the non-beltline regions are relocated from the SONGS Unit 3 Technical Specifications and incorporated into this PTLR.
5.1 RCS TEMPERATURE RATE-OF-CHANGE LIMITS Information describing the rate-of-change linits for SONGS Unit 3 will be removed from the current Technical Specifications and updated limits valid through 32 EFPY relocated into Appendix A of this PTLR. The specific heatup and cooldown rate limits specified in LCO 3.4.3, SR 3.4.3.1, and SR 3.4.3.2 are replaced with text describing that the allowable limits are located in the PTLR. Technical Specification Figures 3.4.3-1 through 3.4.3-5 and Table 3.4.3-1 are removed in their entirety.
5.2 RCS PRESSURE-TEMPERATURE LIMITS Tables 5-1 through 5-3, shown plotted in Figures 5-1 through 5-3, provide heatup, cooldown, in-service hydrostatic and leak testing, and criticality pressure and temperature limits for SONGS Unit 3 through 32 EFPY.
_c7 Rev 0 WCAP-161 67-NP, Rev 00 Page 5-1 November 2004 S023-923-M97
Table 5-1 SONGS Unit 3 Heatup at 60°F/hr RCS Pressure-Temperature Limits through 32 EFPY Control Room Instrumentation Indicated RCS Indicated RCS Temp (IF)
Pressure (psia) 58.5 579.9 68.5 571.7 78.5 593.7 88.5 570.7 98.5 557.7 108.5 551.7 118.5 554.7 128.5 564.7 138.5 581.7 148.5 606.7 158.5 587.7 168.5 631.7 178.5 687.7 188.5 757.7 198.5 844.7 208.5 951.7 218.5 1084.7 228.5 1245.7 238.5 1442.7 248.5 1686.7 258.5 1983.7 268.5 2343.7 278.5 2789.7 288.5 3331.7 298.5 3727.7 308.5 3727.7 318.5 3727.7 Note: Pressure and temperature values shown are adjusted for instrument uncertainty, and for RCS pressure and elevation effects. The pressure shift at 606.7 psia results from the change in pressure correction factors applied to the low-range vice the wide-range pressure instrumentation.
,9 23 3 -t 4-7 f~ot5 WCAP-16167-NP, Rev 00 November 2004 Page 5-2 S023-923-M97
Table 5-2 SONGS Unit 3 Cooldown via Control Room Instrumentation RCS Pressure-Temperature Limits through 32 EFPY RCS PSIA at PSIA at PSIA at PSIA at PSIA at PSIA at PSIA at R
100°F/hr 80F/hr 60OF/hr 40°F/hr 30°F/hr 20°F/hr 10°F/hr Temp ( ;)
Cooldown Cooldown Cooldown Cooldown Cooldown Cooldown Cooldown 58.5 228.9 294.9 362.9 432.9 468.9 505.9 542.9 68.5 252.7 314.7 379.7 447.7 482.7 517.7 553.7 78.5 282.7 339.7 400.7 465.7 498.7 532.7 567.7 88.5 318.7 369.7 426.7 486.7 518.7 550.7 583.7 98.5 361.7 406.7 457.7 512.7 542.7 572.7 582.7 106.9 624.9 106.9 571.5 108.5 415.7 451.7 495.7 545.7 571.7 599.7 116.1 624.9 116.1
-5 71.5 118.5 480.7 506.7 541.7 584.7 607.7 122.4 624.9 122.4 571.5 127.0 624.9 127.0 571.5 128.5 560.7 573.7 598.7 1323 624.9 132.3 571.5 134.7 624.9 134.7 571.5 135.1 624.9 135.1 571.5 138.5 604.7 148.5 722.7 158.5 822.7 168.5 898.7 178.5 990.7 188.5 1104.7 198.5 1242.7 208.5 1411.7 218.5 1617.7 228.5 1869.7 238.5 2177.7 248.5 2553.7 258.5 3012.7 268.5 3168.7 278.5 3168.7 288.5 3169.7 298.5 3170.7 308.5 3172.7 318.5 3173.7 Note: Pressure and temperature values shown are adjusted for instrument uncertainty, and for RCS pressure and elevation effects.
The pressure shift from 624.9 psia to 571.5 psia results from the change in pressure correction factors applied to the low-range vice the wide-range pressure instrumentation.
- Data above 108.5'F (at 10°F/hr) to 135.1°F (at 80'F/hr) are not used to establish PTLR cooldown limits.
WCAP-16167-NPRevOO 5,0 2 3'a -..
2-Mrq-7 P'.ev o November 2004 Pae 4 -3 Page 5-3 S023-923-M97
Table 5-3 SONGS Unit 3 Cooldown via Remote Shutdown Panel Instrumentation RCS Pressure-Temperature Limits through 32 EF1Y PSIA at PSIA at PSIA at PSIA at PSIA at PSIA at PSA at RCS 1 00F/hr 80°F/hr 60°F/hr 40°Fthr 30°F/hr 20°F/hr 10°F/hr Temp (IF)
Cooldown Cooldown Cooldown Cooldown Cooldown Cooldown Cooldown 585 180.4 246.4 314.4 384.4 420.4 457.4 494.4 685 204.7 266.7 331.7 399.7 433.7 469.7 504.7 785 233.7 290.7 352.7 416.7 450.7 484.7 518.7 885 269.7 321.7 377.7 438.7 469.7 502.7 535.7 985 313.7 358.7 408.7 464.7 493.7 524.7 555.7 108.5 366.7 403.7 446.7 496.7 523.7 551.7 580.7 118.5 432.7 128.5 511.7 138.5 609.7 148.5 727.7 158.5 827.7 168.5 903.7 178.5 995.7 188.5 1109.7 198.5 1247.7 208.5 1416.7
_ =
218.5 1617.7 228.5 1869.7 238.5 2177.7 248.5 2553.7 258.5 3012.7 268.5 3168.7 278.5 3168.7 288.5 3169.7 298.5 3170.7 308.5 3172.7 318.5 3173.7 Note: Pressure and temperature values shown are adjusted for instrument uncertainty, and for RCS pressure and elevation cffects. The pressure shift from the low-range to the wide-range pressure instrumentation occurs at approximately 1600 psia for 100°F/hr cooldown.
- Data above 108.S°F for 10°F/hr to 80°F/hr are not used to establish PTLR cooldown limits.
90 23 - 'q 2 M q -7
?Z ev- 0 WCAP-1 6167-NP, Rev 00 November 2004 Page 5-4 S023-923-M97
Figure 5-1 SONGS Unit 3 RCS Heatup Pressure-Temperature Limits through 32 EFPY-Normal Operation*
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I!I 100 200 300 Indlcaled Reactor Coolant Temperature, T. (F) 400
- The more conservative of either the Lowest Service Temperature or the minimum temperature requircmcnts for the reactor vessel when the RCS is pressurized to greater than 20% of PHTP should be used in the development of plant P-T limits.
9023_
M q-7 R ev. 0 P..(De I4i5 WCAP-16167-NP, Rev 00 November 2004 Page 5-5 S023-923--M97
I Figure 5-2 SONGS Unit 3 RCS Cooldown Pressure-Temperature Limits through 32 EFPY-Normal Operation*
zluu JE A I; I I I IIl I I t5
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ICO 200 31 Indicated Reactor Coolanl Tompratura, T. ('IF) 3^w 400
- The more conservative of either the Lowest Service Tcmpcrature or the minimum temperature requirements for the reactor vessel when the RCS is pressurized to greater than 20% of PHTP should be used in the development of plant P-T limits. For clarity, data for cooldown rates from I 0°F/hr through 80°F/hr are ormitted above nominally 140°F.
9So3j2_3- 0q7
,ev 0 WCAP-16167-MP,Rcv OO November 2004 S023-923-M97 Page 5-6
Figure 5-3 SONGS Unit 3 RCS Cooldown Pressure-Temperature Limits through 32 EFPY-Remote Shutdown Panel Operation*
2500 2000 M
v 1500 N
0.
% 1000 500 0
f UNACCEPTABLE I
Temperature I
II 7f I1 Temp.. 208.5F 20Ffi 2Dji I
eer a
100 200 300 400 Indicated Reactor Coolant Tomperature. T. (IF) a The more conservative of either the Lowest Service Temperature or the minimum temperature requirements for the reactor vessel when the RCS is pressurized to greater than 20% of PHTP should be used in the development of plant P-T limits. For clarity, data for cooldown rates from I 0°F/hr through 80'F/hr are omitted above 108.5°F.
9023-q23 -
Mq-7 Rev. 0o 4-7 WCAP-16167-NP, Rev 00 November 2004 S023-923-M97 Page 5-7
This page intentionally blank.
5023 - q233-M q 7 WCAP-1 61 67-NP, Rev 00 November 2004 Rev.
O Page 5-8 S023-923-M97
6.0 MINIMUM TEMPERATURE REQUIREMENTS IN THE PRESSURE-TEMPERATURE CURVES The minimum temperature requirements specified in Appendix G to 10 CFR 50 are applied to the pressure-temperature curves using the NRC-approved methodologies as described in Section 6.0 of Reference 1.
The minimum temperature values applied to the pressure-temperature curves of SONGS Unit 3 corresponding to 32 effective full power years are:
Table 6-1 Minimum Temperature Requirements for SONGS Unit 3 through 32 EFPY Requirement Minimum Temperature Minimum Bolt-Up Temperature 650F Minimum Hydrotest Temperature 229.30F Lowest Service Temperature 208.50F Minimum Flange Limit (NOP) 178.50F Minimum Flange Limit (Hydrotest) 148.50F The lowest service temperature is established for CE NSSSs based on the limiting RTNDT for the reactor coolant pumps.
In the development of pressure-temperature limits for CE NSSSs, the intent is to utilize the more conservative of either the Lowest Service Temperature or the other minimum temperature requirements for the reactor vessel when the RCS is pressurized to greater than 20% of the preservice hydrostatic test pressure.
The "minimum pressure criteria" specified in 10 CFR 50 Appendix G serves as a regulatory breakpoint in the development of pressure-temperature limits and is defined as twenty percent of pre-service hydrostatic test pressure. For CE NSSS plants, the preservice hydrostatic test pressure is defined as 1.25 times the design pressure. The function of minimum pressure in the development of pressure-temperature limits is to provide a transition between the various temperature only based pressure-temperature limits, such as minimum bolt up and the lowest service temperature or flange limits.
For SONGS Unit 3, the minimum pressure is calculated as follows:
Minimum Pressure, Uncorrected:
= (1.25 x Design Pressure) x 0.20 = (1.25 x 2500 psia) x 0.20 625 psia 902 3-L23-t4-7 RevP 0 q
WCAP-16167-NP. Rev 00 Page 6-1 November 2004 S023-923-M97
With pressure corrections due to flow, elevation, and instrument uncertainties, the limiting minimum pressure for SONGS Unit 3 through 32 EFPY becomes:
Table 6-2 Limiting Indicated Minimum Pressure Requirements for SONGS Unit 3 through 32 EFPY For the Control Room:
Minimum Pressure Requirement*
TRCS < 340.00F PRcs = 527.2 psia TRcS > 340.0F PQcs = 507.2 psia For the Remote Shutdown Panel Minimum Pressure Requirement*
TRcS < 340.00F PRcS = 478.7 psia TRcs > 340.00F PRCS = 458.7 psia
- Note: The limiting temperature of 340'F results from the assumed transition from two-reactor coolant pump operation to three-reactor coolant pump operation.
9023-923 -
Aq7 RQev- 0 Plose5(
5'0 WCAP-1 6167-NP, Rev 00 November 2004 Page 6-2 S023-923-M97
7.0 APPLICATION OF SURVEILLANCE DATA TO ADJUSTED REFERENCE TEMPERATURE CALCULATIONS Post-irradiation surveillance capsule test results for SONGS Unit 3 are given in References 4 and 15. The test results were evaluated with respect to the credibility criteria of Regulatory Guide 1.99, Reference 13; data supporting the credibility analysis are presented in Reference 4. The criteria were assessed as follows:
The surveillance program plate or weld duplicates the controlling reactor vessel beltline material in terms of ART. There is one controlling plate in terms of the ART, which is the same heat as the surveillance program plate.
Charpy data scatter does not cause ambiguity in the determnination of the 30 fl-lb shift, The measured shifts are consistent with the predicted shifts, The capsule irradiation temperature is comparable to that of the vessel, and Correlation monitor data are available and meet the credibility test to be (well) within the scatter band of the database for that material.
Credible surveillance data were used to refine the chemistry factor and the margin term in accordance with the methodology prescribed in Position 2.1 ofRegulatory Guide 1.99, Revision 2. Calculated chemistry factor values for the surveillance plate and weld materials are shown in Table 7-1. The credibility test for the surveillance capsule measurement is shown in Table 7-2. In addition, the measured shift for the correlation monitor material from the 263-degree capsule was 1421F, from Table 7.6 of Reference 4, versus the predicted shift of 1 59.80F (see Table 7-3). Therefore, the correlation monitor material meets the credibility test to be within the scatter band of the database for that material.
The derived chemistry factor for the limiting plate C-6802-1 is 720F. The margin on the shift, crE, is 170F.
The value of the margin on initial RINDT, as, is zero because there are measured values for the plate. The total margin as prescribed in Position 2.1 of Regulatory Guide 1.99, Revision 2 is then taken as:
Margin = 2 (c2u + c2 )"' = 2*
when vi is 0.
When the surveillance data have been shown to be credible, the margin on the shift can be halved; i.e., c,& is 8.50 F and the total margin (2aA) is 1 70F.
The calculation of adjusted reference temperature, ART, for use in determination of the pressure-temperature limits is described in Section 4.
902-3-923-7 (3.
O P-os-e.
5*/
WCAP-16167-NP, Rev 00 Page 7-1 November 2004 S023-923-M97
Table 7-1 Calculation of Chemistry Factors for Surveillance Plate and Weld Material Location 97 degree 263 degree Sum Chemistry Capsule Capsule Factor Fluence (xlO"9 n/cm2) 0.800 2.471 Fluence factor (ii) 0.937 1.243 (fi)2 0.879 1.546 2.425 PLATE C-6802-1 g
-- '-5 MeasuredARTNDT 58 0F l
96°F if*ARTNDT 54.4 T 119.4 J
173.8 720F WELD Heat 90069 I U a -1
=
Measured,RTN 300 F 72°F F ff*ARTNDT 28.1 89.5 l
117.6 48.5SF Table 7-2 Credibility Test for Surveillance Plate and Weld Material Material
_A Chemistry Fluence Fluence ARTNDT ARTNDT+a ARTNT<-c Measured Factor (sIO" Factor (CF-fO ARTNDry nlcm ')
(if )
Plate 0.8 0.937 67.50F 84.5 0F 50.50F 580F C-6802-1 17 720F 2.471 1.243 89.50F 106.5°F 72.50F 960F Weld 0.8 0.937 45.4°F 73.4 0F 17.4AF 30°F Heat 28 48.50F 2.471 1.243 60.30F 88.3°F 32.30F 72°F 90069 *
Table 7-3 Credibility Test for Standard Reference Material (SRM)
- CF calculated using NRC Regulatory Guide 1.99, Rev. 2, Table 2, with copper content of 0.174% and nickel content of 0.665% for the standard reference material (from Reference 20).
902- - q23-MMq7 Rev-o P--5e..
5Z WCAP-16167-NP, Rev 00 November 2004 Page 7-2 S023-923-M97
8.0 REFERENCES
- 1.
CE NPSD-683-A, Rev 06, 'The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications," September 2000.
- 2.
NRC Generic Letter 96-03, "Relocation of Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, 'January 31, 1996.
- 3.
U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.
- 4.
"Analysis of the 263-degree Capsule, Southern California Edison Company, San Onofre Unit 3 Nuclear Generating Station, Reactor Vessel Material Surveillance Program," BANV-2454, Lynchburg, Virginia, January 2004.
- 5.
"Program for Irradiation Surveillance of San Onofre Reactor Vessel Materials, San Onofre Nuclear Generating Station, Units 2 and 3," Combustion Engineering, Inc., Windsor, Connecticut, S-NLM-002, Revision 2, October 1974.
- 6.
"Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of San Onofre Unit 3 Reactor Vessel Materials," Combustion Engineering, Inc., Windsor, Connecticut, TR-S-MCM-003, June 25, 1979.
- 7.
J. R. Worsham, et al., "Fluence and Uncertainty Methodologies," BAW-2241P-A. Revision I, Framatome ANP, Lynchburg, Virginia, April 1999.
- 8.
L. A. Hassler, and N. M. Hassan, "SORREL, DOT Input Generation Code User's Manual," NPGD-TM-427, Revision 10, Framatome ANP, Lynchburg, Virginia, May 2001.
- 9.
D. T. Ingersoll, et. al., "BUGLE-93, Production and Testing of the VITAMIN--B6 Fine Group and the BUGLE-93 Broad Group Neutron/photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data," ORNL-DLC-175, Radiation Safety Information Computational Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April 1994.
- 10.
L. A. Hassler, and N. M. Hassan, "GIP User's Manual for B&W Version, Group Organized Cross Section Input Program," NPGD-TM-456, Revision 11,Fraratome ANP, Lynchburg, Virginia, August 1994.
- 11.
M. A. Rutherford, N. M. Hassan, eL al., Eds., "DORT: Two Dimensional Discrete Ordinates Transport Code," BWNT-TM-107, ramatome Technologies, Inc., Lynchburg, Virginia, May 1995.
- 12.
J. R. Worsham, "BUGLE-93 Response Functions," FRA-ANP, Document Number 32-1232719-00, Revision 0, Framatome ANP, Lynchburg, Virginia, June 1995
- 13.
U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials,"
Regulatory Guide 1.99, Revision 2, May 1998.
- 14.
A. Ragl, "Southern California Edison Company, San Onofre Unit 3, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program," Combustion Engineering, Inc., Windsor, Connecticut, TR-S-MCM-004, November 30, 1979.
WCAP-16167-NP, RevOO 30 3 zq3 -
3 q q7 p
e v
5-3 Page 8-I November 2004 S023-923-M97
- 15.
"Analysis of the Southern California Edison Company San Onofre Unit 3 Reactor Vessel Surveillance Capsule Removed from the 970 Location," WCAP-12920, Rev. 2, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, May 1994.
- 16.
SCE Calculation No. N-0220-020, Revision 4, "SONGS 2/3 Adjusted Reference Temperature (ART) for 8 EFPY."
- 17.
Low Temperature Reactor Coolant System Overpressure Protection for San Onofre Units 2 and 3, December 15, 1977.
- 18.
WVCAP-1 5688, Rev 00, "CE-NSSS LTOP Energy Addition Transient Analysis Methodology," May 2001.
- 19.
SCE Calculation No. M-0011-074, Rev. 0, "SONGS Unit 3 Adjusted Reference Temperature for 20 and 32 EFPY," March 3,2004.
- 20.
E.D. Eason, et al., "Improved Embrittlcmcnt Correlations for Reactor Pressure Vessel Steels,"
NITREGICR-655 1, Nuclear Regulatory Commission, November 1998.
5o23-T2.3-P-49'7 WCAP-16167-NP, Rev 00 November 2004 RRVr 0
rRPe5 5
2 Page 8-2 S023-923-M97
I APPENDIX A Technical Specification References to the PTLR (Provided by SCE)
The P-T limits information contained in Appendix A is extracted from the PTLR and displayed in a format similar to SCE's existing Technical Specification. This Appendix provides a convenient centralized location for information relocated from the Technical Specification to the PTLR in the format familiar to SCE's Operation Group.
This Appendix is currently a sample representation. SCE will replace this sample Appendix with final information.
S02-3-_c(3-Mq-7 Rev. 0 1
6-e-55 WCAP-16167-NP, Rev OO November 2004 A-l S023-923-M97
APPENiDixA TECHNICAL SPECIFICATION REFERENCES TO THE PTLR APPENDIX A TECHNICAL SPEcIFICAllON REFERENCES TO THE PTLR 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PM1) Limits LCO 3.4.3 The combination of RCS pressure, RCS temperature and RCS heatup and cooldown rates shall be maintained within the limits as specified in the RCS PRESSURE-TEMPERATURE LIMITS REPORT (PTLR).
With the reactor vessel head bolts tensioned*, the Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures A-1 through A-3 and Table A-1 during heatup, cooldown, and Inservice leak and hydrostatic testing with:
- a.
A maximum heatup of 60 OF in any 1-hour period with RCS cold leg temperature greater than or equal to 65 "F.
- b.
A maximum cooldown of,100 r ny 1-hour period with RCS cold leg temperature greater thaf)
"tornormal operation. A maximum cooldown of 100 "F in any I-h 7' 0rio ith RCS cold leg temperature greater than 123 OF for remot sh'
- c.
A mail own of 80 OF In any 1-hourperiod with RCS cold leg t
greater than 111 OF for normal operation. A maximum cooldown of 80.iny 1-hour period with RCS cold leg temperature greater than 120 OF for remote shutdown operation.
- d.
A maximum cooldown of 60 OF In any 1-hour period with RCS cold leg termperature greater than 101 OF for normal operation. A maximum cooldown of 60 Fin any 1-hourperiod with RCS cold leg temperature greaterthan 113 "F for remote shutdown operation.
- e.
A maximum cooldown of 40 OF in any 1-hourperiod with RCS cold leg temperature greater than or equal to 78 OF for normal operation. A maximum cooldown of 40 OFin any 1-hour period with RCS cold leg temperature greater than 99 OF for remote shutdown operation.
- f.
A maximum cooldown of 30 "F in any 1-hourperiod with RCS cold leg temperature greater the 65 OF for normal operation. A maximum cooldown of 30°F in any 1-hour period with RCS cold leg temperature greater than 86 OF for remote shutdown operation.
- g.
A maximum cooldown of 20 "Fin any 1-hourperiod with RCS cold leg temperature greater than 65 OF for remote shutdown operation.
- h.
A maximum temperature change of 10 OF in any 1-hourperiod during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
3029 3
9q23_ M q7 l-ev o
-e_
5 '
WCAP-1 6167-NP, Rev 00 A-2 November 2004 S023-923-M97
APPENDIXA TECHNICAL SPECIFICATION REFERENCES TO THE PTLR
- i.
A minimum temperature of 65 OF to tension reactor vessel head bolts.
With the reactor vessel head bolts detensioned, the Reactor Coolant System (except the pressurizer) temperature shall be limited to a maximum heatup or cooldown of 60 OF in any 1-hour period.
With the reactor vessel head bolts detensioned, RCS cold leg temperature may be less than 65 OF.
(iS S02 3 - 12 3 _
7 q 7 Revt. C Pe5 5 7 WCAP-16167-NP, Rev 00 November 2004 A-3 S023-923-M97
APPENDDX A TECHNICAL SPECIFICATION REF-ERENCES TO THE PTLR APPENDIX A TECHNICAl.. SPECIFICATION REFERENCES TO THE PTLR 3/27/03 APPLICABILITY:
At all times.
ACTIONS 3.4.3 CONDITION REQUIRED ACTION COMPLETION TiME A. _NOTE--
Required Action A.2 shall be completed whenever this Condition is entered.
Requirements of LCO not met in MODE 1, 2. 3. or 4.
(
B.
,'-)
Required Action and ass;ated Completion Time of Condition A not met.
C.
NOTE-Required Action C.2 shall be completed whenever this Condition is entered.
Requirements of LCO not met any time in other than MODE 1, 2,3, or4.
A.1 Restore parameter(s) to within limits.
30 minutes AND A.2 B.1 Be in MODE 3.
AND B.2 Be in MODE 5 with RCS pres
< 500 psia.
C.1 Initiate action to restore parameter(s) to within limits.
AND C.2 Determine RCS is acceptable continued operation.
for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 6 hours sure 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Immediately for Prior to entering MODE 4
$V023 - 012 3 - Mq 7 I, ev. 0 F1 c (5 e-5-S WCAP-1 61 67-NP, Rev 00 November 2004 A-4 S023-923-M97
APPENDix A TECHNICAL SPECIFICATION REFERENCES TO THE PTLR APPENDIX A TECHNICAL SPECIFICATiON REFERENCES TO THE PTLR 3/27/03 3.4.3 Continued SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.
NOTE-Only required to be performed during RCS heatup and cooldown operations and RCS Inservice leak and hydrostatic testing.
Verify RCS pressure, RCS temperature, and rates within limits specified in Figures A-1 thr Iup and cooldown 30 minutes 4-The reactor vessel materipli t'o. *eatce specimens shall be removed and examined' fio.'et changes in material properties, as required by 10 CFR 50 AIpepdi(. The results of these examinations shall be used to update tt$TLR In accordance with requirements of 10CFR 50 Appendix H S023 3 -Mq7 kcv-0 ag 5 7
WCAP-16167-NP, Rev 00 November 2004 A-5 S023-923-M97
APPENDIX A TECHNICAL SPECIFICATiON REFERENCES TO THE PTLR 3/27103 3.4.3 Continued
'Arm~---
I_
I IL Lr
=.I Test Presur (2556 psla) 2500 2000 C
a'i A.
IL c1500v la I.0.
-Acceptable operating region -to
- the right of the hestbp curve In all
- modes. In addition. in mo~des I and
_ 2 the cperating region Is b the right t
of Ihe care critical curve.
I Y..
,I
-Accetpablet oper abng ntegon - tot bth
_right of the inse#rvle tests curve
_(Applicatbl in modes cthe than Mod..
_1 end 2)
I I
9-.9 i.
9 99 9
,99 t
t I
99I.
9
.9 I
l I
9, I
9' I
' I t:
9I.
,99
.9..
99 9,,,.
19
.9.*,
9.
I
.9 i
9 99, 9
., 9 9999 t,99
.t 9 9 I*.
1 9
.9 9,
j:,
I J7 T
i_
Lowest Servce Temnp.. 208.5-F 1
- 9 9
9 I
I 99i.9 I
9999$I
.99
- 99 991$
9,*99I9
,9999'l i
9 I
999#
- 99 999l 1
.99 1
j 99,i
,9.
9 99*
.,,99.9 1
9 i i
9
.,,.9i I Nr rts= lity Based o~n Inservice Hydrotstatic; Test Tempeaie (28.9F) fto Specified RTemr
.9 9.
i 9 9 9
99, i9.
- 99.
t 400
- I i I I I I.11II I I ;.;_I jII I
A I
I I
I I
I I
0 100 200 300 Indicateid Reactor CoolantTemperaturs. T. rF)
FIGURE A-1 SONGS Unit 3 RCS Heatup P-T Limits through 32 EFPY Normal Operation 5023-q23-m-1 7 Re.
6 o WCAP-1 6167-NP. Rev O0 November 2004 A-6 S023-923-M97
APPENDKx A TECHNICAL SPECIFIcATIoN REFERENCES TO THE PTLR APPEDIXA TCHNIAL PECFIC~ON EFEENCE TOTHEPTL 3/27/03 3.4.3 Continued 114-WI
[ 1 1 I til}
r l,
I-ll i
I l
'i!
IiIIUnacctab
_-II_
f~~
piang J
-I Cooldown 1l0OF/hr Rein ALL-JL! I_.'
.jl 2,500[1.0 I
.-uI I,
lr-V III 1t~II 1--iir '+/-
x: ly, r -:-I' U-L-1: I-I I
S C.
0.
C I
l I
I Li.
fAAIA Ilf'
- Jt U
-4 I /
15tuu.u.0-I II I
I I
I i
I I I
-I
-1 I
- I 1 ! I I I Lowest Service Temp., 208.5-F I iz~ZUWL1 4
500.0 0.0 Acceptable Operating Region Minimum Boltup Temp., 65'F I
71 tj 0
10o 200 300 400 Indicated Reactor CoolantTemporature, T. (-F)
FIGURE A-2 SONGS Unit 3 RCS Cooldown P-T Limits through 32 EFPY Normal Operation S023-q23 - P-1 q 7 Rev.. 0 J 'E.e ' (
WCAP-1 6167-NP, Rev 00 November 2004 A-7 S023-923-M97
APPENDIX A TECHNICAL SPECFICATION REFERENCES TO THE PTLR 3/27/03 3.4.3 Continued 3000.0 2500.0 2000.0 T
Is E
10 Q.
0.
h 1500.0 0.
1000.0 500.0 0.0 I:17 I
ICoodown.
100'F/hr[+4 i -i-Unacceptable I f I t
- ~Lowest Service Temp., 208.5'F I
I Ij.
I Acceptble I.
_4
---I.-,-T 0.0 100.0 200.0 300.0 400.0 Indicated Reactor Coolant Temperature, T. (IF)
FIGURE A-3 SONGS Unit 3 RCS Cooldown P-T Limits through 32 EFPY Remote Shutdown Panel Operation
-S3- '/23-M1q7 PI o
JZ)7e WCAP-1 61 67-NP, Rev 00 November 2004 A-8 S023-923-M97
APPEN 3/27/C MDD A
TECHNICAL SPECIFICATION REFERENCES TO THE PTLR
)3 3.4.3 Continued TABLE A-1 Low Temperature RCS Overpressure Protection Ranqe Operating Period, EFPY Cold Leg Temperature. 'F During Heatup
< 239.5 Through 32 (Normal Operation)
Through 32 (Remote shutdown 4 During Cooldown
< 214.3
< 214.3 Heatup operations are not normally performed from the Remote Shutdown Panel 5-023 -
'2 3-M- 7 RE -vo P
3 WCAP-16167-NP, Rev 00 A-9 November 2004 S023-923-M97
APPENDIX A TECHNICAL SPECIFICATION REFERENCES TO THE PTLR 327/03 3.4.12.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature s 239.5 IF LCO 3.4.12.1 No more than two high pressure safety injection pumps shall be OPERABLE, the safety injection tanks shall be isolated or depressurized to less than the limit specified in Figure A-2 and at least one of the following overpressure protection systems shall be OPERABLE:
- a.
The Shutdown Cooling System Relief Valy jUnit 3, PSV9349) with:
- 1)
A lift setting of 4061 +/-0 psig,
- 2)
Relief Valve isolation valves,(
7, 3HV9339, 3HV9377, and 3HV9378 open, or, Ht
- b.
The Reactor Coolan stem depressurized with an RCS vent of greater than or equal to 5.6 square inches.
APPLICABILITY:
MODE 4 when the temperature of any one RCS cold leg is less than or equal to the enable temperatures specified in Table A-1, MODE 5, and MODE 6 when the head is on the reactor vessel and the RCS is not vented.
NOTET SIT isolation or depressurization to less than the Figure A-2 limit Is only required when SIT pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in Figure A-1 and Figure A-2.
3o2S-q2 3 -mq-1 Re V 0
f-% 64 WCAP-16167-NP, Rev 00 November 2004 A-10 S023-923-M97
APPENDIX A TECCHNICAL SPECIFIcAnIoN REFERENCES TO THE PTLR APPENDIX A TECHNICAL SPECIFICATiON REFERENCES TO THE PTLR 3127/03 3.4.12.1 Continued ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
A.1 With more than two HPSI Initiate action to verify a maximum Immediately pumps capable of injecting into of two HPSI pumps capable of the RCS.
injecting into the RCS.
B.
B.1 SIT pressure is greater than or Isolate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> equal to the maximum RCS pressure for existing cold leg temperature allowed in Figure A-1 and Figure A-2.
..(
C.
C.1 Required Action and associated Depressurize affected SIT to less 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time of Condition B than the maximum RCS pressure not met.
for existing cold leg temperature allowed in Figure A-1 and Figure A-2.
D.
D.1 With one or both SDCS Relief Open the closed valve(s).
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Valve Isolation valves in a single SDCS Relief Valve OR isolation valve pair (Unit 3 valve 0.2 pair 3HV9337 and 3HV9339 or Power-lock open the OPERABLE valve pair 3HV9377 and SDCS Relief Valve isolation valve 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3HV9378) closed.
pair.
(continued)
S?022 -
9 M7q-7 65-WCAP-161 67-NP. Rev 00 November 2004 A-11 S023.-923-M97
APPENDIXA TECHNICAL SPECIFicAnON REFERENCES To THE PTLR 3/27/03 3.4.12.1 Continued CONDITION REQUIRED ACTION COMPLETION TIME E.
E.1 SDCS Relief Valve inoperable.
Reduce Teg to less than 200 0F, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> depressurize RCS and establish OR RCS vent of > 5.6 square inches.
Required Action and associated Completion Time of Condition A, C, or D not met.
OR LTOP System inoperable for any 5
reason other than Condone C~
or D.
529 23- '123 - M7
,e Vv 0
P-,5e e_ 6 G WCAP-16167-NP, Rev 00 November 2004 A-12 S023-923-M97
APPENDixA TECHNICAL SPECIFICATION REFERENCES TO THE PTLR APPENDIX A TECHNICAL SPECIFICATION REFERENCES TO THE PTLR 3/27/03 3.4.12.1 Continued SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.
_NNOTE A HPSI pump is secured by verifying that its motor circuit breaker is not racked-in, or its discharge valve is locked closed. The requirement to rack out the HPSI pump breaker is satisfied with the pump breaker racked out to its disconnected or test position.
Verify a maximum of two HPSI pumps are I RCS.
, ai
)ecting into the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
~NOTE
-`-"
Required to be performen complying with the LCO 3.4.12.1 Note.
Verify each SIT is isolated or depressurized less than the limit specified in Figure A-2.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Verify RCS vent 2 5.6 square inches is open when in use for overpressure protection.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for unlocked open vent valve(s)
AND 31 days for locked, sealed, or otherwise secured open vent valve(s), or open flanged RCS penetrations (continued)
S-o23-2Z3 m-q7 Rev. o 67 WCAP-161 67-NP, Rev 00 November 2004 A-13 S023-923-M97
APPENMAA TECHNICAL SPECIFICATION REFERENCES TO THE PTLR APPENDIX A TECHNICAL SPECIFICATION REFERENCES TO THE PTLR 3127/03 3.4.12.1 Continued SURVEILLANCE FREQUENCY
- 1. Only required to be performed when the SDCS Relief Valve isolation valve pair is inoperable.
- 2. The power-lock open requirement is satisfied either with the AC breakers open for valve pair 3HV9337 and 3HV9339 or the inverter input and output breakers open for valve pair 3HV9377 and 3HV9378, whichever valve pair is OPERABLE.
Verify the OPERABLE SDCS Relief V valve pair 3HV9337 and 3HV9339 -o' 3HV9378) is in the power.l c
.~1 ?,_1 (Unit 3 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> g;0 Verify that SDCS Relief ve isolation valves (Unit 3) 3HV9337, 3HV9339, 3HV9377, and 3HV9378 are open when the SDCS Relief Valve is used for overpressure protection.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Verify SDCS Relief Valve Setpoint.
In accordance with the Inservice Testing Program 5V023_ -2 3 -
M 4? 7 WCAP-1 6167-NP. Rev 00 November 2004 A-14 S023-923-M97
APPENDIX A 3/27/03 TECHNICAL SPECIFICATION REFERENCES TO THE PTLR 3.4.12.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > 239.5.0 IF LCO 3.4.12.2 At least one of the following overpressure protection systems shall be OPERABLE:
- a.
The Shutdown Cooling System Relief Valve (Unit 3, PSV9349) with:
- 1)
A lift setting of 406 i 10 psig,
- 2)
Relief Valve Isolation valves (Unit 328+\\(,
3HV9339, 3HV9377, and 3HV9378 open, or,
- b.
A minimum of one' er code safety valve with a lift setting of 2500 psia + 11%.
APPLICABILITYMODE 4 when the temperature of all RCS cold legs are greater than the enable temperatures specified In Table A-1.
NOTES
- 1.
The lift setting pressure of the pressurizer code safety valve shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
- 2.
The SDCS Relief Valve lift setting assumes valve temperatures less than or equal to 130 "F.
s023-q23- ('17 WCAP-16167-NP, Rev 00 November 2004 R e".0 A-15 S023-923-M97
N APPENDIX A TECHNICAL SPECIFICATION REFERENCES TO THE PTLR 3127/03 3.4.12.2 Continued ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
A.1 No pressurizer code safety Be in MODE 5 and vent the RCS 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> valves OPERABLE.
through a greater than or equal to 5.6 square inch vent.
AND The SDCS Relief Valve INOPERABLE.
t(q-B.
With one or both SDCS iFetlff(
en the closed valve(s).
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Valve Isolation valves inQa_
'J single SDCS Relief Valve 1 OR isolation valve pair (valve air B.2 3HV9337 and 3HV9339 or Power-Lock open the OPERABLE valve pair 3HV9377 and SDCS Relief Valve isolation valve 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3HV9378) closed.
pair.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Note Only required when the SDCS Relief Valve is being used for overpressure protection.
Verify that the SDCS Relief Valve Isolation valves (Unit 3) 3HV9337, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 3HV9339, 3HV9377, and 3HV9378 are open.
Verify relief valve setpoint.
In accordance with the Inservice Testing Program S023-923-M q-7
/?gev. 0 70 WCAP-16167-NP, Rev 00 November 2004 A-16 S023-923-M97
WCAP-16167, Rev 00 Non-Proprietary, Class 3
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- Westinghouse Westinghouse Electric Company, LLC 20 International Drive Windsor, Connecticut 06095
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