View source for RS-03-010, Supporting Documentation for Amendment to Technical Specifications Section 3.4.11, RCS Pressure and Temperature (P/T) Limits. Attachment 4, Table of Contents - Appendix F

{{Adams
| number = ML030410050
| issue date = 01/31/2003
| title = Supporting Documentation for Amendment to Technical Specifications Section 3.4.11, RCS Pressure and Temperature (P/T) Limits. Attachment 4, Table of Contents - Appendix F
| author name = Jury K
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000373, 05000374
| license number = NPF-011, NPF-018
| contact person = 
| case reference number = RS-03-010
| document report number = GE-NE-0000-0003-5526-01a, Rev 0
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Report, Technical
| page count = 125
}}

=Text=
{{#Wiki_filter:GE NuclearEnergy Engineering and Technology                          GE-NE-0000-0003-5526-01a General Electric Company                                            Revision 0 175 Curtner Avenue                                                     Class I San Jose, CA 95125                                                  June 2002 Pressure-Temperature Curves For Exelon LaSalle Unit 2 Prepared by:
L.iSenioqngineer Structural Mechanics and Materials Verified by:
B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved by:
B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Exelon and GE, Fluence Analysis, effective 11/14/01, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract.
The use of this information by anyone other than Exelon, or for any purpose other than that for which it is furnished by GE, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
Copyright, General Electric Company, 2002
                                          - iii -

GE Nuclear Energy                                              GE-NE-0000-0003-5526-01a EXECUTIVE 

==SUMMARY==

This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the
-following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of K~c of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. This report incorporates a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER
[14b], and is in compliance with Regulatory Guide 1.190.
CONCLUSIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
                     "* Closure flange region           (Region A)
                     "* Core beltline region            (Region B)
                     "* Upper vessel                    (Regions A & B)
                     "* Lower vessel                    (Regions B & C)
For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 1000F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the
                                              - iv -

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20&deg;F/hr or less must be maintained at all times.
The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beitline curves this approach has added conservatism because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature.
Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beitline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions. A composite P-T curve was also generated for the Core Critical condition at 20 EFPY.

GE Nuclear Energy                             GE-NE-0000-0003-5526-O la TABLE OF CONTENTS

==1.0    INTRODUCTION==
I 2.0    SCOPE OF THE ANALYSIS                                            3 3.0    ANALYSIS ASSUMPTIONS                                             5 4.0    ANALYSIS                                                         6 4.1   INITIAL REFERENCE TEMPERATURE                                  6 4.2   ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE                   13 4.3   PRESSURE-TEMPERATURE CURVE METHODOLOGY                        18

==5.0    CONCLUSION==
S AND RECOMMENDATIONS                                50

==6.0    REFERENCES==
65
                                 - vi -

GE Nuclear Energy                           GE-NE-0000-0003-5526-01a TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 430 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)
                              - vii -

GE Nuclear Energy                                    GE-NE-0000-0003-5526-01a TABLE OF FIGURES FIGURE 4-1    SCHEMATIC OF REACTOR VESSEL ARRANGEMENT SHOWING PLATES AND WELDS                                                        10 FIGURE 4-2. CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS           30 FIGURE 4-3. FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT           36 FIGURE 5-1:   BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]                                    53 FIGURE 5-2:  UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [20&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                                   54 FIGURE 5-3:  BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 20 EFPY
[20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       55 FIGURE 5-4:  BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 32 EFPY
[2 0 'F/HR OR LESS COOLANT HEATUP/COOLDOWN]                     56 FIGURE 5-5:  BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] []00&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                                 57 FIGURE 5-6:  UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                                 58 FIGURE 5-7:  BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 20 EFPY
[100'F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       59 FIGURE 5-8:  BELTLINE P-T CURVES FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY
[100&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       60 FIGURE 5-9:  COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 20 EFPY
[100&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       61 FIGURE 5-10: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY
[20&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                        62 FIGURE 5-11: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY
[100&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       63 FIGURE 5-12: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY
[100&deg;F/HR OR LESS COOLANT HEATUP/COOLDOWN]                       64
                                       - viii -

GE Nuclear Energy                                    GE-NE-0000-0003-5526-01a TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR LASALLE UNIT 2 VESSEL MATERIALS                   1 TABLE 4-2: RTNr VALUES FOR LASALLE UNIT 2 NOZZLE & WELD MATERIALS             12 TABLE 4-3: LASALLE UNIT 2 BELTLINE ART VALUES (20 EFPY)                       16 TABLE 4-4: LASALLE UNIT 2 BELTLINE ART VALUES (32 EFPY)                       17 TABLE 4-5: 

==SUMMARY==
OF THE 10CFRS0 APPENDIX G REQUIREMENTS                     20 TABLE 4-6: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B                                            22 TABLE 4-7: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B                                              22 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES                                                                52
                                     - ix-

GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a

==1.0       INTRODUCTION==

The pressure-temperature (P-T) curves included in this report have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittiement effects in the beltline.
Complete P-T curves were developed for 20 and 32 effective full power years (EFPY).
The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. The P-T curves incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190.
The methodology used to generate the P-T curves in this report is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640 [4]. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. P-T curves are developed using geometry of the RPV shells and discontinuities, the initial    RTNDT   of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.
The initial  RTNDT  is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the initial    RTNDT  values are tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTNDT is documented in Section 4.1.
Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [7] provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 20 and 32 EFPY are included in Section 4.2. The 32 EFPY peak ID fluence value of GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a 1.09 x 1018 n/cm2 used in this report is discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.
Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect the discontinuities.
Guidelines and requirements for operating and temperature monitoring are included in Appendix C. GE SIL 430, a GE service information letter regarding Reactor Pressure Vessel Temperature Monitoring is included in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the LPCI nozzle) are outside the beltline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).
GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a 2.0      SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curves in this report incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190. A detailed description of the P-T curve bases is included in Section 4.3. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of K1c of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. Other features presented are:
         "* Generation of separate curves for the upper vessel in addition to those generated for the beltline, and bottom head.
         "* Comprehensive description of discontinuities used to develop the non-beltline curves (see Appendix A).
The pressure-temperature (P-T) curves are established to the requirements of 10CFR50, Appendix G [8] to assure that brittle fracture of the reactor vessel is prevented. Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Rev. 2 [7].
In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable LaSalle Unit 2 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].
Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring GE Nuclear Energy                                          GE-N E-0000-0003-5526-O1 a requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the LPCI nozzle) are outside the beltline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).
GE Nuclear Energy                                          GE-NE-0000-0003-5526-01 a 3.0     ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:
For end-of-license (32 EFPY) fluence an 80% capacity factor is used to determine the EFPY for a 40-year plant life. The 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (refueling outages, etc. -20%
of the year).
The shutdown margin is calculated for a water temperature of 680 F, as defined in the LaSalle Unit 2 Technical Specification, Section 1.1.
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a 4.0     ANALYSIS 4.1     INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:
: a.      Test specimens shall be longitudinally oriented CVN specimens.
: b.      At the qualification test temperature (specified in the vessel purchase specification), no impact test result shall be less than 25 ft-lb, and the average of three test results shall be at least 30 ft-lb
: c.      Pressure tests shall be conducted at a temperature at least 60&deg;F above the qualification test temperature for the vessel materials.
The current requirements used to establish an initial    RTNDT   value are significantly different. For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section III, Subsection NB-2300 are as follows:
: a. Test specimens shall be transversely oriented (normal to the rolling direction) CVN specimens.
: b. RTNDT   Is defined as the higher of the dropweight NDT or 60&deg;F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion is met.
: c. Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest     RTNDT  of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever is greater.
I OCFR50 Appendix G [8] states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a must be supplemented in an approved manner. GE developed methods for analytically converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data in WRC Bulletin 217 [9] and from data collected to respond to NRC questions on FSAR submittals in the late 1970s. In 1994, these methods of estimating RTNDT were submitted for generic approval by the BWR Owners' Group [10], and approved by the NRC for generic use [11].
4.1.2 Values of Initial RTNDT and Lowest Service Temperature (LST)
To establish the initial RTNDT  temperatures for the LaSalle Unit 2 vessel per the current requirements, calculations were performed in accordance with the GE method for determining   RTNDT. Example   RTNDT  calculations for vessel plate, weld, HAZ, and forging, and bolting material LST are summarized in the remainder of this section.
For vessel plate material, the first step in calculating  RTNDT   is to establish the 50 ft-lb transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs [12]). For LaSalle Unit 2 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 2&deg;F per ft-lb energy difference from 50 ft-lb.
For example, for the LaSalle Unit 2 beltline plate heat C9404-2 in the lower-intermediate shell course, the lowest Charpy energy and test temperature from the CMTRs is 29 ft-lb at 400F. The estimated 50 ft-lb longitudinal test temperature is:
TSOL = 40&deg;F + [(50 - 29) ft-lb *20F/ft-lb] = 820 F The transition from longitudinal data to transverse data is made by adding 30&deg;F to the 50 ft-lb transverse test temperature; thus, for this case above, T-0T = 82 0F + 30&deg;F = 112 0 F GE Nuclear Energy                                                  GE-N E-0000-0003-5526-01 a The initial  RTNDT  is the greater of nil-ductility transition temperature (NDT) or (Tso-r 60 0F). Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is 100 F. Thus, the initial         RTNDT  for plate heat C9404-2 is 520 F.
For the LaSalle Unit 2 beltline weld heat 3P4966 with flux lot 1214 (contained in the lower-intermediate shell), the CVN results are used to calculate the initial       RTNDT. The 50 ft-lb test temperature is applicable to the weld material, but the 30&deg;F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat 3P4966 has a lowest Charpy energy of 28 ft-lb at 10&deg;F as recorded in weld qualification records. Therefore, TsoT  = 10F + [(50 - 28) ft-lb
* 2&deg;F/ft-lb] = 54&deg;F The initial  RTNDT  is the greater of nil-ductility transition temperature (NDT) or (TSoT - 60&deg;F). For LaSalle Unit 2, the dropweight testing to establish NDT was not recorded for most weld materials. GE procedure requires that, when no NDT is available for the weld, the resulting RTNDT should be -50&deg;F or higher. The value of (T5oT  - 60 0F) in this example is -6&deg;F; therefore, the initial RTNDT was -6&deg;F.
For the vessel HAZ material, the     RTNDT   is assumed to be the same as for the base material, since ASME Code weld procedure qualification test requirements and post weld heat treat data indicate this assumption is valid.
For vessel forging material, such as nozzles and closure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the feedwater nozzle at LaSalle Unit 2 (Heat Q2Q25W), the NDT is -20&deg;F and the lowest CVN data is 28 ft-lb at -200 F. The corresponding value of (Ts0T - 60 0F) is:
(T5oT  - 60&deg;F) = {[-20 + (50 - 28) ft-lb "2&deg;F/ft-lb] + 30 0 F} - 60OF = -60F.
GE Nuclear Energy                                                   GE-NE-0000-0003-5526-01a Therefore, the initial  RTNDT   is the greater of nil-ductility transition temperature (NDT) or (TsoT- 60 0F), which is -60F.
In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNOT. For the lower torus shell of LaSalle Unit 2 (Heat C9306-2), the NDT was not available and the lowest CVN data was 33 ft-lb at 40&deg;F. The corresponding value of (TS0T - 60 0F) was:
(T50T - 60 0F) = {[40 0F + (50 - 33) ft-lb
* 2&deg;F/ft-lb] + 300F} - 60&deg;F = 440 F.
Therefore, the initial RTNDT was 440 F.
For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 60&deg;F (as discussed in Section 4.3.2.3) is the LST for the bolting materials. Charpy data for the LaSalle Unit 2 closure studs do not meet the 45 ft-lb, 25 MLE requirement at 10&deg;F. Therefore, the LST for the bolting material is 70&deg;F. The highest      RTNDT   in the closure flange region is 260F, for the vessel upper shell materials. Thus, the higher of the LST and the          RTNDT +600 F is 86&deg;F, the boltup limit in the closure flange region.
The initial RTNDT  values for the LaSalle Unit 2 reactor vessel (refer to Figure 4-1 for LaSalle Unit 2 schematic) materials are listed in Tables 4-1 and 4-2. This tabulation includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves.
GE Nuclear Energy                                                       GE-NE-0000-0003-5526-0 a TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL #4 0
SHELL #3
                                                              */I         0/////////
LPCI NOZZLE TOP OF BELTLINE REGION 376.3125" TOP OF ACTIVE FUEL                                                                SHELL #2 (TAF) 366.3125"                                     ELDS
                                                              ,BF ELDAB SHELL #1 BOTTOM OF ACTIVE FUEL (BAF) 216.3125" BOTTOM OF BELTLINE REGION 206.3125" BOTTOM HEAD SUPPORT SKIRT Notes:  (1) Refer to Tables 4-1 and 4-2 for reactor vessel components and their heat identifications.
(2) See Appendix E for the definition of the beitline region.
Figure 4-1: Schematic of the LaSalle Unit 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy                                                   GE-NE-0000-0003-5526-01 a Table 4-1:        RTNDT Values for LaSalle Unit 2 Vessel Materials IPH TEST TEM,        CHARPY ENERGY JDO (Ts.r-60) DROP COMPONENT               HEAT      TEMP.                                               RTNDT (F
(FT-LB)         ("F)WEIGHT        RTND CF)                             (IF)       NDT     (F)
PLATES & FORGINGS:
Top Head & Flange:
Top Head: Torus Plate          B3269-1             10     75     72     71        -20       10 Torus Plate         B3269-2             10     50     60     50        -20       10 Dollar Plate       C9195-3             40     30     30     33         50       50 Top Head Flange                BWK-446             10     70    122    142        -20       20 Shell Flange                   BRC424              10    103    110    105       -20        10 Shell Courses:
Upper Shell                    C9678-1             10     47     66     66       -14        10 Mk-24                          A8453-1             10     44     27     42         26       26 C9507-1             10     51     53     56       -20        10 Upper Int. Shell               C9569-1            40      73     69     64         10       40 Mk-23                          C9481-2            40      46     55     62         18       40 C9602-2            40      56     45     58         20       40 Low-Int. Shell                 C9404-2            40      48     44     29         52       10 Mk-22                          C9481-1            40     103     61     85         10     -30 C9601-2            40      85     93     74         10     -30 Lower Shell                    C9425-1            40      39     43     48         32        0 Mk-21                          C9425-2            40      44     40     49         30     -30 C9434-2            40      91     58     72         10     -10 Bottom Head:                   C9306-2            40      36     38     33         44      44 C9514-2            40      50     59     71         10      40 C9621-1            40      63     64     72         10      40 C9245-1            40      61     53     62         10      40 Support Skirt:                 A8699-3            40      30     31     37         50       50 A8879-1 B          40      30     30     33         50      50 A8418-4            40      34     33     30         50       50 C9491-IB           40      77     74     68         10      40 STUDS:                                                                          LST Studs                          82552              10      44     45     40         70 Nuts                           10134-48            10     51     52     43         70
                                                      -11
                                          -  11  -

GE Nuclear Energy                                                     GE-NE-0000-0003-5526-O1 a Table 4-2:            RTNDT    Values for LaSalle Unit 2 Nozzle & Weld Materials 1
* TEST                                    DRO COMPONENT                 HEAT    TEMP.
T(OF ESI      C CKARPY   ENERGY (FT-LB)       Te)IDO Ts-0
('F)     WEIGHT DRP NDT     RTNOT
('F)
NOZZLES:
Recirculation Outlet Nozzle, N1     Q2Q32W           40    54      39   40        32        40       40 Recirculation Inlet Nozzle. N2      02Q33W           40    54      62   49        12        40       40 Q2Q25W           40    82      50    77       10        40       40 Q2Q36W           40    94      87  105        10        40       40 02Q42W           40    82      98    98       10        40       40 Steam Outlet Nozzle, N3             Q2Q30W           40    45     49    48        20        40       40 Q2032W           40    62     49     58       12        40       40 Feedwater Nozzle, N4                02033W          -20    35      37   43      -20        -20      -20 02Q25W          -20    38      35    28       -6       -20        -6 02Q29W          -20    63      52    38     -26        -20      -20 LP Core Spray Nozzle, N5            02025W          -20    40      26    36       -2       -20        -2 HP Core Spray Nozzle, N16           02029W          -20    38      42    53     -26        -20      -20 RHR/LPCI Nozzle, N6                 02036W          -20    44      37    28       -6       -20        -6 Q2042W          -20    57      49    38     -26        -20      -20 Head Spray Nozzle, N7               02Q33W          -20    64      70    93     -50        -20      -20 Vent Nozzle, N8                     02019W           40    69      65    58       10        40       40 Jet Pump Instrumentation Nozzle, N9 02026W           40    29      30    41       52        52        52 CRD Hyd. System Return Nozzle, N10  02Q23W          -10    39      30    43         0      -10          0 Drain Nozzle. N15                   265M-1          -10    55      34    42       -8         -8       -8 WELDS:
Vertical Welds:
BA, BB, BC                          3P4000           10    86      87    90     -50        -50      -50 BD, BE, BF                          3P4966           10    28      84    63       -6       -50        -6 BG. BJ, BH &BK, BM, BN              4P4784           10    71      73  73.5     -50        -50      -50 Girth Welds:
AA.AC                               3P4966           10    28      84    63       -6       -50        -6 AB                                  5P6771           10    57      55    42      -34       -50      -34 AD                                  5P6214B          10    37      54    47      -24       -50      -24 Bottom Head Assembly Welds:
DA, DB, DC, DD, DE, DF              3P4000           10    86      87    90      -50       -50      -50 Top Head Assembly Welds:
DM, DN, DP. DH, DJ, DK              3P4966            10   28      84    63        -6      -50         -6 AG                                  5P6214B           10   37      54    47      -24       -50      -24 GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a 4.2     ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (Rev 2) provides the methods for determining the ART. The Rev 2 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beitline plates and welds was made and summarized in Table 4-3 for 20 EFPY and Table 4-4 for 32 EFPY.
4.2.1 Regulatory Guide 1.99, Revision 2 (Rev 2) Methods The value of ART is computed by adding the SHIFT term for a given value of effective full power years (EFPY) to the initial RTNDT. For Rev 2, the SHIFT equation consists of two terms:
SHIFT = ARTNDT + Margin where,          ARTNDT = [CF]*f (0.28-0.10logt Margin = 2(al 2 + 0. )0.5 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = 1/4Tfluence / 1019 2
Margin = 2(cra + y2)0.5 0., = standard deviation on initial RTNDT, which is taken to be 0&deg;F.
c&= standard deviation on ARTNDT, 28OF for welds and 17&deg;F for base material, except that a, need not exceed 0.50 times the ARTNDT value.
ART = Initial RTNDT    +  SHIFT The margin term aA has constant values in Rev 2 of 170F for plate and 28*F for weld.
However, a,&need not be greater than 0.5
* ARTNDT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in GE Nuclear Energy                                            GE-N E-0000-0003-5526-01 a Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of a, is taken to be 0&deg;F for the vessel plate and weld materials.
4.2.1.1             Chemistry The vessel beltline chemistries for LaSalle Unit 2 were obtained from several sources.
The vessel plate copper values were obtained from the plate manufacturer [5a] and the nickel values were obtained from the CMTRs [12]. Submerged arc weld properties were obtained from separate evaluations [13a, 13b, and 13c]. The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively.
4.2.1.2             Fluence A LaSalle Unit 2 flux for the vessel IDwall [14a] was calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190.The flux as documented in [14] is determined for the currently licensed power of 3489 MWt using a conservative power distribution and is conservatively used from the beginning to the end of the licensing period (32 EFPY).
The peak fast flux for the RPV inner surface from Reference 14 is 1.08e9 n/cm2-s. The peak fast flux for the RPV inner surface determined from surveillance capsule flux wires removed during the outage following Fuel Cycle 6 at 6.98 EFPY and at a full power of 3323 MWt is 5.22e8 n/cm 2-s [5b]. Linearly scaling the Reference 5 flux by 1.05 to the currently licensed power of 3489 MWt results in an estimated flux of 5.48e8 n/cm&#xfd;-s.
Therefore, the Reference 14 flux bounds the flux determined from the surveillance capsule flux wire results by 197%.
The time period 32 EFPY is 1.01e9 sec, therefore the RPV IDsurface fluence is as follows: RPV ID surface fluence = 1.08e9 n/cm2 -s*1.01e9 s = 1.09e18 n/cm2 . This fluence applies to the lower-intermediate plates and welds. The fluence is adjusted for the lower plates and welds and the girth weld based upon a peak I lower shell location ratio of 0.88 (at an elevation of 277" above vessel "0"); hence the peak ID surface fluence for these components is 9.59e17 n/cm 2. Similarly, the fluence is adjusted for the GE Nuclear Energy                                              GE-NE-0000-0003-5526-01a LPCI nozzle based upon a peak / LPCI nozzle location ratio of 0.244 (at an elevation of 355" above vessel "0"and at 45&deg;. 1350, and 2250 azimuths); hence the peak IDsurface fluence used for this component is 2.66e17 n/cm2 .
4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART. For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY as discussed in Section 4.3.2.2.2. At 20 EFPY, the P-T curves are not beltline limited.
Table 4-3 lists values of beltline ART for 20 EFPY and Table 4-4 lists the values for 32 EFPY. Sections 4.3.2.2.2 and 4.3.2.2.3 provide a discussion of the limiting material.
GE Nuclear Energy                                                                                GE-NE-0000-0003-5526-01a Table 4-3: LaSalle Unit 2 Beltline ART Values (20 EFPY)
Lower-4murdIte IPlat" ad WehdiED,3=. 31 Thlickness in indhes        619             Ratio Peak/Location , 1.00              32 EFPY Peak I.D. iluence , 1.09E-18     nlom^2 32 EFPY Peak 1/4 T fluenoe     7.5E+17    nVanI2 20 EFPY Peak 114T fluence= 4.7E+17         rdan^2 Loer rlata and Welds A. M M Girth Wdd AS Thickness in hinces         8.19            Rado Peak/ Locatn = 0.88                32 EFPY Peak i.D. fluence
* 9.59E+17     nicmt2 Ele.allon - 27r                      32 EFP`Y Peak 1/4 T Ikience
* 6.6E017    n/cmn2 20 EFPYPeakl/4Tfluence.         4.10E17    n/tmi2 3.10 i nI=.
Thickeaa in inches          6.19            Ratio Peak/Location - 0244               32 EFPY Peak I.D. Iluence , 2.66E17     rVcm^2 Eilevllon -355                       32 EFPY Peak 114 T luenoe = 1.6E017       rVo2 20 EFPYPeakl4Tfluenoe=          1.10E17    n/cmW2 hnital   114T    20 EFPY                             20 EFPY 20 EFPY COMPONENT            HEAT OR HEATALOT %Cu         %m         CF     RTmr    FRuence a RTmT         a,       aA     margin   Shift    ART
                                                                             "*F    ,VcMA2      -F                           F       *F      -F PLATES:
Lower She.
21-1                C9425-2     0.120      0.510       81       30    4.1E017      21        0        11      21        43      73 21-2                C9425-1     0.120      0.510       81       32    4.1E+17       21       0        11      21        43      75 21-3                C9434-2     0.090      0.510       58        10   4.1E+17       15       0        8       15        31      41 Lower4ntenmedlate shell 22-1                C9481-1     0.110      0.500       73        10   4.7E+17       21       0        10      21        41      51 22-2                C9404-2     0.070      0.490       44       52    4.7E+17       12       0         6       12       25      77 22-3                C9601-2     0.120      0.500       81        10   4.7E+17       23       0        11      23        46      58 WELDS:
Lower Vertical BA.BB,BC              3P4000 13933  0.020      0.930       27      -. 0   4.1E+17       7        0         4       7        14      -36 Lower-lntermedlate Vertical BD.BE.BF              3P4966 / 1214 0.026      0.920       41        -6   4.7E+17       12       0         6       12       23      17 Girth AB               5P677110342   0.040      0.940       54       -34   4.1E.17       14       0         7       14       28      .6 LPCI 02036W      0.220      0.830       177       .6    1.1+E17      21       0        11      21        43      37 GE Nuclear Energy                                                                                  GE-NE-0000-0003-5526-01 a Table 4-4: LaSalle Unit 2 Beltline ART Values (32 EFPY)
Lm-.ttrmedhAt. IPlateand WeddBD.33 311 Thickness in inchmes        6.19             Ratio Peak/ Location
* 1.00              32 EFPY Peak 1.0.fluence
* 1092.18       wNcm'"2 32 EFPY Peak 14 T fluence      7.5E+17 32 EFPYPeak1/4Tfluence,         7.5E17     n(cmA2 Lower Pln andWeld BA"S3. W. Glrth WeldAN Thickness In inches         6.19            Ratio Peak/Location- 0.88                 32 EFPY Peak I.D. Ituence - 9.59E+17 Flevation - 277                       32 EFPY Peak 114 T fluence
* 8.BE+17      n/cnn'2 32 EFPYPeakl14Tfiuence.         8.6E*17    n/crrn2 1.1,C Nozzle Thicknmesin indies.         6.19            Ratio Peak/ Location - 0.244              32 EFPY Peak I.0. fluence - 2.66E+17     n/cm`2 Elevation -355"                       32 EFPY Peak 1/4 T fluence.. 1.8E+17 32 EFPY Peak 1/4 T fluence -    1.8$E17    N'cm"2 Initial   114T    32 EFPY                             32 EFPY      32 EFPY COMPONENT            HEAT OR HEAT/LOT  %Cu        %M          CF      RTa,    Fluence A RT. 7       a,       a. Margin    Shift        ART
                                                                               -F     rVCM"2       F                         "F       "F            -F PLATES:
Lower Shen 21-1                 C9425-2     0.120      0.510       81         30    6.6E+17      27       0        14       27       55            85 21-2                 C9425-1     0.120      0.510       81         32    6.6E-17      27       0        14       27       55            87 21-3                 C9434-2     0090       0.510       58         10    6.6E+17      20       0        10       20       39            49 Lower4-ntenedlate Shell 22-1                 C9481-1     0.110      0.500       73         10    7.5E&#xf7;17      28       0        13       28       53            63 22-2                 C9404-2     0.070      0.490       44         52    7.5E+17      16       0        8        16       32            84 22-3                 C9601-2     0.120      0.500       81         10    7.5E+17      29       0       15        29       59            69 WELDS:
Lower Vertical BA, BB, BC           3P4000/ 3933   0020       0930        27        -50    8.6E&#xf7;17       9       0        5         9       18           -32 Lower-hltarmedlatl vertical BO, BE. BF           3P4966    1214 0.026      0.920       41         -6    7.5E+17      15       0        7        15       30            24 GMrth AS               5P6771 /0342   0.040      0.940       54        -34    6.6E-17      18       0        9        18       37             3 LP21 02036W       0.220   10.830        177         -6    1.82.17      20       0       14        29       58      1     52 GE Nuclear Energy                                              GE-NE-0000-0003-5526-O1a 4.3      PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over its service lifetime. The ASME Code (Appendix G of Section XI of the ASME Code [6]) forms the basis for the requirements of 10CFR50 Appendix G. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
                    "* Closure flange region         (Region A)
                    "* Core beltline region          (Region B)
                    "* Upper vessel                  (Regions A & B)
                    "* Lower vessel                  (Regions B & C)
The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNDT. The remaining portion of the vessel (i.e., upper vessel, lower vessel) include shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.
For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1OOF/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20 &deg;F/hr or less must be maintained at all times.
The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, K4r, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.
The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-5:
GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a Table 4-5: Summary of the 10CFR50 Appendix G Requirements Operating Condition and Pressure                    Minimum Temperature Requirement
: 1. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A
: 1. At < 20% of preservice hydrotest              Larger of ASME Limits or of highest pressure                                     closure flange region initial RTNDT + 60 0F*
: 2. At > 20% of preservice hydrotest               Larger of ASME Limits or of highest pressure                                     closure flange region initial RTNDT + 90&deg;F I1. Normal operation (heatup and cooldown),
including anticipated operational occurrences
: a. Core not critical - Curve B
: 1. At < 20% of preservice hydrotest              Larger of ASME Limits or of highest pressure                                     closure flange region initial RTNDT + 60 0F*
: 2. At > 20% of preservice hydrotest               Larger of ASME Limits or of highest pressure                                     closure flange region initial RTNDT + 120&deg;F
: b. Core critical - Curve C
: 1. At < 20% of preservice hydrotest              Larger of ASME Limits + 40&deg;F or of a.1 pressure, with the water level within the normal range for power operation
: 2. At > 20% of preservice hydrotest               Larger of ASME Limits + 40&deg;F or of pressure                                     a.2 + 40=F or the minimum permissible temperature for the inservice system hydrostatic pressure test 60&deg;F adder     is included   by GE   as   an  additional  conservatism as discussed in Section 4.3.2.3 There are four vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the two regions in the remainder of the vessel (i.e., the upper vessel and lower vessel non-beltline regions). The closure flange region limits are controlling at lower pressures primarily because of 10CFR50 Appendix G [8]
requirements. The non-beltline and beltline region operating limits are evaluated according to procedures in IOCFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.
GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a GE PROPRIETARY INFORMATION DELETED 4.3.2 P-T Curve Methodology 4.3.2.1              Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0e17 n/cm2 ) to cause any significant shift of RTNDT (see Appendix E). Non-beltline components include nozzles, the closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations.
Detailed stress analyses of the non-beitline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 100&deg;F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWR/6 components: the feedwater nozzle (FW) and the CRD penetration (bottom GE Nuclear Energy                                                GE-N E-0000-0003-5526-01 a head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-6 and 4-7.
Table 4-6: Applicable BWR/5 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle LPCI Nozzle CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Instrumentation Nozzle Shell CRD and Bottom Head (Bonly)
Top Head Nozzles (Bonly)
Recirculation Outlet Nozzle (Bonly)
Table 4-7: Applicable BWR/5 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity Identification CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shefl**
Support Skirt**
Shroud Support Attachments" Core AP and Liquid Control Nozzle**
These discontinuities are added~to the bottom head curve discontinuity list to assure that the entire bottom head is covered, since separate bottom head P-T curves are provided to monitor the bottom head.
GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a The P-T curves for the non-beltline region were conservatively developed for a large BWR/6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for LaSalle Unit 2 as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at LaSalle Unit 2 by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.
This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.
4.3.2.1.1       PressureTest - Non-Beitline, Curve A (Using Bottom Head)
In a         finite element analysis [ ], the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K4. The evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section XI Appendix G [6] and shown below. The results of that computation were K* = 143.6 Rsi-in1 2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 84&deg;F.
GE Nuclear Energy                                                GE-N E-0000-0003-5526-O1 a The limit for the coolant temperature change rate Is 20&deg;Flhr or less.
The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on a thickness of 8.0 inches; hence, tV2 = 2.83. The resulting value obtained was:
Mm = 1.85 for     _<2 M, = 0.926 ft for 2<f <3.464 = 2.6206 Mm = 3.21 for N >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Klb is calculated from the equation in Paragraph G-2214.2 [6]:
KIm = Mm *apm   =             ksi-in"2 KIb = (2/3) Mmpb =                ksi-in"2 The total K,is therefore:
K  = 1.5 (K*m+  Kib) + Mm" (aSm + (2/3)"  Osb)  = 143.6 ksi-inl2 GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNoT) for a specific K,is based on the K, equation of Paragraph A-4200 in ASME Appendix A [17]:
(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02 (T - RTNDT)  = In [(144 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 84&deg;F The generic curve was generated by scaling 143.6 ksi-in"2 by the nominal pressures
                                                                                         .. W and calculating the associated (T -   RTNDT):
The highest  RTNDT  for the bottom head plates and welds is 440 F, as shown in Tables 4-1 and 4-2.
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a Second, the P-T curve is dependent on the calculated K,value, and the K,value is proportional to the stress and the crack depth as shown below:
K* cc c (7ra)"*                                               (4-1)
The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, i, is proportional to RI(t)f 2 . The generic curve value of R/(t)"2 , based on the generic BWR/6 bottom head dimensions, is:
Generic:         R / (t)t2 = 138 / (8)"12 = 49 inch1 2                                 (4-2)
The LaSalle Unit 2 specific bottom head dimensions are R = 126.7 inches and t =7.13 inches minimum [19], resulting in:
LaSalle Unit 2 specific:      R / (t)'12 = 126.7 / (7.13)12 = 47.5 inch"r              (4-3)
                                             -26   -

GE Nuclear Energy                                               GE-NE-0000-0003-5526-01a Since the generic value of R/(t)'" is larger, the generic P-T curve is conservative when applied to the LaSalle Unit 2 bottom head.
4.3.2.1.2        Core Not CriticalHeatup/Cooldown- Non-Beltilne Curve B (Using Bottom Head)
As discussed previously, the CRD penetration region limits were established primarily for consideration of bottom head discontinuity stresses during pressure testing.
Heatup/cooldown limits were calculated by increasing the safety factor in the pressure testing stresses (Section 4.3.2.1.1) from 1.5 to 2.0.
The calculated value of K,for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the K4 value for the core not critical condition is (143.6 / 1.5)
* 2.0 = 191.5 ksi-in"2 .
GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a h..
Therefore, the method to solve for (T - RTNDT) for a specific K* is based on the KI, equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:
(T - RTNDT) = In [(14 - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT)   = 102&deg;F The generic curve was generated by scaling 192 ksi-in'" by the nominal pressures and calculating the associated (T - RTNDT):
Core Not Critical CRD Penetration K,and (T - RTNDT) as a Function of Pressure Nominal Pressure                    !(              T - RTNDT (psig)                  (ksi-in" 2)            (-F) 1563                      192                 102 1400                      172                  95 1200                      147                  85 1000                      123                  73 800                        98                  57 600                        74                  33 400                        49                 -14 The highest   RTNDT  for the bottom head plates and welds is 440 F, as shown in Tables 4-1 and 4-2.
As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Tables 4-6, 4-7, and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe
                                               -28  -

GE Nuclear Energy                                         GE-NE-0000-0003-5526-01 a than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.
GE Nuclear Energy                                 GE-NE-0000-0003-5526-01a Figure 4-2. CRD Penetration Fracture Toughness Limiting Transients GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a 4.3.2.1.3        PressureTest - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)
The stress intensity factor, KI, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was K,= 200 ksi-in r2 for an applied pressure of 1563 psig preservice hydrotest pressure.
The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the corner thickness.
To evaluate the results, K,is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or XI). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of K1 is shown below using the BWR/6, 251-inch dimensions:
Vessel Radius, R,                126.7 inches Vessel Thickness, t,             6.1875 inches Vessel Pressure, P,              1563 psig Pressure stress: a = PR / t = 1563 psig. 126.7 inches / (6.1875 inches) = 32,005 psi.
The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.97 ksi. The factor F (a/rn) from Figure A5-1 of WRC-1 75 is 1.4 where:
a = %(t,    2 + tv 2)1/2                       =2.36 inches t, = thickness of nozzle                        = 7.125 inches t, = thickness of vessel                       = 6.1875 inches r, = apparent radius of nozzle                   = ri + 0.29 rc=7.09 inches r1 = actual inner radius of nozzle              = 6.0 inches rc = nozzle radius (nozzle corner radius)        = 3.75 inches GE Nuclear Energy                                              GE-N E-0000-0003-5526-01 a Thus, airm = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an aift of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (na)1 2 . F(a/rn):
2 Nominal K,= 1.5 34.97. (x. 2.36)12 .1.4 = 200 ksi-in1 The method to solve for (T - RTNDT) for a specific K,is based on the Kk, equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:
(T - RTNDT) = In [(K1- 33.2) / 20.734] / 0.02 (T - RTNoT) = In [(200 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 104.2&deg;F The generic pressure test P-T curve was generated by scaling 200 ksi-in' 2 by the nominal pressures and calculating the associated (T - RTNDT),
GE Nuclear Energy                                            GE-NE-0000-0003-5526-01a The highest RTNDT for the feedwater nozzle materials is 40&deg;F as described below. The generic pressure test P-T curve is applied to the LaSalle Unit 2 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 400 F.
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01a Second, the P-T curve is dependent on the K,value calculated. The LaSalle Unit 2 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and KIare shown below:
Vessel Radius, Rv                    126.7 inches Vessel Thickness, t,                 6.19 inches Vessel Pressure, P,                    1563 psig Pressure stress: a = PR / t = 1563 psig. 126.7 inches / (6.19 inches) = 31,992 psi. The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.96 ksi. The factor F (a/r) from Figure A5-1 of WRC-175 is determined where:
a    I=(t    2 + tv 2 )112                      =2.36 inches tn = thickness of nozzle                          = 7.125 inches t, = thickness of vessel                         = 6.19 inches rn = apparent radius of nozzle                    = r1 + 0.29 r,=6.8 inches ri =  actual inner radius of nozzle              = 6.0 inches r, =   nozzle radius (nozzle comer radius)        = 2.75 inches Thus, alrn = 2.36 / 6.8 = 0.35. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an air, of 0.35, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (na)7". F(a/rn):
Nominal K,= 1.5- 34.96 - (7 - 2.36)"2 .1.4 = 199.9 ksi-in" 2 GE Nuclear Energy                                            GE-NE-0000-0003-5526-01 a 4.3.2.1.4        Core Not CriticalHeatup/Cooldown- Non-Beitline Curve B (Using FeedwaterNozzle/Upper Vessel Region)
The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.
Stresses were taken from a                   finite element analysis done specifically for the purpose of fracture toughness analysis [ ]. Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40&deg;F feedwater injection, which is equivalent to hot standby, see Figure 4-3.
The non-beltline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (WRC)
Bulletin 175 [15].
The stress intensity factor for a nozzle flaw under primary stress conditions (Kip) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:
Kip = SF .a (7a)1/2
* F(a/rJ)                                                  (4-4) where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/rQ) is the shape correction factor.
GE Nuclear Energy                                         GE-NE-0000-0003-5526-01 a Figure 4-3. Feedwater Nozzle Fracture Toughness Limiting Transient Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rn) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [15].
The stresses used in Equation 4-4 were taken from             design stress reports for the feedwater nozzle. The stresses considered are primary membrane, a.r, and primary bending, ap. Secondary membrane, am, and secondary bending, ao1 , stresses are included in the total KI by using ASME Appendix G [6] methods for secondary portion, Ks:
KI, = Mm (asm + (2/3)" asb)                                              (4-5)
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].
However, the correction was not applied to primary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. KIp and K1, are added to obtain the total value of stress intensity factor, K1. A safety factor of 2.0 is applied to primary stresses for core not critical heatup/cooldown conditions.
Once KIwas calculated, the following relationship was used to determine (T - RTNDT).
The method to solve for (T - RTNDT) for a specific K,is based on the Kc equation of Paragraph A-4200 in ASME Appendix A [17]. The highest            RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.
(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02                       (4-6)
Example Core Not Critical HeatuplCooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the feedwater nozzle               analysis, where feedwater injection of 40&deg;F into the vessel while at operating conditions (551.4&deg;F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle corner stresses were obtained from finite element analysis [ ]. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation.
However, a thickness of 7.5 inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (apm) was adjusted for the actual vessel thickness of 6.1875 inches (i.e., 0 pm = 20.49 ksi was revised to 20.49 ksi
* 7.5 inches/6.1875 inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:
apm = 24.84 ksi           as, = 16.19 ksi         a   = 45.0 ksi        tv = 6.1875 inches apb   = 0.22 ksi          a*b = 19.04 ksi         a = 2.36 inches       r, = 7.09 inches t, = 7.125 inches GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a In this case the total stress, 60.29 ksi, exceeds the yield stress, ac,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root is the inside surface temperature.)
R = [ay - om + ((ototai - cy) / 30)] / (aOota - apr)                                   (4-7)
For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for om. The resulting stresses are:
Crpm = 24.84 ksi         asm = 9.44 ksi apb = 0.13 ksi           aCb    11.10 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on the 4a thickness; hence, t"2 = 3.072. The resulting value obtained was:
Mm = 1.85 for N/tV<2 Mm  = 0.926 fti for 2<4:<3.464 = 2.845 Mm = 3.21 for ft">3.464 The value F(a/r,), taken from Figure A5-1 of WRC Bulletin 175 for an aft, of 0.33, is therefore, F (a / rn) = 1.4 Kip is calculated from Equation 4-4:
Kip = 2.0 (24.84 + 0.13) . (7 . 2.36)'2 . 1.4 Kip = 190.4 ksi-in'12 GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a K15 is calculated from Equation 4-5:
Kis = 2.845. (9.44 + 2/3- 11.10) 2 K1 s = 47.9 ksi-in 1 The total K,is, therefore, 238.3 ksi-in1'2.
The total K4 is substituted into Equation 4-6 to solve for (T - RTNDT):
(T - RTNDT) = In [(238.3- 33.2) / 20.7341 / 0.02 (T- RTNDT)=     115&deg;F The            curve was generated by scaling the stresses used to determine the K1; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40&deg;F water injected into the hot reactor vessel nozzle. In the base case that yielded a K,value of 238 ksi-in'12, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4&deg;F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tswraon - 40) / (551.4 - 40). From K, the associated (T - RTNDT) can be calculated:
GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a Core Not Critical Feedwater Nozzle K,and (T -      RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp.                  R            K*           (T - RTNDT)
(psig)                   (OF)                        (ksi-in'")          (OF) 1563                    604             0.23           303              128 1400                    588             0.34           283              124 1200                    557             0.48           257              119 1050                    551             0.58           238              115 1000                    546             0.62           232              113 800                    520             0.79           206              106 600                     489              1.0           181               98 400                     448              1.0           138               81
*Note: For each change in stress for each pressure and saturation temperature condition, there is a corresponding change to R that influences the determination of K4.
The highest non-beltline   RTNDT  for the feedwater nozzle at LaSalle Unit 2 is 40&deg;F as shown in Tables 4-1 and 4-2 and previously discussed. The jet pump instrumentation nozzle is not limiting, as previously discussed. The generic curve is applied to the LaSalle Unit 2 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT  value of 40&deg;F as discussed in Section 4.3.2.1.3.
4.3.2.2          CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.
GE Nuclear Energy                                              GE-NE-0000-0003-5526-01a The stress intensity factors (K1), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through wall temperature distribution of a flat plate; values were calculated for 100&deg;F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.
4.3.2.2.1        Beitline Region - PressureTest The methods of ASME Code Section XI, Appendix G [6] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (trn) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:
c'm = PR / trin                                                                (4-8)
The stress intensity factor, Km, is calculated using Paragraph G-2214.1 of the ASME Code.
The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIc, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.
The relationship between Kic and temperature relative to reference temperature (T - RTNDT) is based on the K1c equation of Paragraph A-4200 in ASME Appendix A [17]
for the pressure test condition:
Kim SF = Kic = 20.734 exp[0.02 (T - RTNDT)] + 33.2                            (4-9)
GE Nuclear Energy                                              GE-NE-0000-0003-5526-01a This relationship provides values of pressure versus temperature (from        KIR and (T-RTNDT), respectively).
GE's current practice for the pressure test curve is to add a stress intensity factor, Kt, for a coolant heatup/cooldown rate of 20 &deg;F/hr to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatup/cooldown rate of 100&deg;F/hr. The K1t calculation for a coolant heatup/cooldown rate of 100lF/hr is described in Section 4.3.2.2.3 below.
4.3.2.2.2        Calculationsfor the Beitline Region - PressureTest This sample calculation is for a pressure test pressure of 1105 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:
Adjusted  RTNDT =   Initial RTNDT + Shift            A = 32 + 55 = 87&deg;F (Based on ART values in Section 4.2)
Vessel Height                                        H = 870.5 inches Bottom of Active Fuel Height                         B = 216.3 inches Vessel Radius (to inside of clad)                    R = 126.5 inches Minimum Vessel Thickness (without clad)              t = 6.19 inches Pressure is calculated to include hydrostatic pressure for a full vessel:
P = 1105 psi + (H - B) 0.0361 psi/inch = P psig                                        (4-10)
    = 1105 + (870.5 - 216.3) 0.0361 = 1129 psig Pressure stress:
a  =  PR/t                                                                             (4-11)
    =  1.129 - 126.5 / 6.19 = 23.1 ksi GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on a thickness of 6.19 inches (the minimum thickness without cladding);
hence, t" 2 = 2.49. The resulting value obtained was:
Mm = 1.85 for ftV<2 am = 0.926 f-"for 2< Ft<3.464 = 2.30 Mm = 3.21 for ft >3.464 The stress intensity factor for the pressure stress is Kim = Mm " a. The stress intensity factor for the thermal stress, K,, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20&deg;F/hr instead of 100&deg;F/hr.
Equation 4-9 can be rearranged, and 1.5 Kim substituted for Kic, to solve for (T - RTNDT).
Using the K1c equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 53.1, and Kit = 2.58 for a 20&deg;F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:
(T - RTNDT)    = ln[(1.5. Kim + Kit - 33.2) / 20.734] / 0.02                 (4-12)
                        = ln[(1.5. 53.1 + 2.58 - 33.2) / 20.734] / 0.02
                        = 43.0&deg;F T can be calculated by adding the adjusted RTNDT:
T = 43.0 + 87 = 130&deg;F                  for P = 1105 psig For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY. The beltline pressure test P-T curves provided in Section 5.0 of this report are calculated in the same manner as the Feedwater Nozzle pressure test P-T curves as described in Section 4.3.2.1.3. The initial RTNDT for the LPCI nozzle materials is -6 0 F as shown in Table 4-2. The generic pressure test P-T curve is applied to the LaSalle Unit 2 Feedwater Nozzle curve by shifting the P vs. (T - RTNDT) values in Section 4.3.2.1.3 to reflect the ART value of 52 0 F. The 20 EFPY beltline pressure test
                                           -43  -

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01a P-T curves are non-beitline limited and the beltline material calculations are performed as described in this section.
4.3.2.2.3         Beltline Region - Core Not CriticalHeatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section XI Appendix G [6]:
KIc = 2.0- KIm +K1 t                                                        (4-13) where Kim is primary membrane Kdue to pressure and KIt is radial thermal gradient K due to heatup/cooldown.
The pressure stress intensity factor KIm is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.
The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-1 of ASME Appendix G [6] by the through-wall temperature gradient ATw, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [6]. The relationship used to compute the through-wall ATf is based on one-dimensional heat conduction through an insulated flat plate:
a 2T(x,t) / a x2 = 1 / p (aT(x,t) / 8t)                                     (4-14) where T(x,t) is temperature of the plate at depth x and time t, and 13 is the thermal diffusivity.
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that 8l(x,t) / at = dT(t) / dt = G, where G is the coolant heatup/cooldown rate, normally 100OF/hr. The differential equation is integrated over x for the following boundary conditions:
: 1. Vessel inside surface (x = 0) temperature is the same as coolant temperature, To.
: 2. Vessel outside surface (x = C) is perfectly insulated; the thermal gradient dT/dx = 0.
The integrated solution results in the following relationship for wall temperature:
T = Gx 2 / 20 - GCx / 3 + To                                                  (4-15)
This equation is normalized to plot (T - TO) / ATw versus x / C.
The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [6]. Therefore, AT, calculated from Equation 4-15 is used with the appropriate Mt of Figure G-2214-1 of ASME Appendix G [6] to compute Kt for heatup and cooldown.
The Mt relationships were derived in the Welding Research Council (WRC)
Bulletin 175 [15] for infinitely long cracks of 1/4T and 1/8T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.
For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY. The beltline core not critical P-T curves provided in Section 5.0 of this report are calculated in the same manner as the Feedwater Nozzle core not critical P-T curves as described in Section 4.3.2.1.4. The initial RTNDT for the LPCI nozzle materials is -6&deg;F as shown in Table 4-2. The generic core not critical P-T curve is applied to the LaSalle Unit 2 Feedwater Nozzle curve by shifting the P vs. (T - RTNDT) values in Section 4.3.2.1.4 to reflect the ART value of 52&deg;F. The 20 EFPY beltline core not critical P-T curves are non-beltline limited and the beltline material calculations are performed as described in this section.
                                            -45   -

GE Nuclear Energy                                              GE-NE-0000-0003-5526-01a 4.3.2.2.4       Calculationsfor the Beltine Region Core Not Critical Heatup/Cooldown This sample calculation is for a pressure of 1105 psig for 32 EFPY. The core not critical heatup/cooldown curve at 1105 psig uses the same Kim as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. Inaddition, there is a K,term for the thermal stress.
The additional inputs used to calculate Kit are:
Coolant heatup/cooldown rate, normally 100&deg;F/hr               G = 100 &deg;F/hr Minimum vessel thickness, including clad thickness            C = 0.552 ft (6.625 inches)
(the maximum vessel thickness is conservatively used)
Thermal diffusivity at 550&deg;F (most conservative value)        13 = 0.354 ftl/ hr [21]
Equation 4-15 can be solved for the through-wall temperature (x = C), resulting in the absolute value of AT for heatup or cooldown of:
AT = GC 2 / 21                                                                      (4-16)
                = 100. (0.552)/     (2 0.354) = 43&deg;F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.30) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kit = Mt- AT = 12.9, can be calculated. Kim has the same value as that calculated in Section 4.3.2.2.2.
The pressure and thermal stress terms are substituted into Equation 4-9 to solve for (T - RTNDT):
(T - RTNDT)     =    ln[((2* Kim + Kit)- 33.2)/20.734]/0.2                    (4-17)
                        =    In[(2
* 53.1 + 12.9- 33.2) / 20.734] /0.02
                        =    71.1 OF GE Nuclear Energy                                           GE-N E-0000-0003-5526-01 a T can be calculated by adding the adjusted   RTNDT:
T=71.1+87=158.1&deg;F             forP= 1105psig 4.3.2.3          CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region   RTNDT. In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves.
However, some closure flange requirements do impact the curves, as is true with LaSalle Unit 2 at low pressures.
The approach used for LaSalle Unit 2 for the bolt-up temperature was based on a conservative value of (RTNDT + 60), or the LST of the bolting materials, whichever is greater. The 60&deg;F adder is included by GE for two reasons: 1) the pre-1971 requirements of the ASME Code Section III, Subsection NA, Appendix G included the 60&deg;F adder, and 2) inclusion of the additional 60 0 F requirement above the RTNDT provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 and 4-2, the limiting initial RTNDT for the closure flange region is represented by both the top head and vessel shell flange materials at 26&deg;F, and the LST of the closure studs is 70&deg;F; therefore, the bolt-up temperature value used is 86&deg;F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.
10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT   of the closure region. Curve A temperature must be no less than     (RTNDT  + 900F) and Curve B temperature no less than (RTNDT + 1200 F).
GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a For pressures below 20% of preservice hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at    RTNDT  or greater as described above. At low pressure, the ASME Code [6] allows the bottom head regions to experience even lower metal temperatures than the flange region RTNDT. However, temperatures should not be permitted to be lower than 68&deg;F for the reason discussed below.
The shutdown margin, provided in the LaSalle Unit 2 Technical Specification, is calculated for a water temperature of 68&deg;F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68&deg;F limit, further extensive calculations would be required to justify a lower temperature. The 86&deg;F limit for the upper vessel and beltline region and the 68&deg;F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel.
When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.
4.3.2.4         CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8], Table 1. Table I of [8] requires that core critical P-T limits be 40&deg;F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40&deg;F for pressures above 312 psig.
Table I of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60&deg;F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 86&deg;F, based on an        RTNDT  of 26&deg;F. In GE Nuclear Energy                                          GE-NE-0000-0003-5526-01 a addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160&deg;F or the temperature required for the hydrostatic pressure test (Curve A at 1105 psig). The requirement of closure region RTNDT + 160&deg;F does cause a temperature shift in Curve C at 312 psig.
GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a

==5.0      CONCLUSION==
S AND RECOMMENDATIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.
There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:
* Closure flange region         (Region A) 0   Core beltline region          (Region B)
* Upper vessel                  (Regions A & B)
* Lower vessel                  (Regions B & C)
For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of IOO0 F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20F/hr or less must be maintained at all times.
The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 314T for a given metal temperature.
GE Nuclear Energy                                            GE-NE-0000-0003-5526-01a The following P-T curves were generated for LaSalle Unit 2.
"* Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.
"  Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.
"  A composite P-T curve was also generated for the Core Critical condition at 20 and 32 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.
Using the flux from Reference 14 the P-T curves are not beltline limited through 1400 psig for curve A and curve B for 20 EFPY. The P-T curves are beltline (LPCI nozzle) limited above 760 psig for curve A and 550 psig for curve B for 32 EFPY.
Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves is presented in Appendix B.
GE Nuclear Energy                                                GE-NE-0000-0003-5526-01 a Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure          Table Numbers Curve                Curve Description                   Numbers for                 for Presentation of       Presentation of
_the                             P-T Curves       the P-T Curves Curve A Bottom Head Limits (CRD Nozzle)                  Figure  5-1          B-1 & B-3 Upper Vessel Limits (FW Nozzle)                  Figure  5-2          B-1 & B-3 Beltline Limits for 20 EFPY                      Figure  5-3              B-3 Beltline Limits for 32 EFPY                      Figure  5-4              B-1 Curve B    I Bottom Head Limits (CRD Nozzle)                  Figure  5-5           B-1 & B-3 Upper Vessel Limits (FW Nozzle)                  Figure  5-6           B-1 & B-3 Beitline Limits for 20 EFPY                      Figure  5-7              B-3 Beltline Limits for 32 EFPY                      Figure  5-8              B-1 Curve C Composite Curve for 20 EFPY**                    Figure 5-9               B-4 A, B, & C    Composite Curves for 32 EFPY Bottom Head and Composite Curve A                Figure 5-10              B-2 for 32 EFPY*
Bottom Head and Composite Curve B                Figure 5-11              B-2 for 32 EFPY*
Composite Curve C for 32 EFPY**                  Figure 5-12              B-2
* The Composite Curve A        &  B curve is the more limiting of three limits: 10CFR50   Bolt up Limits, Upper Vessel Limits (FW Nozzle), and Beltline Limits. A separate Bottom Head Limits (CRD Nozzle) curve is individually included on this figure.
**  The Composite Curve C curve is the more limiting of four limits: 10CFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Limits (FW Nozzle), and Beltline Limits.
GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 1000 z
: 9. 900 0                                                                  iINITIAL RTndt VALUE ISI 49LF FOR BOTOM HEADJ 0    800 0700                                                             HEATUP/COOLDOWN RATE OF COOLANT
                                                                           < 20"F/HR
* 600
* 500 0
ILl 300 200 100 0
0     25     50      75    100    125   150    175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve A]
[20 &deg;F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                                   GE-NE-0000-0003-5526-O1a 1400 1300 1200 1100 1000 0    9800                                        '   'IINITIAL                    RTndt VALUE IS I
_______40_F      FOR UPPER VESSEL w
U) 800 U)
U) 0 700 HEATUP/COOLDOWN Uj                                                                           RATE OF COOLANT S600           -----                                                            20"F/HR S500 f    400 300 200                                    -FLANGE  REGION 16F
                                                                            -UPPER VESSEL LIMITS (Including 100                                                                       Flange and FW Nozzle Limbt) 0 0     25      50      75     100      125 150     175       200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-2: Upper Vessel P-T Curve for Pressure Test [Curve A]
[20 0 F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a 1400 1300 1200 INITIAL RTndt VALUE IS 52&deg;F FOR BELTLINE 1100 C.
a   1000 R
 -I   900
 -I BELTLINE CURVE to   800                                                           ADJUSTED AS SHOWN:
EFPY SHIFT (F) 20       25 700 U)                                                                 HEATUP/COOLDOWN 600                                                            RATE OF COOLANT
                                                                           < 200F/HR
  ,I 500 400 uJ 300 200
                                                                  "-BELTLINELIMITS 100 0
0     25     50     75     100   125    150   175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 20 EFPY
[201F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                             GE-NE-O000-0003-5526-01 a 1400 1300 1200                                                            INITIAL RTndt VALUE IS
                                                                       -6&deg;F FOR LPCI NOZZLE 1100                                                          Note: The LPCI Nozzle is the limiting material for the beltline region.
IL 900 BELTLINE CURVE ADJUSTED AS SHOWN:
10 EFPY SHIFT (&deg;F) o)
U 700                                                                     32         58 Uj I-1                                                                    HEATUP/COOLDOWN 1% 600
  -I                                                                      RATE OF COOLANT Z                                                                              < 20F/IHR
_j 500 w
      ~400 w
300 200                                                           -BELTLINE           LIMITS 100 0
0     25     50     75     100   125   150   175    200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-4: Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY
[20&deg;F/hr or less coolant heatup/cooldown]
                                            -56  -

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 0.. 1000 a.
  $A 900                                                          INITIAL RTndt VALUE IS I                                                                 49&deg;F FOR BOTTOM HEAD w
0     800 700 600 500 w
: 0)    400 300 200 100 0
0     25     50    75    100    125    150   175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (&deg;F)
Figure 5-5: Bottom Head P-T Curve for Core Not Critical [Curve B]
[100IF/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                                     GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 1000 ILu x
 "IL 900 0
I.                                                                        40"F FOR UPPER VESSELI
 .1                                                                        I INITIAL RTndt VALUE IS I
   .Il "U) 0 LU 800 w    o700                                                                    HEATUP/COOLDOWN RATE OF COOLANT I                                                                                  < 100&deg;FIHR
      , 600 500 F I.
LU CVn400 (n
300          3712 PSI G       ___
I                   a 200
                                                                !~             *-UPPER VESSEL I      i                 LIMITS (Including 100                                    I   FLANGE REGION 86&deg;F             Flange and FW I                        Nozzle Umits) 0 0     25       50       75    100     125   150    175     200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B]
[1 00&deg;F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                                  GE-N E-0000-0003-5526-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 52&deg;F FOR BELTLINE 1100
* F a 1000                                                                     BELTLINE CURVE 9(
z                                                                        ADJUSTED AS SHOWN:
I
  -j
      ="900 I                  j     W EFPY 20 SHIFT (&deg;F) 25 C
1,1   800 O     700 I_    _
HEATUP/COOLDOWN 31     SIG                                           RATE OF COOLANT
* 600                                                                     < 100&deg;FIHR
_   _     _ _  I, _
    -500 LU                                              1QF5
________J
___   _ __ILU uo    400 U) 3U                      l                  86F 300 200
                                                                         -BELTLINE       LIMITS 100 0
0      25         50     75    100     125 150    175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-7: Beltline P-T Curve for Core Not Critical [Curve B] up to 20 EFPY
[100&deg;F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a 1400 1300 I                        I 1200                 T                                               INITIAL RTndt VALUE IS II                      I-6=F                    FOR LPCI NOZZLE 1100                 +     [                                      Note: The LPCI Nozzle is the limiting material for the beitline region.
1000 a.
0    900  ,     t                    i                                       BELTLINE CURVE ADJUSTED AS SHOWN:
0                                                                            EFPY       SHIFT (-F) 800                                                                         32         58 700        r I-l w                                    Ii                                  HEATUP/COOLDOWN RATE OF COOLANT u'J                                                                              < 100&deg;F/HR 600 I-l 500                         _              -         _
400   _       _                              -      -      -
0, a.
300   _     j-31--S 200 "rl
                                               ,RS0                      -      BELTLINE LIMITS 100        T        F                    F 0     25       50   75    100     125      150     175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-8: Beltline P-T Curves for Core Not Critical [Curve B] up to 32 EFPY
[1OO0 F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                               GE-NE-0000-0003-5526-01 a 1400 INITIAL RTndt VALUES ARE 1300                                                              52&deg;F FOR BELTLINE, 40&deg;F FOR UPPER VESSEL, 1200                                                                        AND 49&deg;F FOR BOTTOM HEAD 1100
     =31000                                                                BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (&deg;F)
: 0. 900                                                                  20         25 I     800 HEATUP/COOLDOWN RATE OF COOLANT
_<100&deg;F/HR 600 w
300 600 2
SBELTLINE AND 100                                                                  NON-BELTLINE LIMITS 0
0    25   50    75    100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(&deg;F)
Figure 5-9: Composite Core Critical P-T Curves [Curve C] up to 20 EFPY
[100 0 F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a 1400 1300 1200 INITIAL RTndt VALUES ARE
                                                                    -6&deg;F FOR LPCI NOZZLE, 1100                                                        40"F FOR UPPER VESSEL, AND 49&deg;F FOR BOTTOM HEAD S1000 Note: The LPCI Nozzle is the limiting material for the 0    900                                                                 beitline region.
800 I
V-.
0O 700 BELTLINE CURVES EOS600                                                            ADJUSTED AS SHOWN:
EFPY SHIFT (&deg;F)
S500                                                                        32      58 w
o 400                                                             HEATUP/COOLDOWN U                                                                   RATE OF COOLANT
                                                                             < 20F/HR 300
                                                                    -UPPER              VESSEL 200                                                                    AND BELTLINE LIMITS
                                                                     ...... BOTTOM HEAD 100                                                                    CURVE 0
0     25    50      75    100    125   150   175   200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-10: Composite Pressure Test P-T Curves [Curve A] up to 32 EFPY
[20 &deg;F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                                                  GE-NE-0000-0003-5526-01 a 1400       -.                             -
1300              -~        _                               ."_':_
1200                               r                                                 INITIAL RTndt VALUES I/                                       ARE
_                           "-6&deg;FFOR LPCI NOZZLE, 1100                      t                                                              409F FOR UPPER S                                                                             VESSEL,
_   ____    __    __     I___
____I          IAND "01000                                                                                  49F FOR BOTTOM HEAD
                      -0                                                   i              Note: The LPCI Nozzle is S900   -                         tthe                                                 limiting material for the beltline region.
cao 80 0     _           __
0     700                                           _                                         BELTLINE CURVES i                   IADJUSTED                                                                            AS SHOWN:
oi[                            IEFPY                                                                      SHIFT (F) a: 600                                     .."                               I                      32       58
 -j0                      HEAD 68T                                                              HEATUP/COOLDOWN RATE OF COOLANT 1a_400       _
100&deg;F/HR 300 200                      [                   FAND
                                                                                             ._UPPER LIMITS VESSEL BELTLINE 100             __                                REGION                             ...... BOTTOM HEAD SIO                                       CURVE 0      25       50        75         100       125        150   175     200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(&deg;F)
Figure 5-11: Composite Core Not Critical P-T Curves [Curve B] up to 32 EFPY
[1 00&deg;F/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                           GE-NE-O000-0003-5526-01 a 1400 1300                                                          INITIAL RTndt VALUES ARE
                                                                   -6&deg;F FOR LPCI NOZZLE, 1200 40"F FOR UPPER VESSEL, AND 1100                                                        49"F FOR BOTTOM HEAD "1000                                                        Note: The LPCI Nozzle is 4                                                               the limiting material for the uJ                                                                     beltline region.
M. 900 0
iI
  -J
  'f   800 o 700 w
BELTLINE CURVE ADJUSTED AS SHOWN:
4                                                                 EFPY SHIFT (&deg;F)
Uj   5 00                                                              32           58 w
S4500                                                              HEATUP/COOLDOWN RATE OF COOLANT
: u.                                                                       < 100&deg;F/HR a.
300 200 I--BELTLINED NON-BELTLINE 100                                                               LIMITS 0
0   25   50    75  100   125   150 175  200  225  250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
Figure 5-12: Composite Core Critical P-T Curves [Curve C] up to 32 EFPY
[100OF/hr or less coolant heatup/cooldown]
GE Nuclear Energy                                         GE-N E-0000-0003-5526-01 a

==6.0    REFERENCES==
: 1. Carey, R.G., "Pressure-Temperature Curves for ComEd LaSalle Unit 2," GE-NE, San Jose, CA, May 2000, (GE-NE-B13-02057-00-05Rl, Revision 1)(GE Proprietary).
: 2. GE Drawing Number 761 E581, "Reactor Vessel Thermal Cycles," GE-NED, San Jose, CA, Revision 1 (GE Proprietary).
: 3. GE Drawing Number 15888136, "Reactor Vessel Nozzle Thermal Cycles,"
GE-NED, San Jose, CA, Revision 6 (GE Proprietary).
: 4.  "Alternative Reference Fracture Toughness for Development of P-T Limit Curves Section XI, Division 1," Code Case N-640 of the ASME Boiler & Pressure Vessel Code, Approval Date February 26, 1999.
: 5. a) T.A. Caine, "LaSalle County Station Units 1 and 2 Fracture Toughness Analysis per 10CFR50 Appendix G", GE-NE, San Jose, CA, March 1988 (SASR 88-10).
9 b)  E.W. Sleight, "LaSalle Unit 2 RPV surveillance Materials Testing and Analysis,"
GE-NE, San Jose, CA, February 1996, (GE-NE-B1301786-01, Revision 0).
: 6.  "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section III or XI of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.
: 7.  "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.
: 8.  "Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.
: 9. Hodge, J. M., "Properties of Heavy Section Nuclear Reactor Steels," Welding Research Council Bulletin 217, July 1976.
GE Nuclear Energy                                          GE-NE-0000-0003-5526-01 a
: 10. GE Nuclear Energy, NEDC-32399-P, "Basis for GE RTNDT Estimation Method,"
Report for BWR Owners' Group, San Jose, California, September 1994 (GE Proprietary).
: 11. Letter from B. Sheron to R.A. Pinelli,"Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994, " USNRC, December 16, 1994.
: 12. QA Records & RPV CMTR's:
LaSalle Unit 2 -QA Records & RPV CMTR's LaSalle Unit 2 GE PO# 205-AE020, Manufactured by CBIN.
: 13. a) Letter from L. Loflin (Shearon Harris Nuclear Power) to NRC dated September 8, 1989, transmitting BAW-2083, "Analysis of Capsule U, Carolina Power & Light Company, Shearon Harris Unit No. 1, Reactor Vessel Material Surveillance Program", August 1989.
b) "Carolina Power & Light Company, Shearon Harris Unit No. 1, Reactor Vessel Radiation Surveillance Program", WCAP-10502, May 1984.
c) Letter from R. M. Krich to the NRC, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity - Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 - LaSalle County Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374 - Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265," Commonwealth Edison Company, Downers Grove, IL., July 30, 1998.
GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a
: 14. a) Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation," GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).
b) Letter, S.A. Richards, USNRC to J.F. Klapproth, GE-NE, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)", MFN 01-050, September 14, 2001.
: 15. "PVRC Recommendations on Toughness Requirements for Ferritic Materials,"
Welding Research Council Bulletin 175, August 1972.
16.
: 17. "Analysis of Flaws," Appendix A to Section XI of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.
18.
: 19. Bottom Head and Feedwater Nozzle Dimensions:
a) CBIN Drawing, GE Number VPF 3073-1-7, "Vessel Outline," GE-APED, San Jose, CA, Revision 7.
b) GE Drawing Number VPF 3073-52, "Feedwater Nozzle", GE-NED, San Jose, CA, Revision 7.
20.
: 21. "Materials - Properties," Part D to Section IIof the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.
                                         -67  -

GE Nuclear Energy                       GE-NE-0000-0003-5526-01 a APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy     GE-NE-0000-0003-5526-01a A-2

GE Nuclear Energy                                              GE-N E-0000-0003-5526-01 a Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5" or less provided the lowest service temperature is not lower than RTNDT plus 60 0F. Nozzles and appurtenances made from Alloy 600 (Inconel) do not require fracture toughness analysis. Components that do not require a fracture toughness evaluation are listed below:
Nozzle or Appurtenance     Nozzle or Appurtenance          Material   Reference                 Remarks Identification Nl1              Core Differential Pressure &    Alloy 600               Thickness is < 2.5" and made of Liquid Poison - Penetration                              Alloy 600; therefore, no further
                 < 2.5"                                                  fracture toughness evaluation Is required.
N15              Drain- Penetration < 2.5" -   SA-508 Cl. 1 1.5.9 &        The discontinuity of the CRD Bottom Head                  (Heat 265M-1) 1.5.21           nozzle listed in Table A-1 bounds this discontinuity; RTNDT=-8"F                 therefore, no further fracture toughness evaluation is required.
N17              Seal Leak Detection -           Alloy 600  1.5.9 &          Not a pressure boundary Penetration -1"                            1.5.28        component; therefore, requires no fracture toughness evaluation.
Top Head Lifting Lugs        SA-533 GR. B  1.5.9 &          Not a pressure boundary CL. 1    1.5.14       component and loads only occur on this component when the reactor is shutdown during an outage. Therefore, no fracture toughness evaluation is required.
* The high/low pressure leak detector, and the seal leak detector are the same nozzle, these nozzles are the closure flange leak detection nozzles.
A-3

GE Nuclear Energy                                        GE-NE-0000-0003-5526-01 a APPENDIX A 

==REFERENCES:==

1.5. RPV Drawings 1.5.1. CBI #32, Rev. 5, "Top Head Assembly," (GE VPF # 3073-032, Rev. 5) 1.5.2. CBI #30, Rev. 3, "Top Head Flange Assembly," (GE VPF # 3073-030, Rev. 3) 1.5.3. CBI #26, Rev. 8, "Shell Flange Assembly w/ N17 Nozzle,"
(GE VPF # 3073-026, Rev. 9) 1.5.4. CBI #21, Rev. 2, "#1 Shell Ring Assembly," (GE VPF # 3073-021, Rev. 4) 1.5.5. CBI #22, Rev. 3, "#2 Shell Ring Assembly," (GE VPF # 3073-022, Rev. 4) 1.5.6. CBI #23, Rev. 2, "#3 Shell Ring Assembly," (GE VPF # 3073-023, Rev. 4) 1.5.7. CBI #24, Rev. 3, "#4 Shell Ring Assembly," (GE VPF # 3073-024, Rev. 4) 1.5.8. CBI #13, Rev. 5, "Bottom Head Assembly," (GE VPF # 3073-013, Rev. 6) 1.5.9. CBI #R1 3, Rev. 7, "Vessel, Nozzle & Outside Bracket As-Built Dimensions," (GE VPF # 3073-104, Rev. 8) 1.5.10. CBI #58, Rev. 5, "RHRPLPCI Mode Nozzle N6," (GE VPF # 3073-058, Rev. 5) 1.5.11. CBI #69, Rev. 4, "Instrumentation Nozzle N12," (GE VPF # 3073-069, Rev. 4) 1.5.12. CBI #19, Rev. 4, "Shroud Support Assembly," (GE VPF # 3073-019, Rev. 5) 1.5.13. CBI #17, Rev. 2, "Shroud Support Stubs," (GE VPF # 3073-017, Rev. 2) 1.5.14. CBI #40, Rev. 2, "Top Head Lift Lugs," (GE VPF # 3073-040, Rev. 3) 1.5.15. CBI #51, Rev. 8, "N3 Nozzle," (GE VPF # 3073-051, Rev. 8) 1.5.16. CBI #52, Rev. 7, "N4 Nozzle," (GE VPF # 3073-052, Rev. 7) 1.5.17. CBI #55, Rev. 6, "N5 Nozzle," (GE VPF # 3073-055, Rev. 7) 1.5.18. CBI #61, Rev. 2, "N7 Nozzle," (GE VPF # 3073-061, Rev. 3) 1.5.19. CBI #63, Rev. 5, "N9 Nozzle," (GE VPF # 3073-063, Rev. 5) 1.5.20. CBI #65, Rev. 85, "N10 Nozzle," (GE VPF # 3073-065, Rev. 5) 1.5.21. CBI #72, Rev. 4, "N1 5 Nozzle," (GE VPF # 3073-072, Rev. 4) 1.5.22. CBI #51, Rev. 8, "N3 Nozzle," (GE VPF # 3073-051, Rev. 8) 1.5.23. CBI #73, Rev. 5, "N16 Nozzle," (GE VPF # 3073-073, Rev. 5) 1.5.24. CBI #76, Rev. 2, "N18 Nozzle," (GE VPF # 3073-076, Rev. 3)
A-4

GE Nuclear Energy                                         GE-NE-0000-0003-5526-01 a 1.5.25. CBI #80, Rev. 2, "Stabilizer Brackets,' (GE VPF# 3073-080, Rev. 2) 1.5.26. CBI #9, Rev. 7, "Support Skirt Knuckle," (GE VPF # 3073-009, Rev. 7) 1.5.27. CBI #62, Rev. 2, "N8 Nozzle,' (GE VPF # 3073-062, Rev. 3) 1.5.28. GE Drawing 732E143, Rev. 16, "Purchase Part, Reactor Vessel,"
GE-NED, San Jose, CA.
1.5.29. CBI #46, Rev. 5, "Ni Nozzle," (GE VPF # 3073-046, Rev. 5) 1.5.30. CBI #48, Rev. 6, "N2 Nozzle," (GE VPF # 3073-048, Rev. 6) 1.6. Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation', GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)
(GE Proprietary).
A-5

GE Nuclear Energy               GE-NE-0000-0003-5526-01a APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM      UPPER        32 EFPY      BOTTOM       UPPER    32 EFPY HEAD      VESSEL     BELTLINE         HEAD       VESSEL  BELTIINE PRESSURE     CURVE A CURVEA CURVEA                  CURVEB CURVE B CURVE B (PSIG)         (OF)        (OF)        (F)           (OF)        (F)       (F) 0           68.0       86.0         86.0          68.0       86.0      86.0 10          68.0       86.0         86.0          68.0       86.0      86.0 20          68.0       86.0         86.0          68.0       86.0      86.0 30          68.0       86.0         86.0          68.0       86.0      86.0 40           68.0       86.0         86.0          68.0       86.0      86.0 50         68.0        86.0         86.0          68.0       86.0      86.0 60         68.0        86.0         86.0          68.0       86.0      86.0 70         68.0        86.0         86.0          68.0       86.0      86.0 80         68.0        86.0         86.0          68.0       86.0      86.0 90          68.0        86.0         86.0          68.0       86.0      86.0 100         68.0        86.0         86.0          68.0       86.0      86.0 110         68.0        86.0         86.0          68.0       86.0      86.0 120         68.0        86.0         86.0          68.0       86.0      86.0 130         68.0        86.0         86.0          68.0       86.0      86.2 140         68.0        86.0         86.0          68.0       86.0      89.4 150         68.0        86.0         86.0          68.0       86.0      92.2 160         68.0        86.0         86.0          68.0       86.0      94.9 170         68.0        86.0         86.0          68.0       86.0      97.5 180         68.0        86.0         86.0          68.0       87.9      99.9 190         68.0        86.0         86.0          68.0       90.2     102.2 200          68.0        86.0         86.0          68.0       92.3     104.3 210          68.0        86.0         86.0          68.0       94.3     106.3 220          68.0        86.0         86.0          68.0       96.3     108.3 230          68.0        86.0         86.0          68.0       98.1     110.1 240          68.0        86.0         86.0          68.0       99.9     111.9 B-2

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM       UPPER       32 EFPY      BOTTOM       UPPER    32 EFPY HEAD       VESSEL      BELTLINE        HEAD       VESSEL  BELTLINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVE B CURVE B (PSIG)         (&deg;F)        (&deg;F)        (OF)          (&deg;F)         (OF)     (OF) 250         68.0        86.0         86.0          68.0        101.6    113.6 260         68.0        86.0         86.0          68.0        103.2    115.2 270         68.0        86.0         86.0          68.0        104.8    116.8 280         68.0        86.0         86.0          68.0        106.3    118.3 290         68.0        86.0         86.0          68.0        107.8    119.8 300         68.0        86.0         86.0          68.0        109.2    121.2 310         68.0        86.0         86.0          68.0        110.5    122.5 312.5        68.0        86.0         86.0          68.0        110.9    122.9 312.5        68.0        116.0       116.0          68.0        146.0    146.0 320         68.0        116.0       116.0          68.0       '146.0    146.0 330         68.0        116.0       116.0          68.0        146.0    146.0 340         68.0        116.0       116.0          68.0        146.0    146.0 350         68.0        116.0       116.0          68.0        146.0    146.0 360         68.0        116.0       116.0          68.0        146.0    146.0 370         68.0        116.0       116.0          68.0        146.0    146.0 380         68.0        116.0       116.0          68.0        146.0    146.0 390         68.0        116.0       116.0          68.0        146.0    146.0 400         68.0        116.0       116.0          68.0        146.0    146.0 410         68.0        116.0       116.0          68.0        146.0    146.0 420         68.0        116.0       116.0          68.0        146.0    146.0 430         68.0        116.0       116.0          68.0        146.0    146.0 440         68.0        116.0       116.0          68.0        146.0    146.0 450         68.0        116.0       116.0          68.0        146.0    146.0 460         68.0        116.0       116.0          68.0        146.0    146.0 470         68.0        116.0       116.0          68.0        146.0    146.0 480         68.0        116.0       116.0          68.0        146.0    146.0 490         68.0        116.0       116.0          68.0        146.0    146.0 B-3

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM        UPPER       32 EFPY      BOTTOM        UPPER   32 EFPY HEAD       VESSEL      BELTLINE        HEAD       VESSEL  BELTLINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVE B CURVE B (PSIG)         (*F)        (OF)         (OF)          (F)         (OF)     (OF) 500         68.0        116.0        116.0         68.0        146.0    146.0 510         68.0        116.0        116.0         68.0        146.0    146.0 520         68.0        116.0        116.0         68.2        146.0    146.0 530         68.0        116.0        116.0         70.2        146.0    146.0 540         68.0        116.0        116.0         72.1        146.0    146.0 550         68.0        116.0        116.0         73.9        146.0    146.6 560         68.0        116.0        116.0         75.7        146.0    147.4 570         68.0        116.0        116.0         77.4        146.0    148.1 580         68.0        116.0        116.0         79.0        146.0    148.9 590         68.0        116.0        116.0         80.6        146.0    149.6 600         68.0        116.0       116.0          82.2        146.0   150.1 610         68.0        116.0       116.0          83.7        146.0   150.6 620         68.0        116.0       116.0          85.1        146.0   151.0 630         68.0        116.0       116.0          86.5        146.0   151.4 640         68.0        116.0       116.0          87.9        146.0   151.8 650         68.0        116.0       116.0          89.2        146.0   152.2 660         68.0        116.0       116.0          90.5        146.0   152.7 670         68.0        116.0       116.0          91.8        146.0   153.1 680         68.0        116.0       116.0          93.1        146.0   153.5 690         68.0        116.0       116.0          94.3        146.0   153.9 700         69.2        116.0       116.0          95.4        146.0   154.3 710         70.7        116.0       116.0          96.6        146.0   154.7 720         72.1        116.0       116.0          97.7        146.0   155.1 730         73.5        116.0       116.0          98.8        146.0   155.5 740         74.8        116.0       116.0          99.9        146.0   155.9 750         76.1        116.0       116.0          101.0       146.0   156.2 760         77.4        116.0       116.8          102.0       146.0   156.6 B-4

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM        UPPER      32 EFPY       BOTTOM       UPPER    32 EFPY HEAD      VESSEL BELTLINE             HEAD       VESSEL BELTIINE PRESSURE     CURVE A CURVE A CURVE A               CURVE B CURVE B CURVE B (PSIG)         (*F)        (OF)        (-F)           (OF)       (*F)     (*F) 770          78.6       116.0       117.6          103.0      146.0    157.0 780          79.8       116.0       118.3          104.0      146.0    157.4 790          81.0       116.0       119.1          105.0      146.0    157.8 800          82.2       116.0       119.9          105.9      146.1    158.1 810          83.3       116.0       120.6          106.9      146.5    158.5 820          84.4       116.0       121.4          107.8      146.9    158.9 830          85.5       116.0       122.1         108.7       147.2    159.2 840          86.5       116.0       122.8         109.6       147.6    159.6 850          87.6       116.0       123.5         110.4       147.9    159.9 860          88.6       116.0       124.2         111.3       148.3    160.3 870           89.6       116.0       124.9         112.1       148.6    160.6 880           90.5       116.0       125.6         113.0       149.0    161.0 890           91.5       116.0       126.3         113.8       149.3    161.3 900           92.4       116.0       126.9         114.6       149.7    161.7 910          93.4        116.0       127.6         115.4       150.0    162.0 920          94.3        116.2       128.2         116.1       150.4    162.4 930          95.1        116.9       128.9         116.9       150.7    162.7 940          96.0       117.5        129.5         117.7       151.0    163.0 950          96.9       118.1        130.1         118.4       151.4    163.4 960          97.7       118.7        130.7         119.1       151.7    163.7 970          98.6       119.3        131.3         119.9       152.0    164.0 980          99.4       119.9        131.9         120.6       152.4    164.4 990          100.2      120.5        132.5         121.3       152.7    164.7 1000         101.0      121.1        133.1         122.0       153.0    165.0 1010         101.7      121.7        133.7         122.6       153.3    165.3 1020         102.5      122.2        134.2         123.3       153.6    165.6 1030         103.3      122.8        134.8         124.0       154.0    166.0 B-5

GE Nuclear Energy                                                 GE-N E-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM      UPPER        32 EFPY      BOTTOM        UPPER    32 EFPY HEAD       VESSEL      BELTINE         HEAD       VESSEL   BELTLINE PRESSURE      CURVE A CURVE A CURVE A               CURVE B CURVE B CURVE B (PSIG)         (&deg;F)       (F)          (F)           (OF)         (OF)      (F) 1040         104.0      123.4        135.4         124.6        154.3     166.3 1050         104.7      123.9        135.9         125.3        154.6     166.6 1060         105.4      124.5        136.5         125.9        154.9     166.9 1070         106.2      125.0        137.0         126.5        155.2     167.2 1080         106.9      125.5        137.5         127.2        155.5     167.5 1090         107.6      126.1        138.1         127.8        155.8     167.8 1100         108.2      126.6        138.6         128.4        156.1    168.1 1105         108.6      126.8        138.8         128.7        156.3    168.3 1110         108.9      127.1        139.1         129.0        156.4    168.4 1120         109.6      127.6        139.6         129.6      '156.7     168.7 1130         110.2      128.1        140.1         130.2        157.0    169.0 1140         110.9      128.6        140.6         130.7        157.3    169.3 1150         111.5      129.1        141.1         131.3        157.6    169.6 1160         112.1      129.6        141.6         131.9        157.9    169.9 1170         112.8      130.1        142.1         132.4        158.2    170.2 1180         113.4      130.6        142.6         133.0        158.5    170.5 1190         114.0      131.1        143.1         133.5        158.7    170.7 1200         114.6      131.5        143.5         134.1        159.0    171.0 1210         115.2      132.0        144.0         134.6        159.3    171.3 1220         115.8      132.5        144.5         135.2        159.6    171.6 1230         116.3      132.9        144.9         135.7        159.9    171.9 1240         116.9      133.4        145.4         136.2        160.2    172.2 1250         117.5      133.8        145.8         136.7        160.4    172.4 1260         118.0      134.3        146.3         137.2        160.7    172.7 1270         118.6      134.7        146.7         137.7        161.0    173.0 1280         119.1      135.2        147.2         138.2        161.2    173.2 1290        119.7       135.6        147.6         138.7        161.5    173.5 B-6

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM SUPPER             32 EFPY      BOTTOM       UPPER    32 EFPY HEAD      VESSEL BELTLINE             HEAD       VESSEL BELTLINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVE B CURVE B (PSIG)         (OF)        (OF)        (OF)           (OF)       (OF)      (OF) 1300         120.2       136.0       148.0          139.2      161.8     173.8 1310         120.7       136.5       148.5          139.7      162.1     174.1 1320         121.3       136.9       148.9         140.2       162.3     174.3 1330         121.8      137.3        149.3         140.6       162.6     174.6 1340         122.3      137.7        149.7         141.1       162.8     174.8 1350         122.8      138.1        150.1         141.6       163.1     175.1 1360         123.3      138.6        150.6         142.0       163.4     175.4 1370         123.8      139.0        151.0         142.5       163.6     175.6 1380         124.3      139.4        151.4         142.9       163.9    175.9 1390         124.8      139.8        151.8         143.4       164.1    176.1 1400         125.3      140.2        152.2         143.8       164.4    176.4 B-7

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV & NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A           CURVE A       CURVE B         CURVE B          CURVE C (PSIG)        (&deg;F)          (OF)           (OF)          (OF)               (OF) 0          68.0         86.0           68.0           86.0               86.0 10         68.0         86.0           68.0           86.0               86.0 20         68.0         86.0           68.0           86.0              86.0 30         68.0         86.0           68.0           86.0              86.0 40          68.0         86.0           68.0           86.0              86.0 50         68.0         86.0           68.0           86.0              86.0 60         68.0         86.0           68.0           86.0              92.0 70         68.0         86.0           68.0           86.0              99.2 80         68.0         86.0           68.0           86.0              105.2 90         68.0         86.0           68.0           86.0              110.3 100         68.0         86.0           68.0           86.0              114.8 110         68.0         86.0           68.0           86.0              118.9 120         68.0         86.0           68.0           86.0              122.7 130         68.0         86.0           68.0           86.2              126.2 140         68.0         86.0           68.0           89.4              129.4 150         68.0         86.0           68.0           92.2              132.2 160         68.0         86.0           68.0           94.9              134.9 170         68.0         86.0           68.0           97.5              137.5 180         68.0         86.0           68.0           99.9              139.9 B-8

GE Nuclear Energy                                             GE-N E-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV &                     NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A           CURVE A       CURVE B         CURVE B          CURVE C (PSIG)        (&deg;F)          (OF)           (.F)           (OF)              (OF) 190         68.0         86.0            68.0          102.2             142.2 200         68.0         86.0            68.0          104.3             144.3 210         68.0         86.0            68.0          106.3             146.3 220         68.0         86.0            68.0          108.3             148.3 230         68.0         86.0            68.0          110.1             150.1 240         68.0         86.0            68.0          111.9             151.9 250         68.0         86.0            68.0          113.6             153.6 260         68.0         86.0            68.0          115.2             155.2 270         68.0         86.0            68.0          116.8             156.8 280         68.0         86.0            68.0          118.3             158.3 290         68.0         86.0            68.0          119.8             159.8 300         68.0         86.0            68.0          121.2             161.2 310         68.0         86.0            68.0          122.5             162.5 312.5        68.0         86.0            68.0          122.9             162.9 312.5        68.0         116.0           68.0          146.0             186.0 320         68.0         116.0           68.0          146.0             186.0 330         68.0         116.0           68.0          146.0             186.0 340         68.0         116.0           68.0          146.0             186.0 350         68.0         116.0           68.0          146.0             186.0 360         68.0         116.0           68.0          146.0             186.0 B-9

GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV &      NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A          CURVE A        CURVE B        CURVE B           CURVE C (PSIG)        (*F)          (OF)           (OF)           (OF)              (OF) 370         68.0         116.0           68.0          146.0             186.0 380         68.0         116.0           68.0          146.0             186.0 390         68.0         116.0           68.0          146.0             186.0 400          68.0         116.0           68.0          146.0             186.0 410          68.0         116.0           68.0          146.0             186.0 420         68.0         116.0           68.0          146.0             186.0 430         68.0         116.0           68.0          146.0             186.0 440         68.0         116.0           68.0          146.0             186.0 450         68.0         116.0           68.0          146.0             186.0 460         68.0         116.0           68.0          146.0             186.0 470         68.0         116.0           68.0          146.0             186.0 480          68.0         116.0           68.0          146.0             186.0 490         68.0         116.0           68.0          146.0             186.0 500         68.0         116.0           68.0          146.0             186.0 510         68.0         116.0           68.0          146.0             186.0 520         68.0         116.0           68.2          146.0             186.0 530         68.0         116.0           70.2          146.0             186.0 540         68.0         116.0           72.1          146.0             186.0 550         68.0         116.0           73.9          146.6             186.6 560         68.0         116.0           75.7          147.4             187.4 B-10

GE Nuclear Energy                                             GE-NE-O000-0003-5526-01a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV &      NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A          CURVE A        CURVE B        CURVE B           CURVE C (PSIG)         (*F)         (&deg;F)           (OF)           (&deg;F)              (&deg;F) 570         68.0         116.0           77.4          148.1             188.1 580         68.0         116.0           79.0          148.9             188.9 590         68.0         116.0           80.6          149.6             189.6 600         68.0         116.0           82.2          150.1             190.1 610         68.0         116.0           83.7          150.6             190.6 620         68.0         116.0           85.1          151.0             191.0 630         68.0         116.0          86.5           151.4             191.4 640         68.0         116.0          87.9           151.8             191.8 650          68.0         116.0          89.2           152.2             192.2 660          68.0        116.0           90.5           152.7             192.7 670          68.0        116.0           91.8           153.1             193.1 680          68.0        116.0           93.1           153.5             193.5 690          68.0        116.0           94.3           153.9             193.9 700          69.2        116.0           95.4           154.3             194.3 710          70.7        116.0           96.6           154.7             194.7 720          72.1        116.0           97.7           155.1             195.1 730          73.5        116.0           98.8           155.5             195.5 740          74.8        116.0           99.9           155.9             195.9 750          76.1        116.0          101.0           156.2             196.2 760          77.4        116.8          102.0           156.6             196.6 B-11

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV &      NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTL1NE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A          CURVE A        CURVE B         CURVE B          CURVE C (PSIG)        (&deg;F)          (&deg;F)           (7F)           (OF)              (OF) 770         78.6         117.6          103.0          157.0             197.0 780          79.8         118.3          104.0          157.4             197.4 790         81.0         119.1          105.0          157.8             197.8 800          82.2         119.9          105.9          158.1             198.1 810          83.3         120.6          106.9          158.5             198.5 820          84.4         121.4          107.8          158.9             198.9 830         85.5         122.1          108.7          159.2             199.2 840         86.5         122.8          109.6          159.6             199.6 850          87.6         123.5          110.4          159.9             199.9 860          88.6         124.2          111.3          160.3             200.3 870         89.6         124.9          112.1          160.6             200.6 880          90.5         125.6          113.0          161.0             201.0 890          91.5         126.3          113.8          161.3             201.3 900          92.4         126.9          114.6          161.7             201.7 910          93.4         127.6          115.4          162.0             202.0 920         94.3         128.2          116.1          162.4             202.4 930          95.1         128.9          116.9          162.7             202.7 940         96.0         129.5          117.7          163.0             203.0 950         96.9         130.1          118.4          163.4             203.4 960          97.7         130.7          119.1          163.7             203.7 B-12

GE Nuclear Energy                                             GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM                UPPER RPV & NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT       & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A           CURVE A       CURVE B         CURVE B          CURVE C (PSIG)        (OF)          (OF)           (&deg;F)           (OF)              (OF) 970         98.6         131.3          119.9          164.0             204.0 980         99.4         131.9          120.6          164.4             204.4 990        100.2         132.5          121.3          164.7             204.7 1000        101.0         133.1          122.0          165.0             205.0 1010        101.7         133.7          122.6          165.3             205.3 1020        102.5         134.2          123.3          165.6             205.6 1030        103.3         134.8          124.0          166.0             206.0 1040        104.0         135.4          124.6          166.3             206.3 1050        104.7         135.9          125.3          166.6             206.6 1060        105.4         136.5          125.9          166.9             206.9 1070        106.2         137.0          126.5          167.2             207.2 1080        106.9         137.5          127.2          167.5             207.5 1090        107.6         138.1          127.8          167.8             207.8 1100        108.2         138.6          128.4          168.1             208.1 1105        108.6         138.8          128.7          168.3             208.3 1110        108.9         139.1          129.0          168.4             208.4 1120        109.6         139.6          129.6          168.7             208.7 1130        110.2         140.1          130.2          169.0             209.0 1140        110.9         140.6          130.7          169.3             209.3 1150        111.5         141.1          131.3          169.6             209.6 B-1 3

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV & NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE    CURVE A       CURVE A       CURVE B         CURVE B          CURVE C (PSIG)       (OF)          (OF)           (OF)           (OF)              (&deg;F) 1160       112.1         141.6          131.9          169.9             209.9 1170       112.8         142.1          132.4          170.2             210.2 1180       113.4         142.6          133.0          170.5             210.5 1190       114.0         143.1          133.5          170.7             210.7 1200       114.6         143.5          134.1          171.0             211.0 1210       115.2         144.0          134.6          171.3             211.3 1220       115.8         144.5          135.2          171.6             211.6 1230       116.3         144.9          135.7          171.9             211.9 1240       116.9         145.4          136.2          172.2             212.2 1250       117.5         145.8          136.7          172.4             212.4 1260       118.0         146.3          137.2          172.7             212.7 1270       118.6         146.7          137.7          173.0             213.0 1280       119.1         147.2          138.2          173.2             213.2 1290       119.7         147.6          138.7          173.5             213.5 1300       120.2         148.0          139.2          173.8             213.8 1310       120.7         148.5          139.7          174.1             214.1 1320       121.3         148.9          140.2          174.3             214.3 1330       121.8         149.3          140.6          174.6             214.6 1340       122.3         149.7          141.1          174.8             214.8 1350       122.8         150.1          141.6          175.1             215.1 B-14

GE Nuclear Energy                                            G E-N E-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM     UPPER RPV & BOTTOM            UPPER RPV & NONBELTLINE HEAD     BELTLINE AT        HEAD       BELTLINE AT        & BELTLINE 32 EFPY                       32 EFPY           32 EFPY PRESSURE CURVE A           CURVE A       CURVE B         CURVE B          CURVE C (PSIG)        (*F)          (&deg;F)           (&deg;F)           (*F)              (F) 1360        123.3         150.6          142.0          175.4             215.4 1370        123.8         151.0          142.5          175.6             215.6 1380        124.3         151.4          142.9          175.9             215.9 1390        124.8         151.8          143.4          176.1             216.1 1400        125.3         152.2          143.8          176.4             216.4 B-15

GE Nuclear Energy                                                GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM      UPPER        20 EFPY      BOTTOM       UPPER    20 EFPY HEAD      VESSEL      BELTLINE        HEAD       VESSEL  BELT.INE PRESSURE     CURVE A CURVE A CURVE A               CURVE B CURVE B       CURVE B (PSIG)        (F)         (OF)        (OF)           (OF)       (OF)      (.F) 0          68.0       86.0         86.0          68.0        86.0      86.0 10         68.0       86.0         86.0          68.0        86.0      86.0 20          68.0       86.0         86.0          68.0        86.0      86.0 30          68.0       86.0         86.0          68.0        86.0      86.0 40          68.0       86.0         86.0          68.0        86.0      86.0 50         68.0       86.0         86.0          68.0        86.0      86.0 60          68.0       86.0         86.0          68.0        86.0      86.0 70         68.0       86.0         86.0          68.0        86.0      86.0 80         68.0       86.0         86.0          68.0        86.0      86.0 90         68.0       86.0         86.0          68.0        86.0      86.0 100         68.0       86.0         86.0          68.0        86.0      86.0 110         68.0       86.0         86.0          68.0        86.0      86.0 120         68.0       86.0         86.0          68.0        86.0      86.0 130         68.0       86.0         86.0          68.0        86.0      86.0 140         68.0       86.0         86.0          68.0        86.0      86.0 150         68.0       86.0         86.0          68.0        86.0      86.0 160         68.0       86.0         86.0          68.0        86.0      86.0 170         68.0       86.0         86.0          68.0        86.0      86.0 180         68.0       86.0         86.0          68.0        87.9      86.0 190         68.0       86.0         86.0          68.0        90.2      87.2 200         68.0       86.0         86.0          68.0        92.3      89.3 210         68.0       86.0         86.0          68.0        94.3      91.3 220         68.0       86.0         86.0           68.0       96.3      93.3 230         68.0       86.0         86.0           68.0       98.1      95.1 240         68.0       86.0         86.0           68.0       99.9      96.9 B-16

GE Nuclear Energy                                                GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;Flhr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM       UPPER       20 EFPY      BOTTOM       UPPER    20 EFPY HEAD      VESSEL      BELTLINE        HEAD       VESSEL  BELTLINE PRESSURE     CURVE A CURVE A CURVE A               CURVE B CURVE B CURVE B (PSIG)        (OF)        (F)          ((F)
(F)                      (OF)      (OF) 250         68.0        86.0        86.0          68.0       101.6      98.6 260         68.0        86.0        86.0          68.0       103.2     100.2 270         68.0        86.0        86.0          68.0       104.8     101.8 280         68.0        86.0        86.0          68.0       106.3     103.3 290         68.0        86.0        86.0          68.0       107.8     104.8 300         68.0        86.0        86.0          68.0       109.2     106.2 310         68.0        86.0        86.0          68.0       110.5     107.5 312.5        68.0        86.0        86.0          68.0       110.9     107.9 312.5        68.0       116.0        116.0         68.0       146.0     146.0 320         68.0       116.0        116.0         68.0       146.0     146.0 330         68.0       116.0       116.0          68.0       146.0     146.0 340         68.0       116.0       116.0          68.0       146.0     146.0 350         68.0       116.0        116.0         68.0       146.0     146.0 360         68.0       116.0        116.0         68.0       146.0     146.0 370         68.0       116.0        116.0         68.0       146.0     146.0 380         68.0       116.0       116.0          68.0       146.0     146.0 390         68.0       116.0       116.0          68.0       146.0     146.0 400         68.0       116.0       116.0          68.0       146.0     146.0 410         68.0       116.0        116.0         68.0       146.0     146.0 420         68.0       116.0        116.0         68.0       146.0     146.0 430         68.0       116.0        116.0         68.0       146.0     146.0 440         68.0       116.0        116.0         68.0       146.0     146.0 450         68.0       116.0        116.0         68.0       146.0     146.0 460         68.0       116.0        116.0         68.0       146.0     146.0 470         68.0       116.0        116.0         68.0       146.0     146.0 480         68.0       116.0        116.0         68.0       146.0     146.0 490         68.0       116.0        116.0         68.0       146.0     146.0 B-1 7

GE Nuclear Energy                                                GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM       UPPER       20 EFPY      BOTTOM        UPPER   20 EFPY HEAD      VESSEL      BELTLINE        HEAD       VESSEL   BELTLINE PRESSURE     CURVE A CURVE A CURVE A               CURVE B CURVE B CURVE B (PSIG)        (F)         (OF)         (O)           (OF)        (OF)     f(F) 500         68.0       116.0       116.0          68.0        146.0    146.0 510         68.0       116.0       116.0          68.0        146.0    146.0 520         68.0       116.0       116.0          68.2        146.0    146.0 530         68.0       116.0       116.0          70.2        146.0    146.0 540         68.0       116.0       116.0          72.1        146.0    146.0 550         68.0       116.0       116.0          73.9        146.0    146.0 560         68.0       116.0       116.0          75.7        146.0    146.0 570         68.0       116.0       116.0          77.4        146.0    146.0 580         68.0       116.0        116.0         79.0        146.0    146.0 590         68.0       116.0       116.0          80.6        146.0    146.0 600         68.0       116.0       116.0          82.2        146.0    146.0 610         68.0       116.0       116.0          83.7        146.0    146.0 620         68.0       116.0       116.0          85.1        146.0    146.0 630         68.0       116.0       116.0          86.5        146.0    146.0 640         68.0       116.0       116.0          87.9        146.0    146.0 650         68.0       116.0        116.0         89.2        146.0    146.0 660         68.0       116.0        116.0         90.5        146.0    146.0 670         68.0       116.0        116.0         91.8        146.0    146.0 680         68.0       116.0        116.0         93.1        146.0    146.0 690         68.0       116.0        116.0         94.3        146.0    146.0 700         69.2       116.0        116.0         95.4        146.0    146.0 710         70.7       116.0        116.0         96.6        146.0    146.0 720         72.1       116.0        116.0         97.7        146.0    146.0 730         73.5       116.0        116.0         98.8        146.0    146.0 740         74.8       116.0        116.0         99.9        146.0    146.0 750         76.1       116.0        116.0         101.0       146.0    146.0 760         77.4       116.0        116.0         102.0       146.0    146.0 B-1 8

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM        UPPER       20 EFPY      BOTTOM        UPPER   20 EFPY HEAD       VESSEL      BELTLINE        HEAD       VESSEL  BELTLINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVE B CURVE B (PSIG)         (F)         (F)          (F)           (F)         (OF)     (OF) 770         78.6        116.0        116.0         103.0       146.0    146.0 780         79.8        116.0        116.0         104.0       146.0    146.0 790         81.0        116.0        116.0         105.0       146.0    146.0 800         82.2        116.0        116.0         105.9       146.1    146.0 810         83.3        116.0        116.0         106.9       146.5    146.0 820         84.4        116.0        116.0         107.8       146.9    146.0 830         85.5        116.0        116.0         108.7       147.2    146.0 840         86.5        116.0        116.0         109.6       147.6    146.0 850         87.6        116.0        116.0         110.4       147.9    146.0 860         88.6        116.0        116.0         111.3       148.3    146.0 870         89.6        116.0        116.0         112.1       148.6    146.0 880         90.5        116.0        116.0         113.0       149.0    146.0 890         91.5        116.0        116.0         113.8       149.3    146.3 900         92.4        116.0        116.0         114.6       149.7    146.7 910         93.4        116.0        116.0         115.4       150.0    147.0 920         94.3        116.2        116.0         116.1       150.4    147.4 930         95.1        116.9        116.0         116.9       150.7    147.7 940         96.0        117.5        116.0         117.7       151.0    148.0 950         96.9        118.1        116.0         118.4       151.4    148.4 960         97.7        118.7        116.0         119.1       151.7    148.7 970          98.6       119.3        116.3         119.9       152.0    149.0 980          99.4       119.9        116.9         120.6       152.4    149.4 990         100.2       120.5        117.5         121.3       152.7    149.7 1000         101.0       121.1        118.1         122.0       153.0    150.0 1010         101.7       121.7        118.7         122.6       153.3    150.3 1020         102.5       122.2        119.2         123.3       153.6    150.6 1030         103.3       122.8        119.8         124.0       154.0    151.0 B-1 9

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM        UPPER       20 EFPY      BOTTOM       UPPER    20 EFPY HEAD       VESSEL      BELTLINE        HEAD       VESSEL  BELTLINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVE B CURVE B (PSIG)         (F)         (F)          ("F)          (F)        (F)       (OF) 1040         104.0       123.4        120.4         124.6      154.3     151.3 1050         104.7       123.9        120.9         125.3      154.6     151.6 1060         105.4       124.5        121.5         125.9      154.9     151.9 1070         106.2       125.0        122.0         126.5      155.2     152.2 1080         106.9       125.5        122.5         127.2      155.5     152.5 1090         107.6       126.1        123.1         127.8      155.8     152.8 1100         108.2       126.6        123.6         128.4      156.1     153.1 1105         108.6       126.8        123.8         128.7      156.3     153.3 1110         108.9       127.1        124.1         129.0      156.4     153.4 1120         109.6       127.6        124.6         129.6      156.7     153.7 1130         110.2       128.1        125.1         130.2      157.0     154.0 1140         110.9       128.6        125.6         130.7      157.3     154.3 1150         111.5       129.1        126.1         131.3      157.6     154.6 1160         112.1       129.6        126.6         131.9      157.9     154.9 1170         112.8       130.1        127.1         132.4      158.2     155.2 1180         113.4       130.6        127.6         133.0      158.5     155.5 1190         114.0       131.1        128.1         133.5      158.7     155.7 1200         114.6       131.5        128.5         134.1      159.0     156.0 1210         115.2       132.0        129.0         134.6      159.3     156.3 1220         115.8       132.5        129.5         135.2      159.6     156.6 1230         116.3       132.9        129.9         135.7      159.9     156.9 1240         116.9       133.4        130.4         136.2      160.2     157.2 1250         117.5       133.8        130.8         136.7      160.4     157.4 1260         118.0       134.3        131.3         137.2      160.7     157.7 1270         118.6       134.7        131.7         137.7      161.0     158.0 1280         119.1       135.2        132.2         138.2      161.2     158.2 1290         119.7       135.6        132.6         138.7      161.5     158.5 B-20

GE Nuclear Energy                                                 GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curves B & C and 20 &deg;F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM        UPPER       20 EFPY      BOTTOM        UPPER   20 EFPY HEAD       VESSEL      BELTLINE        HEAD       VESSEL  BELTIINE PRESSURE     CURVE A CURVE A CURVE A                CURVE B CURVEB        CURVE B (PSIG)         (F)         ("F)        (F)            (F)         (F)      (F) 1300         120.2       136.0       133.0          139.2       161.8    158.8 1310         120.7       136.5       133.5          139.7       162.1    159.1 1320         121.3       136.9       133.9          140.2       162.3    159.3 1330         121.8       137.3       134.3          140.6       162.6    159.6 1340         122.3       137.7       134.7          141.1       162.8    159.8 1350         122.8       138.1       135.2          141.6       163.1    160.1 1360         123.3       138.6       135.7          142.0       163.4    160.4 1370         123.8       139.0       136.2          142.5       163.6    160.8 1380         124.3       139.4       136.8          142.9       163.9    161.3 1390         124.8       139.8       137.3          143.4       164.1    161.7 1400         125.3       140.2       137.8          143.8       164.4    162.1 B-21

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curve C For Figure 5-9 UPPER               BOTTOM                  20 EFPY PRESSURE      VESSEL CURVE C         HEAD CURVE C          BELTLINE CURVE C (PSIG)              (@F)                 (OF)                   (OF) 0                86.0                 68.0                   86.0 10                86.0                 68.0                   86.0 20                86.0                 68.0                   86.0 30                86.0                 68.0                   86.0 40                86.0                 68.0                   86.0 50                86.0                 68.0                   86.0 60                86.0                 68.0                   86.0 70                87.2                 68.0                   86.0 80                93.2                 68.0                   86.0 90                98.3                 68.0                   86.0 100              102.8                 68.0                   86.0 110              106.9                 68.0                   86.0 120              110.7                 68.0                   86.0 130              114.2                 68.0                   86.0 140              117.4                 68.0                   86.0 150              120.2                 68.0                   86.0 160              122.9                 68.0                   86.0 170              125.5                 68.0                   86.0 180              127.9                 68.0                   86.0 190               130.2                68.0                   86.0 200               132.3                68.0                   86.0 210               134.3                68.0                   86.0 220               136.3                68.0                   86.0 230               138.1                68.0                   86.0 240               139.9                68.0                   86.0 250               141.6                68.0                   86.0 B-22

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;Flhr for Curve C For Figure 5-9 UPPER               BOTTOM                  .20 EFPY PRESSURE      VESSEL CURVE C         HEAD CURVE C          BELTLINE CURVE C (PSIG)              (OF)                 (O)                     (*F) 260              143.2                 68.0                    86.0 270              144.8                 68.0                    86.0 280              146.3                 68.0                    86.0 290              147.8                 68.0                    86.0 300              149.2                 68.0                    86.0 310              150.5                 68.0                    86.0 312.5             150.9                 68.0                    86.6 312.5             186.0                 68.0                   186.0 320              186.0                 68.0                   186.0 330              186.0                 68.0                   186.0 340              186.0                 68.0                   186.0 350              186.0                 68.0                   186.0 360              186.0                 68.0                   186.0 370              186.0                 68.0                   186.0 380              186.0                 68.0                   186.0 390              186.0                 71.3                   186.0 400              186.0                 75.3                   186.0 410              186.0                 79.0                   186.0 420              186.0                 82.5                   186.0 430              186.0                 85.8                   186.0 440              186.0                 88.8                   186.0 450              186.0                 91.7                   186.0 460              186.0                 94.4                    186.0 470              186.0                 97.0                    186.0 480              186.0                 99.5                    186.0 490              186.0                101.8                    186.0 500              186.0                104.0                    186.0 510               186.0               106.2                    186.0 B-23

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curve C For Figure 5-9 UPPER               BOTTOM                  20 EFPY PRESSURE      VESSEL CURVE C         HEAD CURVE C          BELTLINE CURVE C (PSIG)              ('F)                 (OF)                    (F) 520              186.0                108.2                   186.0 530              186.0                110.2                   186.0 540              186.0                112.1                   186.0 550              186.0                113.9                   186.0 560              186.0                115.7                   186.0 570              186.0                117.4                   186.0 580              186.0                119.0                   186.0 590              186.0                120.6                   186.0 600              186.0                122.2                   186.0 610              186.0                123.7                   186.0 620              186.0                125.1                   186.0 630              186.0                126.5                   186.0 640              186.0                127.9                   186.0 650              186.0                129.2                   186.0 660              186.0                130.5                   186.0 670              186.0                131.8                   186.0 680              186.0                133.1                   186.0 690              186.0                134.3                   186.0 700              186.0                135.4                   186.0 710              186.0                136.6                   186.0 720              186.0                137.7                   186.0 730              186.0                138.8                   186.0 740              186.0                 139.9                  186.0 750              186.0                 141.0                  186.0 760              186.0                 142.0                  186.0 770              186.0                 143.0                  186.0 780              186.0                 144.0                  186.0 790              186.0                 145.0                  186.0 B-24

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curve C For Figure 5-9 UPPER               BOTTOM                  20 EFPY PRESSURE      VESSEL CURVE C         HEAD CURVE C          BELTLINE CURVE C (PSIG)              (&deg;F)                  ("F)                   (F) 800              186.1                145.9                   186.0 810              186.5                146.9                   186.0 820              186.9                147.8                   186.0 830              187.2                148.7                   186.0 840              187.6                149.6                   186.0 850              187.9                150.4                   186.0 860              188.3                151.3                   186.0 870              188.6                152.1                   186.0 880              189.0                153.0                   186.0 890              189.3                153.8                   186.0 900              189.7                154.6                   186.0 910              190.0                155.4                   186.0 920              190.4                156.1                   186.0 930              190.7                 156.9                  186.0 940              191.0                 157.7                  186.0 950              191.4                 158.4                  186.0 960              191.7                 159.1                  186.0 970              192.0                 159.9                  186.0 980              192.4                 160.6                  186.0 990              192.7                 161.3                  186.0 1000              193.0                 162.0                  186.0 1010              193.3                 162.6                  186.0 1020              193.6                 163.3                  186.0 1030              194.0                 164.0                  186.0 1040              194.3                 164.6                  186.0 1050              194.6                 165.3                  186.0 1060              194.9                 165.9                  186.0 1070              195.2                 166.5                  186.1 B-25

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curve C For Figure 5-9 UPPER               BOTTOM                  20 EFPY PRESSURE       VESSEL CURVE C        HEAD CURVE C          BELTLINE CURVE C (PSIG)              (OF)                 (OF)                    (OF) 1080              195.5                167.2                   186.7 1090              195.8                167.8                   187.3 1100              196.1                168.4                   187.8 1105              196.3                168.7                   188.1 1110              196.4                169.0                   188.4 1120              196.7                169.6                   188.9 1130              197.0                170.2                   189.4 1140              197.3                170.7                   190.0 1150              197.6                171.3                   190.5 1160              197.9                171.9                   191.0 1170              198.2                172.4                   191.5 1180              198.5                173.0                   192.0 1190              198.7                173.5                   192.5 1200              199.0                174.1                   193.0 1210              199.3                174.6                   193.5 1220              199.6                175.2                   194.0 1230              199.9                175.7                   194.5 1240              200.2                176.2                   195.0 1250              200.4                176.7                   195.5 1260              200.7                177.2                   195.9 1270              201.0                177.7                   196.4 1280              201.2                178.2                   196.9 1290              201.5                178.7                   197.3 1300              201.8                179.2                   197.8 1310              202.1                179.7                   198.2 1320              202.3                180.2                   198.7 1330              202.6                180.6                   199.1 1340              202.8                181.1                   199.5 B-26

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 &deg;F/hr for Curve C For Figure 5-9 UPPER              BOTTOM                  20 EFPY PRESSURE       VESSEL CURVE C       HEAD CURVE C           BELTLINE CURVE C (PSIG)               (F)                 (F)                     (*F) 1350               203.1               181.6                   200.0 1360               203.4               182.0                   200.4 1370               203.6               182.5                   200.8 1380               203.9               182.9                   201.3 1390               204.1               183.4                   201.7 1400               204.4               183.8                   202.1 B-27

GE Nuclear Energy                       GE-NE-0000-0003-5526-01 a APPENDIX C Operating And Temperature Monitoring Requirements C-1

GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.
One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing. An experiment has been conducted at a BWR-4 that showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltline temperature during pressure testing. Thermocouples on the RPV flange to shell junction outside surface should be used to monitor compliance with upper vessel curve. Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D. First, however, it should be determined whether there are significant temperature differences between the beltline region and the bottom head region.
C.2 DETERMINING WHICH CURVE TO FOLLOW The following subsections outline the criteria needed for determining which curve is governing during different situations. The application of the P-T curves and some of the assumptions inherent in the curves to plant operation is dependent on the proper monitoring of vessel temperatures.
C-2

GE Nuclear Energy                                             GE-NE-0000-0003-5526-O1a C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by .- 0&deg;F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.
C.2.2 Curve B: Non-Nuclear Heatup/Cooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 20&deg;F per hour during a hydrotest and when the core is not critical.
C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core is critical. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.
C.3 REACTOR OPERATION VERSUS OPERATING LIMITS For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves). The operations where P-T curve compliance is typically monitored closely are planned events, such as vessel boltup, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.
The most severe unplanned transients relative to the P-T curves are those that result from SCRAMs, which sometimes include recirculation pump trips. Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high.
Experience with such events has shown that operator action is necessary to avoid P-T curve exceedance, but there is adequate time for operators to respond.
C-3

GE Nuclear Energy                                           GE-NE-0000-0003-5526-01a In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:
        "* Head flange boltup
        "* Leakage test (Curve A compliance)                                          .,
        "* Startup (coolant temperature change of less than or equal to 100OF in one hour period heatup)
        "* Shutdown (coolant temperature change of less than or equal to 100OF in one hour period cooldown)
        "* Recirculation pump trip, bottom head stratification (Curve B compliance)
C-4

GE Nuclear Energy             GE-NE-0000-0003-5526-01a APPENDIX D GE SIL 430 D-1

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01 a September 27, 1985                                                         SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions                   -. A concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stress requirements.
As such, the purpose of this Service Information Letter is to provide a summary of RPV temperature monitoring measurements, their primary and alternate uses and their limitations (See the attached table). Of basic concern is temperature monitoring to comply with brittle fracture temperature limits and for vessel thermal stresses during RPV heatup and cooldown. General Electric recommends that BWR owners/operators review this table against their current practices and evaluate any inconsistencies.
TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)
Measurement                             Use                           Limitations Steam dome saturation              Primary measurement              Must convert saturated temperature as determined          above 212&deg;F for Tech              steam pressure to from main steam instrument         Spec 1OOOF/hr heatup             temperature.
line pressure                      and cooldown rate.
Recirc suction line                Primary measurement              Must have recirc flow.
coolant temperature.               below 212&deg;F for Tech              Must comply with SEL 251 Spec 100&deg;F/hr heatup              to avoid vessel stratification.
and cooldown rate.
Alternate measurement             When above 212&deg;F need to above 212&deg;F.                      allow for temperature variations (up to 10-150F lower than steam dome saturation temperature) caused primarily by FW flow variations.
D-2

GE Nuclear Energy                               GE-NE-0000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement                     Use                    Limitations Alternate measurement for RPV drain line temperature (can use to comply with delta T limit between steam dome saturation temperature and bottom head drain line temperature).
RHR heat exchanger        Alternate measurement       Must have previously inlet coolant             for Tech Spec IOOOF/hr      correlated RHR inlet temperature               cooldown rate when in       coolant temperature shutdown cooling mode.      versus RPV coolant temperature.
RPV drain line            Primary measurement to      Must have drain line coolant temperature       comply with Tech Spec        flow. Otherwise, delta T limit between       lower than actual steam dome saturated        temperature and higher temp and drain line         delta T's will be indicated coolant temperature.        Delta T limit is 100&deg;F for BWR/6s and 1450F for earlier BWRs.
Primary measurement to       Must have drain line comply with Tech Spec        flow. Use to verify brittle fracture             compliance with Tech limits during cooldown. Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).
Alternate information       Must compensate for outside only measurement for        metal temperature lag bottom head inside/         during heatup/cooldown.
outside metal surface       Should have drain line flow.
temperatures.
D-3

GE Nuclear Energy                               GE-NE-0000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement                      Use                    Limitations Closure head flanges     Primary measurement for      Use for metal (not coolant) outside surface T/Cs      BWR/6s to comply with        temperature. Install Tech Spec brittle fracture   temporary T/Cs for metal temperature limit      alternate measurement, if for head boltup.             required.
One of two primary measure ments for BWR/6s for hydro test.
RPV flange-to-shell       Primary measurement for      Use for metal (not coolant) junction outside          BWRs earlier than 6s to      temperature. Response surface T/Cs              comply with Tech Spec        faster than closure head brittle fracture metal       flange T/Cs.
temperature limit for head boltup.
One of two primary           Use RPV closure head flange measurements for BWRs        outside surface as alternate earlier than 6s for          measurement.
hydro test. Preferred in lieu of closure head flange T/Cs if available.
RPV shell outside         Information only.            Slow to respond to RPV surface T/Cs                                          coolant changes. Not available on BWR/6s.
Top head outside          Information only.            Very slow to respond to RPV surface T/Cs                                          coolant changes. Not avail able on BWRI6s.
D-4

GE Nuclear Energy                                            GE-NE-O000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement                                Use                         Limitations Bottom head outside                 I of 2 primary measurements     Should verify that vessel surface T/Cs                       to comply with                   stratification is not Tech Spec brittle fracture       present for vessel hydro.
metal temperature                (see SIL No. 251).
limit for hydro test.
Primary measurement to           Use during heatup to verify comply with Tech Spec            compliance with Tech Spec brittle fracture metal           metal temperature/reactor temperature limits               pressure curves.
during heatup.
Note: RPV vendor specified metal T limits for vessel heatup and cooldown should be checked during initial plant startup tests when initial RPV vessel heatup and cooldown tests are run.
D-5

GE Nuclear Energy                                                  GE-NE-0000-0003-5526-01a Product 

==Reference:==
B21 Nuclear Boiler Prepared By: A.C. Tsang Approved for Issue:                                  Issued By:
B.H. Eldridge, Mgr.                                  D.L. AlIred, Manager Service Information                                  Customer Service Information and Analysis Notice:
SILs pertain only to GE BWRs. GE prepares SILs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained in SILs to any plant or facility other than GE BWRs as designed and fumished by GE. Determination of applicability of information contained in any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of information contained in SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SlLs.
D-6

GE Nuclear Energy                             GE-NE-0000-0003-5526-01 a APPENDIX E Determination of Beitline Region and Impact on Fracture Toughness E-1

GE Nuclear Energy                                              GE-NE-0000-0003-5526-01 a 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:
 "T he region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage" To establish the value of peak fluence for identification of beltline materials (as discussed above), the 10CFR50 Appendix H fluence value used to determine the need for a surveillance program was used; the value specified is a peak fluence (E>1 MEV) of 1.Oe17 n/cm 2 . Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm2 , then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.
The following dimensions are obtained from the referenced drawings:
Shell # 2 - Top of Active Fuel (TAF): 366.31" (from vessel 0) (Reference 1)
Shell # 1 - Bottom of Active Fuel (BAF): 216.31" (from vessel 0) (Reference 1)
Bottom of LPCI Nozzle in Shell # 2: 355.06" (from vessel 0) (Reference 2)
Center line of LPCI Nozzle in Shell # 2: 372" (from vessel 0) (Reference 3)
Top of Recirculation Outlet Nozzle in Shell # 1:197.91" (from vessel 0) (Reference 4)
Center line of Recirculation Outlet Nozzle in Shell # 1: 172.5" (from vessel 0) (Reference 3)
Top of Recirculation Inlet Nozzle in Shell # 1: 198.56" (from vessel 0) (Reference 5)
Center line of Recirculation Inlet Nozzle in Shell # 1: 181" (from vessel 0) (Reference 3)
As shown above, the LPCI nozzle is within the core beitline region. This nozzle is bounded by the feedwater pressure-temperature curve as stated in Appendix A.
From [3], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -18" from BAF and the top of the recirculation outlet nozzle is -18" from BAF), and no other nozzles are within the BAF-TAF region of the reactor vessel. Therefore, if it can be shown that the peak E-2

GE Nuclear Energy                                             GE-NE-0000-0003-5526-01a fluence at this location is less than I.0E17 n/cm 2, it can be safely concluded that all nozzles are outside the beltline region of the reactor vessel.
Based on the axial flux profile [6], the RPV flux level at -10" below the BAF dropped to less than 0.1 of the peak flux level at the same radius. Likewise, the RPV flux level at
-10" above the TAF dropped to less than 0.1 of the peak flux at the same radius.
Therefore, if the RPV fluence is 1.09E18 n/cm2 [6], fluence at -10" below BAF and -10" above TAF are expected to be less than 1.OE17 n/cm 2 at 32 EFPY. The beltline region considered in the development of the P-T curves is adjusted to include the additional 10" above and below the active fuel region. The adjusted beltline region extends from 206.31" to 376.31" above reactor vessel "0".
Based on the above, it is concluded that none of the LaSalle Unit 2 reactor vessel nozzles, other than the LPCI nozzle, which is considered in the P-T curve evaluation, are in the beltline region.
E-3

GE Nuclear Energy                                        GE-NE-0000-0003-5526-01 a Appendix E 

==References:==
: 1. Source of Bottom of Beitline Elevation: Figure Q121.7-2, uWelds in Beltline Region of Reactor Vessel - Unit 2", page Q121.7-12.
: 2.      CBIN Drawing #58, Revision 5, "RHR/LPCI Nozzle N6",
(GE VPF #3073-58-5).
: 3.      CBIN Drawing #R13, Revision 3, "Vessel, Nozzle, & Outside Bracket As Built Dimensions" (GE VPF #3073-104-8).
: 4.      CBIN Drawing #46, Revision 5, "Recirculation Outlet Nozzle NI",
(GE VPF #3073-46-5).
: 5.      CBIN Drawing #48, Revision 6, "Recirculation Inlet Nozzle N2",
(GE VPF #3073-48-6).
: 6.      Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation", GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).
E-4

GE Nuclear Energy                             GE-NE-0000-0003-5526-01a APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)
F-1

GE Nuclear Energy                                            GE-NE-0000-0003-5526-01a Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Reg.
Guide 1.99, Rev. 2 [2] methods, are summarized in Table F-I.
The USE decrease prediction values from Reg. Guide 1.99, Rev. 2 [2] were used for the beltline plates and welds in Table F-I. These calculations are based on the peak 1/4T fluence for all materials other than the LPCI nozzle, for conservatism. Because the Charpy data available for the LPCI nozzle consists of shear energy of 60%, this conservatism is not applied to the 32 EFPY USE calculation for this component; the 1/4T fluence for the LPCI nozzle as provided in Table 4-4 is used. Based on these results, the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. The lowest USE predicted for 32 EFPY is 53 ft-lb (for Lower Shell plate heat C9434-2).
F-2

GE Nuclear Energy                                                                     GE-N E-0000-0003-5526-01 a Table F-I: Upper Shelf Energy Evaluation for LaSalle Unit 2 Beltline Materials Initial        Initial            32 EFPY Test          Longitudinal   Transverse               I11T   % Decrease     32 EFPY Location          Heat     Temperature           USE           USE*          %Cu  Fluence     USEb     Transverse USE6 (OF)            (ft4b)         (ft4b)              (n/cm2 )                 (ft4b)
Plates:
Lower         C9425-1           d               102           66.3         0.12 7.,E+17        12           58 C9425-2           d               94            61.1         0.12 7.5E+17        12           54 C9434-2          40               91            59.2         0.09 7.5E+17        10           53 Lower Intermediate     C9481-10          40               n/a           95.5         0.11 7.5E+17       11            85 C9404-2           d               116           75.4         0.07 7.5E+17       8.5           69 eC960-2             40               107           69.6         0.12 7.5E+17       12            61 Welds:
Vertical:
Intermediate      3P4000           10               n/a            99          0.02 7.5E+17        8            91 Lower         3P4966           10               n/a            84         0.026 7.5E+17       8.5           77 Girth:
Lower to     I                                             IIIII Lower-I Intermediate      5P6771           10               We             61          0.04 7.5E+17       10            55 Nozzles:
LPCI         02a36W'          -10                              66          0.22 1.8E+17      12.5           58 a   Values obtained from [3]
b   Values obtained from Figure 2 of [2] for 32 EFPY 1/4T fluence c  32 EFPY Transverse USE - Initial Transverse USE &deg;[1 - (%Decrease USE /100)]
d   USE values estimated from statistical evaluation inAppendix B of [3]
e  Initial Transverse USE value obtained from baseline transverse data set [4]
f  Average of Charpy V-Notch data for %Shear = 60 F-3

GE Nuclear Energy                                       GE-NE-0000-0003-5526-01a Appendix F 

==References:==
: 1. "Fracture Toughness Requirements", Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.
: 2.  "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.
: 3. T.A. Caine, "Upper Shelf Energy Evaluation for LaSalle Units 1 and 2", GENE, San Jose, CA, June 1990 (GE Report SASR 90-07).
: 4. Letter, dated 3/16/94, G.W. Contreras (GE San Jose) to R. Willems (Oak Brook),
      "LaSalle RPV Archive Material Records Search".
F-4}}