ML20039F867
| ML20039F867 | |
| Person / Time | |
|---|---|
| Site: | Palisades, Saint Lucie  |
| Issue date: | 12/31/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML13308A045 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-F, NUDOCS 8201130490 | |
| Download: ML20039F867 (34) | |
Text
- - __ .
93*h 3 Ein'o"x e 50' -
EVALUATION OF PRESSURIZED THERMAL SHOCK EFFECTS DUETO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE ST. LUCIE 1 & 2 REACTOR VESSELS Prepared for FLORIDA POWER AND LIGHT COMPANY NUC.. EAR POWER SYSTEMS DIVISION P, J M POWER bi ==i SYSTEMS couBusDON ENGINEERING. INC.
820113047d
LEGAL NOTICE THis REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COISUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:
A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR ltrufD INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTA81UTY, WITH RESPECT TO THE ACCURACY,
, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOO, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRlVATELY OWNED RIGHTS;OR
- 8. ASSUMES ANY UA81UTIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
I
. _ _ . _ _ - -_ - _ _ . . _ _ . . . . . . _ _ _ _ _ . _ . , _ . _ _ _ - . . _ _ . _ _ _ - . , _ . _ _ _ _ _ . _ _ , . . ~ _ . _ _ , _ . _. _ . . - - .
ABSTRACT This Appendix to CEN-189 provides the plant-specific
, evaluation of pressurized thennal shock effects due to small break LOCA's with extended loss of feedwater for the St. Lucie 1 & 2 reactor vessels. It is concluded that crack initiation would not occur for the transients considered for more than 32 effective full power years, which is assumed to represent full plant life.
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CEN-189 Appendix F TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT F1. PURPOSE F1 F2. SCOPE F1 F3. INTRODUCTION F1 F4. THERMAL HYDRAULIC ANALYSES F1 F5. FLUENCE DISTRIBUTIONS F2 F6. MATERIAL PROPERTIES F9 .
F7. VESSEL INTEGRITY EVALUATIONS F17 F8. CONCLUSIONS F31 1
F1.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized thermal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the St. Lucie 1 & 2 reactor vessels.
F2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA +
LOFW trsnsients presented in CEN-189, as applied to the St. Lucie 1 & 2 reactor vessels.
Other C-E NSSS reactor vessels are reported in separate Appendices.
F
3.0 INTRODUCTION
This Appendix to CEN-189 was prepared by C-E for Florida Power and Light for their use in responding to Item II.K.2.13 of NUREG-0737 for the St. Lucie 1 & 2 reactor vessels.
This Appendix is intended to be a companion to the CEN-189 report.
The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter F5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter F6 reports the plant-specific material properties and change of properties due to irradiation, based on the methods of Chapter 6.0 of the report. Chapter F7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter F6.
F4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to perform the plant-specific vessel evaluation reported in this Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by performing more detailed plant-specific thermal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.
F5. St. Lucie - Unit 1 Fluence Distribution The fluence distribution for St. Lucie - Unit I was developed using the methodology described in Chapter 5. Florida Power and Light supplied an estimated cumulative energy output to December 31,1981 of 3.74 Effective Full Power Years (EFPY) at 2560 Megawatts-thermal (Mwt).
Since no surveillance capsule data is currently available for St. Lucie Unit I the results obtained for Millstone Point Unit 2 were applied because the plants have very similar designs. The rats of accumulation of the peak fluence was scaled by the applicable full power levels of the two plants (2560/2700) to yield a rate of 0.885 x 1018 (n/cm2) per EFPY at 2560 Mwt. The resulting peak fast fluence rate on i.5e St. Lucie Unit I reactor vessel is then estimated to be 3.31 x 1918 (n/cml) as if December 31, 1981 (3.74 EFPY at 2560 Mwt).
The azimuthal shape of the fluence distribution was updated to a St. Lucie 1 distribution as of December 31, 1981 using the results of the Millstone Point-2 RQ-DOT calculations at:d cdju:;tment factors obtained with SHADRAC as described in Section 5.2.2. The nodalization of the core region for the SHADRAC model is shown in Figures F5-1 and F5-2. The time-averaged detailed power distribution used in the SHADRAC calculations was obtained from a Cycle 5 pin power distribution that haa an assembly power distribu-tion closely approximating the Cycle 1 to Cycle 5 time-averaged assembly power distribution. The resulting azimuthal fluence distribution is shown in Figure F5-3. The 00 reference location is shown in Figure F5-4.
One-eighth core symetry is assumed.
The axial and radial fluence distrioutions in the reactor vessel are obtainea for St. Lucie Unit I from a DOT-RZ calculation using the Millstone Point-Unit 2 reference model. The methodology is described in Sections 5.2.3 and 5.2.4. The resulting axial and radial distributions are shown in Figures F5-5 and F5-6.
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i APPENDIX F ST. LUCIE UNITS #1 AND #2 F.6 MATERIAL PROPERTIES F.6.1 St. Lucie Unit #1 The methods used to develop and evaluate the materials for -he St. Lucie Unit #1 reactor vessel are described in Section 6.0 in the main 'x>dy of the report. The chemistry data (nickel, copper, and phosphorus content) and initial (pre-irradiation) toughness properties of the reactor vessel shell course plates and welds are summarized in Table F6-1.
In cases where the weld metal nickel content was not determined, it was conservatively estimated using information on the type of wire used (eg, high MnMo versus MnMcNi wire). For the St. Lucie Unit #1 weldments, the weld inspection records and welding certification reports for weld seam 1-203 indicated that the weld could be expected to contain high nickel (greater than 0.30 w/o) since it was fabricated with MnMoNi wire, so the nickel content was
' conservatively estimated to be 0.99 w/o as shown in Table F6-1. For weld seam 2-203, the recorch indicated that the weld could be expected to contain low nickel (less than 0.30 w/o) since it was fabricated with high MnMo wire, so the nickel content was conservatively estimated to be 0.20 w/o as shown in the Table.
The toughness properties given in Table F6-1 are the drop weight NDTT (if determined) and the initial reference temperature, RTNDT. For the plate materials, the RTNDT was detemined by converting longitudinal impact data using Branch Technical Position MTEB 5-2*. For the weld material, the RT NDT was estimated using the weld qualification test results benchmarked to the Calvert Cliffs Unit #2 surveillance weld. (RTgyi. has not been determined yet for the St. Lucie Unit #1 surveillance weldment,so data from a similar weldment were used instead.) The methodology used is discussed in Section 6.0 and described below.
- " Fracture Toughness Requirements for Older Plants," U.S. Atomic Energy Commission, Regulatory Standard Review Plan.
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The individual weld qualification test results (three Charpy impact specimens tested at +10F) are listed in Table F6-2. Each weld which exhibited an average Charpy energy of 93 ft-lb or greater (the average Charpy energy for the benchmark weld at 10F) was considered to be at least as tough as the benchmark welds i.e., that weld seam RT was NDT
-60F or less. For those weld qualification test results exhibiting an average Charpy energy less than 93 ft-lb, the RT ** " #**** I "
NDT amount equivalent to the temperature difference between the average Charpy energy transition curve for the benchmark weld and the average Charpy energy for the vessel weld test results. In effect, the temperature at which 50 ft-lb or better exists was determined, and the RT was established at a temperature 60F below that value.
NDT A " map" of the cylindrical portion of the St. Lucie Unit #1 reactor vessel is given in Figure F6-1. It shows the locations of the plates and welds listed in Table F6-1 and their corresponding values of initial RT ## * * "# #* ""9 * "
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NDT NDT for the vertical weld seams (designated 1-203, 2-203, and 3-203 ) are shown at a single seam but apply to all three vertical seams in a given shell course. Included in the Figure are the locations of the inlet and outlet nozzles, the core midplane, and the extremities of the active core.
Figure F6-2 is a map of adjusted RT values for important NDT locations at the inner surface of the St. Lucie Unit #1 vessel predicted for December 31, 1981. The predictions are based on the best estimate neutron fluence, 0 330 x 10 n/em (E>1Mev), (corresponding to 3 55 effective full power years at peak flux location on the inside surface of the vessel), the initial RT and copper, phosphorus, and nickel NDT contents given in Table F6-1, and the normalizd neutron flux profiles given in Section F.5. The values of adjusted RT U NDT NDT plus predicted shift) are lccated in rectangles adjacent to the plate and weld designations. The RT values apply to the inner surface of the NDT l vessel in the region indicated by a circle. The circled regions generally represent areas of peak neutron flux for a given weld seam or plate.
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.h F.6.2 St. Lucie Unit #2 The chemistry and initial (pre-irradiation) toughness properties of the St. Lucie Unit #2 reactor beltline materials are summarized in Table F6-3.
The most controlling material in terms of residual ( . istry (copper and '
phosphorus) and initial RTNDT based on Regulatory Guide 1.99, is plate M-605-2 from the intermediato shell course *. The predicted RT g shift based on the 19 n/cm maximum design fluence, 3.66 x 10 (E T . at the inside surface of the reactor vessel is 172F using Regulatory Guide 1.99. This will result in an adjusted RT at end-of-life (32 effective full power years) of 182F at NDT the vessel inside surface. If the design fluence was increased by a factor 2
or two to 7.32 x 1019n/cm , the RTNDT shift is predicted to be 243F for.an adjusted RTNDT of 253F.
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- the basis of the C-E design curve presented in the St. - Lucie Unit #2 FSAR.
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TABLE F6-1
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' Form Identification NDTT (*F) RTNDT ( F) Nickel Copper Phosphorus Plate C6-1 10 .53 .14 b .012 10'a
-30 .53 .14 .011 Plate C6-2 -30 8 b Plate C6-3 -10 -10 a .53 .14 .011 Plate C7-1 0 D .64 .11 .004 a
Plate C7-2 -30 D .64 .11 .004 a
Plate C7-3 -30 10 .58 .11 .004 a
Plate C8-1 -10 20 .56 .15 .006 a
Plate C8-2 0 O .57 .15 .006 a
Plate C8-3 0 O .58 d .12 .004 c .015 Weld 1-203 A,B,&C _N/A -50 c ,99 " .22 Weld 2-203 A,B,8C N/A -50 c .20 .12 .018 Weld 3-203 A,B,8C N/A -50 C .64 .30 .013 Weld 8-203 N/A -60 c .71 .21 .016 Weld 9-203 N/A -60 .11 .23 .013 N/A Not Available a Determined using Branch Technical Position MTEB 5-2 b Estimated based on average of St. Lucie Unit #1 plates having analyses reported c Estimated (see text and Table F6-2) d Estimated Ni content (high nickel type wire) e Estimated Ni content (low nickel type wire)
TABLE f6-2 ST. LUCIE UtilT 01 REACTOR VESSEL WELD SEAM TOUGilNESS DATA d-Charpy Qualification Test Results Average Energy Estimated Weld Seam at 10 F (ft-lb) at 10*F (ft-lb) RTNDT ( F) 1-203 A/C 62, 59, 60 60.3 -50 2-203 A/C 69, 70, 63 67.3 -50 66, 75, 78 73.0 -50 3-203 A/C 87, 82, 92 87.0 -50 3-203 51, 70, 74 65.0 -50 9-203 108, 112, 119 113.0 -60 b
Benchmark Weld ^ 70.5, 78.5, 39.5 93.0 -60 c l
a Benchmark Weld - Calvert Cliffs Unit f2 surveillance weld b Test results at 0*F c Actual RTflDT based on drop weight and Charpy test results d Estimated using the method described in the text S
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TABLE f6-3 ST. LUCIE UNIT #2 REACTOR VESSEL MATERIALS Product Material Orop Weight Initial Chemical Content (%)
Form Identi fication NOTT ( F) RTNDT ( F) Nickel Copper Phosphorus Plate M-605-1 0 +30 0.61 0.11 0.003 Plate M-605-2 -10 +10 0.62 0.13 0.003 c Plate M-605-3 -20 -10 0.61 0.11 0.009
, Plate M-4116-1 -30 +20 0.57 0.06 0.007 0.60 0.07 0.003 i Plate M-4116-2 -50 +20 Plate M-4116-3 -40 +20 0.60 0.07 0.003 Weld 101-214 A,B,&C # -50 to -80 -50 to -30 0.07 0.04 0.011 b 0.003 Weld 101 -142 A,B,&C -50 -50 0.10 0.04 C
Weld 101-171 A,B,8C -70 to -80 -70 to -80 0.08 0.09 0.011 ,
a Intemediate shell course longitudinal seam welds b Lower shell course . longitudinal seam welds c Intermediate to lower shell girth weld I
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F.7.0 St. Lucie 1 Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the St. Lucie 1 vessel. The attenuation of the peak fluence value is considered in three dimensions (r, z, 9), and the superposition of the fluence profile and the weld geometry map is used in calculating the predicted RT NDT value at all points in the vessel as a function of Effective Full Power Years (EFPY). This information is used in locating the points in the vessel having the highest RTNDT at each of the three axial sections of interest:
- 1) middle of care, z = 136.4 in.
- 2) top of core, z = 66. in.
- 3) above core, z = 40.5 in.
where z is the axial distance below the centerline of the nozzle.
From the predicted RT ilDT values, the material toughness properties K and K are determined from the calculated temperatures for the IC Ia SBLOCA + LOFW transients using the method described in Section 7.3.3.
Critical crack depth diagrams are constructed from the applied K y vs crack depth curves and the calculated material toughness curves.
By performing the same fracture mechanics analysis a number of times for increasing plant life (EFPY) the integrity of the St. Lucie 1 vessel for the SBLOCA + LOFW transient is evaluated.
F.7.1 Sumary of Physics and Materials Data Input to Fracture Mechanics Analysis A detailed survey was performed on the combined fluence and material properties maps of the St. Lucie 1 vessel to determine the most critical locations in tems of radiation embrittlement. The properties are considered independently at the three axial sections. At each section, the combination of fluence and materials data were evaluated for a large number of points around the circumference. The adjusted RT NDT values at the inner vessel radius were compared, and the location with the highest RT NDT value was used in the fracture mechanics analysis.
At the mid-core level the location of highest RT flDT ccurs in the plate material at an azimuthal angle of 0 degrees. The fluence factor at this location is 1.0 of the peak fluence in the vessel.
. - 131_ __- - _ - _ _ _-- _ _ _ _ -_ _ . _
i The materials data at this point are as follows:
PCT. Ni = .58 PCT. Cu = .11 PCT. P = .008 Initial RT = 10 F NDT At the 12/31/81 levelof 3.55 EFPY, and peak fluence of .330 x 10 19 2
n/cm (E >l MeV), this corresponds to a point fluence of .330 x 10 19 n/cm2 and an adjusted surface RT valueof50%.
NDT At the top of core level the location of highest RT NDT ccurs in the plate material at an azimuthal angle of 0 degrees. The fluence factor at this location in the vessel is .33 of the peak fluence. The materials data at this point are as follows:
PCT. Ni = .58 PCT. Cu = .11 PCT. P = .008 Initial RT = 10 %
NDT At the 12/31/81 level of 3.55 EFPY, and peak fluence of .330 x 10 19 19 n/cm (E >l MeV), this corresponds. to a point fluence of .109 x 10 n/cm2 and an adjusted surface RT valueof33%.
NDT At the above-core level (about halfway between the top of core and the inlet nozzle), the location of highest RT NDT occurs in the plate material at an azimuthal angle of 0 degrees. The fluence factor at this point is .006 of the peak fluence in the vessel. The materials data for this point are as follows:
PCT. Ni = .53 PCT. Cu = .14 PCT. P = .012 Initial RT = 10 F NDT
fluence attenuation was used such that:
F = exp (-8.625 in x .33 in.-I) - (a/w) o
= exp (-2.85) -
(a/w) where F = point fluence in wall Fo = peak fluence at surface a/w = fractional wall depth Upper bound materials data were used to conservatively envelope all plate and weld materials, which are as follows:
FCT. Cu = .13 PCT. P = .010 Initial RT = +40 F NDT The shift in the value of the RT NDT was determined using the method of Reg. Guide 1.99. This produces an end-of-life prediction for the surface RT NDT of 182 F using the nominal design fluence. A predicted surface RT NDT value of 253 F is detennined for a fluence double the nominal design fluence.
F.7.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW >
Open PORV's (Case 4)
The stress analysis for this case is presented in Section 7.8.1 of the report. The fracture mechanics analyses were performed using the St. Lucie 1 vessel properties and considering fluence levels up to the assumed end-of-life condition of 32 EFPY. Figure F.7-1 shows the critical crack depth diagram for the mid-core level at 32 EFPY. The calculated shift in RT NDT are relatively low, thus the initiation toughness level is not exceeded for this transient. This is indicated by the lack of data on the critical crack depth diagram. Therefore, no crack initiation would occur under these conditions at the mid-core level of the vessel.
Figure F.7-2 is the critical crack depth diagram for the top of core level of the vessel at 32 EFPY. Similarly, the diagram for the level above the core is given in Figure F.7-3. Both of these diagrams indicate that these are no potential regions for crack initiation or crack arrest at these locations in the vessel for this transient.
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F.7.3 Results of Fracture Mechanics Analysis for SBLOCA + LOFW - P Restoration of Feedwater (Case 5)
The stress analysis for this transient is presented in Section 7.8.2 of the report. Fracture mechanics analyses were performed using the St.
Lucie 1 vessel properties with fluence levels up to the assumed end-of-life condition of 32 EFPY. Critical crack depth diagrams were constructed as shown in Figure F.7-4 for the mid-core level of the vessel at 32 EFPY.
The calculated values of Ky for this transient do not exceed the initiation toughness level under these conditions. This indicates that crack initiation would not occur at the mid-core level of the vessel under this combination of loading and vessel material properties.
The critical crack depth diagram for the top of core level of the vessel at 32 EFPY is given in Figure F.7-5. The plot of critical crack depths for the level of the vessel above the core is shown in Figure F.7-6. The fact that no crack initiation regions exist in these figures indicates that no crack initiation would occur in these areas of the vessel throughout the life of the plant for this transient loading condition.
F.7.4 Conclusien The results demonstrate that the integrity of the St. Lucie 1 vessel would be assured throughout the assumed life of the plant for SBLOCA + LOFW transient with recovery of feedwater, and for the SBLOCA + LOFW transient where PORV's are opened.
F.7.5 St. Lucie 2 Vessel Integrity The fracture mechanics analysis is perfomed using upper bound data for fluence and material properties in the St. Lucie 2 vessel. The peak vessel fluence is considered to occur at the point of highest RT The material toughness properties K IC and Kg , are detemined fiDT.
from the calculated temperatures for the SBLOCA + LOFW transients using the method described in Section 7.3.3 and the predicted RTfiDT values through the depth of the wall. Critical crack depth diagrams are constructed from the applied K vs y crack depth curves at the mid-core level of the vessel and the calculated material toughness curves. In this manner the integrity of the St. Lucie 2 vessel is evaluated for the SBLOCA + LOFW transient. ,
F.7.6 Summary of Physics and Material Data Input to Fracture Mechanics Analysis An upper bound fluence value of 3.66 x 10 19 n/cm 2 (E 71 MeV) was used to approximate the end-of-life fluence for the St. Lucie 2 vessel, as well as a conservative upper bound of 7.32 x 10 19 n/cm 2 or double the predicted end-of-life value. The peak fluence is considered to be uniform around the vessel. A conservative radial Figure F.7-8 is the critical crack depth diagram for a similar case with twice the design fluence. Again, the lack of data on the plot indica';es that no crack initiation would occur for this transient even with a much greater accumulated fluence.
F.7.7 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Open PORV's (Case 4)
The stress analysis for this case is presented in Section 7.8.1 of the report. The fracture mechanics analyses were performed using the St. Lucie 2 vessel properties and conservative end-of-life fluence levels.
Figure F.7-7 shows the critical crack depth diagram for stresses ~
considered at the mid-core level coincident with the peak fluence and material properties. An upper bound fluence of 3.66 x 10 l9 n/cm2 was used in this case. The calculated shifts in RT NDT are relatively low, thus the initiation toughness level is not exceeded for this transient. Therefore, no crack initiation would occur under these conditions.
19 At the 12/31/81 level of 3.55 EFPY, and peak fluence of .330 x 10 19 n/cm2 (E > 1 MeV), this corresponds to a point fluence of .002 x 10 n/cm2 and an adjusted surface RT o 15%.
NDT This represents the materials information available at the time of the analysis.
F.7.8 Results of Fracture Mechanics Analysis for SBLOCA + LOFW >
Restoration of Feedwater (Case 5)
The stress analysis for this case is presented in Section 7.8.2 of the report. The fracture mechanics analyses were perfomed using upper bound properties for the St. Lucie 2 vessel and conservative end-of-life fluence levels. The critical crack depth diagram is constructed using the stresses in the transient at the mid-core level coincident with the peak fluence and material properties. Figure F.7-9 shows the critical crack depth diagram for a nominal design fluence of 3.66 x 10 19 n/cm2 . The calculated shifts in RT are relatively NDT low, and ,for this transient loading condition the initiation toughness level is not exceeded. Therefore, no crack initiation would occur for this combination of loading, fluence, and material properties.
Figure F.7-10 shows the critical crack depth diagram for the same transient loading and upper bound material properties, but twice the nominal design fluence. From the figure it is apparent that no crack initiation would occur for this transient even with fluence levels greatly exceeding the nominal design fluence.
F.7.9 Conclusion These results demonstrate that the integrity of the St. Lucie 2 vessel would be maintained throughout theassumed life of the plant for SBLOCA + LOFW transient with recovery of feedwater, and the SBLOCA + LOFW transient where the PORV's are opened.
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8.0 CONCLUSION
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i- This Appendix to CEN-189 provides the.results of analytical evaluations of pressurized thermel shock effects on the St. Lucie 182 reactor vessels i
for cases of a SBLOCA + LOFW,'in response to the requirements' of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based on remedial actions to prevent inadequate core cooling:
r 1. SBLOCA + LOFW + PORV's opened after 10 minutes
- Themal-hydraulic system transient calculations were performed on a reference-olant basis, as reported in CEN-189 with the parameter ,
variations over the range representing all operating plants. Four different cases were analyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of each of the two different scenarios was analyzed using linear elastic fracture mechanics methods to determine the critical crack tip stress intensity values for comparison to plant specific materials properties at various times in plant life. The effect of the warm prestress phenomenon is identified where applicable for each transient, and credited where appropriate.
l In this Appendix, the results Of plant specific neutron fluence pro-file calculations are superimposed on plant specific material proper-ties to define vessel capability versus plant life. The results of l
the generic LEFM analyses were evaluated using the plant specific material properties. It is concluded that crack initiation would not occur due to the SBLOCA + LOFW transients considered, for more than i
i 32 effective full power years of operation, which is assumed to f
represent full plant life.
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