ML041680109
| ML041680109 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/30/2004 |
| From: | Frew B, Norton K, Tilly L General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0023-5333 NEDC-33144, Rev 0 | |
| Download: ML041680109 (157) | |
Text
GE Nuclear Energy
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GE Nuclear Energy Engineering and Technology General Electric Company 175 Curtner Avenue San Jose, CA 95125 NEDO-33144 DRF 0000-0023-5333 Revision 0 Class I April 2004 Pressure-Temperature Curves For Energy Northwest Columbia Prepared by:
Verified by:
KV Norton, Engineer U Tilly, Senior Engineer OD (Frew B.D. Frew, Principal Engineer Structural Analysis and Hardware Design Approved by:
131 'Branfund B.J. Branlund, Technical Leader Fracture Mechanics and Vessel Analysis
GE Nuclear Energy NEDO-33144 Non-proprietary Version IMPORTANT NOTICE This is a non-proprietary version of the document NEDC-33144P, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((
D.
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 Energy Northwest and GE, PO #00315749, Develop New Pressure-Temperature Limit Curves and Related Work, effective 11/18/03, 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 Energy Northwest, 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, 2004
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GE Nuclear Energy NEDO-33144 GE Nuclear Energy ND-34 Non-proprietary Version 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.
Columbia is currently licensed to P-T curves for 32 EFPY [1]; the P-T curves in this report represent 22 and 33.1 effective full power years (EFPY), where 33.1 EFPY represents the end of the 40 year license, and 22 EFPY is provided as a midpoint between the current EFPY and 33.1 EFPY. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of Kic from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress.
This report incorporates a fluence [4] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190.
The P-T curves presented in this report reflect changes from those currently licensed [1].
These P-T curves have been generated to reflect a revised fluence [4]. The limiting beltline shift for the P-T curves presented herein is 350F, based upon a peak surface fluence of 7.41e17 n/cm2 for 33.1 EFPY (9.64e8 MWh) [4]. Similarly, the limiting beltline shift is 280F, based upon a
peak surface fluence of 5.12e17 n/cm2 for 22 EFPY (6.41e8 MWh) [4].
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]:
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GE Nuclear Energy NEDO-33144 Non-proprietary Version L
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 curves, the P-T curves specify a coolant 1
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 l
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 l
are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 200F/hr or less must be maintained at all times.
The P-T curves apply for both heatup and 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, Kirn 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 l
and Core Critical conditions at 22 and 33.1 EFPY.
The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, l
beltline, upper vessel and closure assembly P-T limits.
Separate P-T curves were developed for the upper vessel, beltline (at 22 and 33.1 EFPY), and bottom head for the l
Pressure Test and Core Not Critical conditions.
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GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 SCOPE OF THE ANALYSIS 3.0 ANALYSIS ASSUMPTIONS 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY
5.0 CONCLUSION
S AND RECOMMENDATIONS
6.0 REFERENCES
I 3
5 6
6 15 22 52 69
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GE Nuclear Energy NEDO-331 44 Non-proprietary Version TABLE OF APPENDICES APPENDIX A APPENDIX B APPENDIX C APPENDIX D APPENDIX E APPENDIX F APPENDIX G APPENDIX H DESCRIPTION OF DISCONTINUITIES PRESSURE-TEMPERATURE CURVE DATA TABULATION OPERATING AND TEMPERATURE MONITORING REQUIREMENTS GE SIL 430 DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)
THICKNESS TRANSITION DISCONTINUITY EVALUATION CORE NOT CRITICAL BOTTOM HEAD (CRD PENETRATION) EXAMPLE CALCULATION
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GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE COLUMBIA RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2: CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 33 FIGURE 4-3: FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 38 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
55 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
56 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 22 EFPY [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
57 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 33.1 EFPY
[20WF/HR OR LESS COOLANT HEATUP/COOLDOWN]
58 FIGURE 5-5: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 22 EFPY [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN]
59 FIGURE 5-6: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 33.1 EFPY [201F/HR OR LESS COOLANT HEATUP/COOLDOWN]
60 FIGURE 5-7: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [I 000F/HR OR LESS COOLANT HEATUP/COOLDOWN]
61 FIGURE 5-8: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100WF/HR OR LESS COOLANT HEATUP/COOLDOWN]
62 FIGURE 5-9: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 22 EFPY
[100WF/HR OR LESS COOLANT HEATUP/COOLDOWN]
63 FIGURE 5-10: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 33.1 EFPY
[100WF/HR OR LESS COOLANT HEATUP/COOLDOWN]
64 FIGURE 5-11: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 22 EFPY
[I00°F/HR OR LESS COOLANT HEATUP/COOLDOWN]
65 FIGURE 5-12: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 33.1 EFPY
[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN]
66 FIGURE 5-13: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 22 EFPY [100WF/HR OR LESS COOLANT HEATUP/COOLDOWN]
67 FIGURE 5-14: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 33.1 EFPY
[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN]
68
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GE Nuclear Energy NEDO-331 44 Non-proprietary Version TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR COLUMBIA PLATE AND FLANGE MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR COLUMBIA NOZZLE MATERIALS 12 TABLE 4-3: RTNDT VALUES FOR COLUMBIA WELD MATERIALS 13 TABLE 4-4: RTNDT VALUES FOR COLUMBIA APPURTENANCE AND BOLTING MATERIALS 14 TABLE 4-5A: COLUMBIA BELTLINE PLATE AND NOZZLE ART VALUES (22 EFPY) 18 TABLE 4-SB: COLUMBIA BELTLINE WELD ART VALUES (22 EFPY) 19 TABLE 4-6A: COLUMBIA BELTLINE PLATE AND NOZZLE ART VALUES (33.1 EFPY) 20 TABLE 4-6B: COLUMBIA BELTLINE WELD ART VALUES (33.1 EFPY) 21 TABLE 4-7:
SUMMARY
OF THE IOCFR50 APPENDIX G REQUIREMENTS 24 TABLE 4-8: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 26 TABLE 4-9: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 26 TABLE 4-10: PRESSURE TEST CRD PENETRATION K, AND (T - RTNDT) AS A FUNCTION OF PRESSURE 29 TABLE 4-11: CORE NOT CRITICAL CRD PENETRATION K, AND (T - RTNDT) AS A FUNCTION OF PRESSURE 32 TABLE 4-12: PRESSURE TEST FEEDWATER NOZZLE K, AND (T - RTNDT) AS A FUNCTION OF PRESSURE 35 TABLE 4-13: CORE NOT CRITICAL FEEDWATER NOZZLE K1 AND (T - RTNDT) AS A FUNCTION OF PRESSURE 42 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 54 ix -
GE Nuclear Energy NEDO-33144 Non-proprietary Version
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 embrittlement effects in the beltline.
Complete P-T curves were developed for 22 and 33.1 effective full power years (EFPY),
where 33.1 EFPY represents the end of the 40-year license, and 22 EFPY is provided as a midpoint between the current EFPY and 33.1 EFPY. The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. This report incorporates a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190.
The P-T curves presented in this report reflect changes from those currently licensed [1].
These P-T curves have been generated to reflect a revised fluence [4]. The limiting beltline shift for the P-T curves presented herein is 350F, based upon a peak surface fluence of 7.41e17 n/cm2 for 33.1 EFPY (9.64e8 MWh) [4]. Similarly, the limiting beltline shift is 280F, based upon a peak surface fluence of 5.12e17 n/cm2 for 22 EFPY (6.41e8 MWh) [4].
The methodology used to generate the P-T curves in this report is presented in Section 4.3. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of Kic from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. 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 GE Nuclear Energy NEDO-33144 l
Non-proprietary Version Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTNDT are documented in Section 4.1.
l Adjusted Reference Temperature (ART) is the reference temperature when including l
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 l
beltline material chemistry. The ART calculation, methodology, and ART tables for 22 and 33.1 EFPY are included in Section 4.2.
The peak ID fluence values of l
5.12 x 1017 n/cm2 (22 EFPY) and 7.41 x 1017 n/cm2 (33.1 EFPY) used in this report are 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 l
report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect each discontinuity.
l Guidelines and requirements for operating and temperature monitoring are included in Appendix C. Temperature monitoring requirements and methods are available in GE I
Services Information Letter (SIL) 430 contained in Appendix D.
Appendix E demonstrates that all reactor vessel nozzles requiring fracture toughness evaluation are I
either included in the development of the P-T curves or are outside the beltline region.
Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf l
energy (USE).
Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region and the bottom head. Finally, Appendix H provides an l
example calculation of the bounding bottom head (CRD penetration) core-not-critical (Curve B) curve.
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GE Nuclear Energy NEDO-33144 Non-proprietary Version 2.0 SCOPE OF THE ANALYSIS A detailed description of the P-T curve bases is included in Section 4.3. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of Kic from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. 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 Columbia 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 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 requiring fracture toughness evaluation are either included in the development of the P-T curves or are outside the beltline region. Appendix F provides the calculation for equivalent margin GE Nuclear Energy NEDO-33144 1
Non-proprietary Version j
analysis (EMA) for upper shelf energy (USE). Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region and bottom head.
- Finally, l
Appendix H provides an example calculation of the bounding bottom head (CRD penetration) core-not-critical (Curve B) curve.
l 1
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GE Nuclear Energy NEDO-331 44 Non-proprietary Version 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:
The hydrostatic pressure test will be conducted within the range between 1020 psig and 1050 psig [1]; the examples cited in this report are performed at 1020 psig.
For Columbia, use of either the minimum or maximum pressure has no impact on the P-T curves.
The shutdown margin, provided in the Definitions Section of the Columbia Technical Specification [1], is calculated for a water temperature of 680F.
GE Nuclear Energy NEDO-33144 Non-proprietary Version 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERA TURE 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 applicable ASME Code for the Columbia RPV is 1971 Edition with Summer 1971 Addenda. 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 600F 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 1I1, 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 600F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are 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.
10CFR50 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 NEDO-33144 l
Non-proprietary Version 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 l
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 l
NRC for generic use [11].
4.1.2 Values of Initial RTNDT and Lowest Service Temperature (LST) l To establish the initial RTNDT temperatures for the Columbia vessel per the current J
requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, forging, and for bolting 4
material LST are summarized in the remainder of this section. The initial RTNDT for all materials remain unchanged from those previously reported for Columbia with the 4
following exceptions: dropweight information was obtained and considered, resulting in a revised initial RTNDT for weld heats 3P4955, Lot 0342/3443 (tandem wire) and 4
3P4966 Lot 1214/3481 (single wire), as shown in Table 4-3.
For vessel plate material, the first step in calculating RTNDT is to establish the 50 ft-lb l
transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs [12]). For Columbia CMTRs, typically six energy values l
were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20F per ft-lb energy difference from l
50 ft-lb.
L For example, for the Columbia beltline plate heat C1272-1 in the lower shell course; the lowest Charpy energy and test temperature from the CMTRs is 26 ft-lb at 100F. The l
estimated 50 ft-lb longitudinal test temperature is:
T50L = 1 0F + [ (50 - 26) ft-lb
- 20F/ft-lb] = 580F I
The transition from longitudinal data to transverse data is made by adding 300F to the 50 ft-lb longitudinal test temperature; thus, for this case above, l
T50T = 580F + 300F = 880F.
l GE Nuclear Energy NEDO-33144 Non-proprietary Version The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5or 600F).
Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -100F. Thus, the initial RTNDT for plate heat C1272-1 is 280F.
For the Columbia beltline weld heat 3P4966 (Tandem Wire) with flux lot 1214 (contained in the lower-intermediate shell course), the CVN results are used to calculate the initial RTNDT.
The 50 ft-lb test temperature is applicable to the weld material, but the 300F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat 3P4966 / 1214 (Tandem) has a lowest Charpy energy of 28 ft-lb at 100F as recorded in weld qualification records. Therefore, T50T = 10OF +[(50 - 28)
- 20F/ft-lb] = 540F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TSOT - 60'F). For Columbia, the dropweight testing to establish NDT was -20 0F. The value of (T50T - 600F) in this example is -60F; therefore, the initial RTNDT was -60F.
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 limiting heat in the feedwater nozzle at Columbia (Heat Q2Q55W 786S-3), the NDT is -200F and the lowest CVN data is 25 ft-lb at -20 0F. The corresponding value of (T5oT - 600F) is:
(T0T - 600F) = {[-20 + (50 -25) ft-lb
- 2°F/ft-lb] + 30°F} - 60°F = 0°F.
Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5oT-600F), which is 0°F. It is noted that an initial RTNDT of -14°F was previously cited in the Columbia FSAR for the feedwater nozzle; however, the current evaluation has utilized the vessel CMTRs, resulting in the initial RTNDT of 0°F as calculated above.
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version l
In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNDT. For the bottom head dollar plate heat of Columbia (Heat B5130-2), the NDT is 100F and the lowest CVN data was 30 ft-lb at 100F.
The corresponding value of J50T - 600F) was:
l (TJOT - 600F) = { [10 + (50 - 30) ft-lb
- 20F/ft-lb ] + 300F } - 601F = 20'F.
l Therefore, the initial RTNDT was 20cF.
I 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 4
of the ASME Code Section 1II, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 600F is the LST for the bolting materials. All of the Charpy data for the Columbia closure studs met the 45 ft-lb, 25 MLE requirements at 100F. Therefore, the LST for the bolting material is 100F. The highest RTNDT in the closure flange region is 201F, for the upper shell. Thus, the higher of the LST and the RTNDT +60°F is 800F, the bolt-up limit in the closure flange region.
l The initial RTNDT values for the Columbia reactor vessel (refer to Figure 4-1 for the Columbia Schematic) materials are listed in Tables 4-1, 4-2, 4-3 and 4-4. This tabulation l
includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves.
The values presented in these tables and I
used to determine the initial RTNDT were obtained from the Columbia vessel CMTRs [12].
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GE Nuclear Energy NEDO-33144 Non-proprietary Version LPCI NOZZLE BOUNDING TOP OF ACTIVE FUEL (TAF) 366.3" BOTTOM OF ACTIVE FUEL (BAF) 216.3" 4~-~>\\\\TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL COURSE #4 SHELL COURSE #3 SHELL COURSE #2 AXIAL WELDS GIRTH WELD X
V"GRT WELDSHELL COURSE #1 BOTTOM HEAD SUPPORT SKIRT Hi,=,
Notes:
(1) Refer to Tables 4-1, 4-2, 4-3 and 4-4 for reactor vessel components and their heat identifications.
(2) See Appendix E for the definition of the beltline region.
Figure 4-1: Schematic of the Columbia RPV Showing Arrangement of Vessel Plates and Welds
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I GE Nuclear Energy NEDO-33144 Non-proprietary Version IA I
Table 4-1: RTNDT Values for Columbia Plate and Flange Materials Test Drop Component Heat or Heat I Flux Tep Charpy Energy (T5OT-60)
Weight RTNOT CmoetI Lot Temp (ft-lb)
(TF)
NOT
('F)
(°F)
Top Head & Flange Shell Closure Flange - MK 26 26-1 2V2243 BSD451 10 90 116 110
-20 10 10 Top Head Flange - MK 30 30-1 4P4740 BAS452 10 194 160 144
-20 10 10 Top Head Dollar - MK 32 32-2 C1256-3 10 58 50 41
-2 1 0 10 Top Head Side Plates - MK 32 32-1-1/3 A0141-1 10 55 65 52
-20 10 10 32-1-4/6 A0203-1 10 35 62 72 10 10 10 Shell Courses Upper Shell Plates - MK 24 Shell #4 24-1-1 C1308-2 10 40 64 60 0
1 0 10 24-1-2 C1307-2 10 30 33 36 20 1 0 20 24-1-3 C1873-2 10 45 69 70
-10 10 10 Upper Int. Plates - MK 23 Shell #3 23-1-1 C1307-1 10 60 50 60
-20 1 0 10 23-1-2 C1308-1 10 57 46 42
-4 1 0 10 23-1-3 C1302-1 10 34 42 50 1 2 1 0 12 Low-Int. Plates - MK 22 Shell #2 22-1-1 B5301-1 10 52 52 55
-20
-30
-20 22-1-2 C1336-1 10 60 44 66
-8
-30
-8 22-1-3 C1337-1 10 70 72 55
-20
-30
-20 22-1-4 C1337-2 10 62 72 82
-20
-50
-20 Lower Shell Plates - MK21 Shell #1 21-1-1 C1272-1 10 26 34 30 28
-10 28 21-1-2 C1273-1 10 30 34 35 20
-20 20 21-1-3 C1273-2 10 38 48 55 4
-30 4
21-1-4 C1272-2 10 40 42 44 0
-30 0
Bottom Head Bottom Head Dollar-MK 13 13-1 B5432-1 10 74 64 60
-20 10 10 13-2-1 B5130-1 10 35 40 31 18 10 18 13-2-2 B5130-2 10 32 33 30 20 10 20 Bottom Head Side Plates - MK 13 13-4-1,2,3 C1578-1 10 96 96 80
-20 1 0 10 13-44,5,6 A0081-1 10 95 84 90
-20 10 10 NOTE: These are minimum Charpy values.
I1 l
I I
I Ii L
I I
I, i
L I, Ii
GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-2: RTNDT Values for Columbia Nozzle Materials Heat or Heat I Flux Test Charpy Energy (TNT-GO)
Drop Weight RTIODT Component I Lot Temp C(ft4b) n ff NOT TF) lfr)
Ni Recirculation Outlet Nozzle 46-1-1 02055W327S-1 10 45 41 41
-2 10 10 46-1.2 02049W327S-2 10 25 28 25 30 10 30 N2 Recrcaulation Inlet Nozzle 49-1-1 210527 53003-01 10 39 34 25 30 10 30 49-1-2 210527 53003-01 10 60 26 31 28 10 28 49-1-3 210527 53003-01 10 26 25 32 30 10 30 49-1.4 210527 53003-01 10 30 50 50 20 10 20 49-1-5 211319 53003-4A 10 37 30 35 20 10 20 49-1-6 211319 53003-IA 10 39 28 41 24 10 24 49-1-7 211319 53003-1R 10 104 80 76
-20 10 10 49-1-8 211319 53003-1R 10 70 44 42
-4 10 10 49-1-9 211319 53003-1R 10 53 50 68
-20 10 10 49-1.10 211319 53003-1R 10 70 64 91
-20 10 10 N3 Steam Outlet Nozzle 53-1-1 Q2054W328S-1 1 0 49 42 36 8
10 1 0 53-1-2 02054W328S-2 10 32 46 25 30 10 30 53-1-3 02063WR328S 10 54 29 44 22 10 22 53-1-4 02057W328S-4 10 33 35 39 14 10 14 N4 Feedwater Nozzle 56-1.1 02055W786S-1
-20 58 31 61
-12
-20
-12 56-1-2 02055W 786S-2
-20 64 52 37
-24
-20
-20 56-1-3 02055W 786S-3
-20 25 65 57 0
-20 0
56-1-4 02055W 786S4
-20 48 58 46 42
-20
-20 56-1-5 Q2055W 786S-5
-20 78 75 53
-50
-20
-20 56-1-6 02055W 786S4-
-20 55 55 43
-36
-20
-20 N5 Core Spray Nozzle (Low Pressure) 60-1 02055W 787S
-20 72 43 55
-36
-20
-20 N6 Residual Heat Removal / Low Pressure Core Isolation 64-1-1 02055W 790S-1
-20 73 55 74
-50
-20
-20 64-1-2 02Q55W 790S-2
-20 48 64 45
-40
-20
-20 64-1-3 02055W 790S-3
-20 66 56 48 46
-20
-20 N7 Head Spray Nozzle 68-2 02055W 173T
-20 30 50 65
-10
-20
-10 N8 Vent Nozzle 70-1 02030W 171T 10 48 67 74
-16 10 10 N9 Jet Pump Instrumentaton Nozzle 72-1-1 210527 Lot 1 10 66 50 46
-12 10 10 72-1-2 210527 Lot 1
10 35 41 80 10 10 10 N10 Control Rod Drive Hyd System Return Nozzle "I 75-1 02034W 789S
-20 61 29 68
-8
-20
-8 N1i Core Difterental Pressure & Liquid Control Nozzle Alloy 600 79-1 NX4256 6212 N12 Instrumentaton Nozzle 82-1-1.2.3 219972 Lot 1
.20 60 90 230
-50 40 40 82-1-7 718259 65363
-20 240 240 240
-50
-20
-20
-N13 Instrumentation Nozzle 82-1-5.6 219972 Lot 1
-20 60 90 230
-50 40 40 N14 Instrumentation Nozzle 85-1-1/4 219972 Lot 1
-20 120 240 240
-50 40 40 N15 Drain Nozzle MK 87-1 B12W295T
-20 33 20 30 10
-20 10 N16 Core Sprav Hioh Pressure"2' 88-1 02055W 788S 40 40 N17 Seal Leak Detector Alloy 600 264 NX4104 6242 N18 Top Head Spare Nozzle 92-2 02059W 172T 10 31 36 71 18 10 18 NOTES:
- 1. The N10 nozzle has been capped off.
- 2. CMTR infomationwasnot available for the N16 nozzle; thus the Purchase Specification requirement was used for evaluation of this component.
These are minimum Charpy values.
I GE Nuclear Energy NEDO-33144 Non-proprietary Version I
Table 4-3: RTNDT Values for Columbia Weld Materials Test Drop Heat or Heat I Welds where Heat is Test Charpy Energy (T
Weight RTWDT Flux I Lot Present Weld Type Temp (ft-lb)
(eF)
('F)
Beltline Welds 492L4871 /A422827AF AB (SMAM)
-20 78 81 82
-80
-50 04T931 /A423B27AG AB (SMAW)
-20 61 63 84
-80
-50 5P6756 /0342 - 3447 AB Single Wire (SAW) 10 72 76 77
-50
-60
-50 5P6756 10342 - 3447 AB Tandem Wire (SAM) 10 72 76 77
-50
-50
-50 3P4955 10342 - 3443 AB Single Wire (SAW) 10 33 37 45
-16
-40
-16 3P4955 /0342-3443121 AB Tandem Wire (SAW) 10 47 49 49
-44
-20
-20 04P04610217A27A BA, BB, BD, BF, BH (SMAW)
-20 34 36 39
-48
-48 07L669 / K004A27A BA, BB (SMAW) 10 50 50 54
-50
-50 3P4966 / 1214 - 3482 BA, BB, BC,BD Single Wire (SAW) 10 40 59 63
-30
-30 3P4966 / 1214 - 3482 BA. BB, BC,BD Tandem Wire (SAVV) 10 49 65 67
-48
-48 C3L46C I J020A27A BB, BC. BD (SMAW) 10 35 39 40
-20
-20 08M365 / G128A27A BB (SMAMW 10 49 50 51
-48
-48 09L853 I A11 1A27A BC (SMAMW 10 78 78 79
-50
-50 3P4966 / 1214 - 3481 121 BE, BF, BG, BH Single Wire (SAW) 10 38 38 39
-26
-20
-20 3P4966 / 1214 - 3481 BE, BF, BG, BH Tandem Wire (SAW) 10 28 63 75
-6
-20
-6 05P018 / D211A27A BF (SMAW)
-20 29 30 31
-38
-38 624063 1 C228A27A BG (SMAW)
-20 37 40 51
-54
-50 624039/ D224A27A BG (SMAW)
-20 28 33 34
-36
-36 6240391 D205A27A BH (SMAW)
-20 41 44 49
-62
-50 Nonbeitline Welds See note 1 10
-1
-~
-1 Weld Identification Table AB Shell 1 to Shell 2 Girth Weld BA, BB, BC, BD Shell 1 Vertical Welds
,BE, BF, B3G, BH Shell 2 Vertical Welds Notes:
- 1. CMTRs for the non-beltline welds are not available. The Purchase Specification requirement is used as the limiting RTNDT.
- 2. Dropweight information was obtained and considered, resulting in a change to the initial RTNDT for these materials.
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GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-4: RTNDT Values for Columbia Appurtenance and Bolting Materials Test Drop Heat or Heat Test Charpy Energy
{
T.,-60)
Weight RTmo?
Component I Flux I Lot TeFp (ft4b)
(F)
(*F)
(F) rF)
Misc Appurtenances:
Skirt Knuckle 9-1-1/4 C9285-2A 10 56 54 61
-20 10 10 Shroud Support 17-1-114 Alloy 600 20-1-1& 2. 20-1-3&4. 20-2 Refueling Bellows Ring - MK 43 43-1-1/6 R0543-1 40 57 50 61
.20
-20
-20 Refueling Bellows Bar 26-2.4. 5 C2924-14B 10 28 24 23 34 10 34 Stabilizer Bracket 95-1-1/8 A1322-2B 50 67 50 51
-10
-20
-10 Guide Rod Bracket Stainless Steel 98-1-1/2 651542 Steam Dryer Support Bracket Inconel 100-1-1/4 NX4055-G Steam Dryer Hold Down Bracket 102-1-1/4 A0141-1 10 55 65 52
-20 10 10 Feedwater Sparger Brackets Stainless Steel 104-1-1/12 131146 Core Spray Brackets Stainless Steel 108-1-1/8 61587 Surveillance Bracket 106-1-1/3, Stainless Steel 106-2-1/3 150368 CRD Penetrations 14 Inconel Thermocouple Clamping Pad and End Pad 44-1. 44-2 A8879-1B 40 30 30 33 50 40 50 Lifting Lugs 40-1-1/4 C1256-3 10 58 50 41
-2 10 10 Jet Pump Riser Support Pads Stainless Steel Test Charpy Energy Min Lit Exp LST Component He Temp (ft-lb)
(mils)
(F)
STUDS:
Closure 35-1 81741 10 46 45 49 27 10 35-1 81646 10 47 49 45 26 10 NUTS:
Closure 36-1 61056 10 47 47 45 26 10 BUSHINGS:
Stud Bushings 94-1 61056 10 47 47 45 26 10 WASHERS:
36-2 61056 10 47 47 45 26 10 NOTE: These are minimum Charpy values.
I1 GE Nuclear Energy NEDO-33144 Non-proprietary Version I
I 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 (RG1.99) provides the methods for determining the ART.
The RG1.99 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 beltline plates and welds was performed and is summarized in Tables 4-5 and 4-6 for 22 and 33.1 EFPY, respectively.
4.2.1 Regulatory Guide 1.99, Revision 2 (RG1.99) 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 RG1.99, the SHIFT equation consists of two terms:
I I
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SHIFT = ARTNDT + Margin
- where, ARTNDT = [CF]. f(0.28-010109f)
Margin = 2(al2 + Ga2) 0 5 CF = chemistry factor from Tables 1 or 2 of RG1.99 f = AT fluence / 1019 Margin = 2(0y2 + GA2 ) 0.5 as,
= standard deviation on initial RTNDT, which is taken to be 0F.
CA
= standard deviation on ARTNDT, 280F for welds and 17'F for base material, except that HA need not exceed 0.50 times the ARTNOT value.
-I J1 A
A I
ART = Initial RTNDT + SHIFT 1.
The margin term CA has constant values of 170F for plate and 281F for weld as defined in RG1.99.
However, GA 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 Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of cay is taken to be 0F for the vessel plate and weld materials.
A I.
I
GE Nuclear Energy NEDO-33144 Non-proprietary Version 4.2.1.1 Chemistry The vessel beltline chemistries were obtained from all known available sources of data for the beltline materials, including the Certified Material Test Reports (CMTR),
Surveillance Capsule Data [13] and best estimates [1].
The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG1.99, to determine a chemistry factor (CF) per Paragraph 1.1 of RG1.99 for welds and plates, respectively.
The chemistry values for plate heat B5301-1 were developed using all available data, including CMTR and surveillance capsule test results. Table 3-1 of [13] cites CMTR unirradiated chemistry values as 0.14% Cu and 0.50% Ni. Table 4-1 of [13] provides test results from irradiated specimens obtained from the Columbia surveillance capsule as shown in the table below.
%Cu
%Ni 0.12 0.51 0.11 0.49 0.11 0.47 The irradiated test results were averaged, the results of which were averaged with the CMTR values noted above.
The resulting chemistry for this heat is 0.13% Cu and 0.495% Ni, which was rounded to 0.50%. Table 8-1 of [13] used chemistry values of 0.13% Cu and 0.49% Ni, however, this report uses 0.13% Cu and 0.50% Ni based upon the calculations described.
For weld heats 3P4955 and 3P4966, the chemistry reflects ISP vessel best estimate data provided in [1]. Heat 5P6756 is the surveillance weld material as defined by the Integrated Surveillance Program (ISP); chemistry and adjusted CF information defined by this program was provided by [1].
For this material, an adjusted CF used in calculating the adjusted reference temperature for 22 and 33.1 EFPY was obtained by multiplying the ISP least-squares fit CF developed in accordance with RG1.99 as defined by BWRVIP-102 [5] by the ratio of the RG1.99 CF for the ISP best estimate GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
vessel chemistry to the RG1.99 CF for the ISP surveillance chemistry. This results in an adjusted CF of: 119.72 * (108 / 82) = 157.68.
-I 4.2.1.2 Fluence The peak fluence for the RPV inner surface, used for determination of the P-T curves, is I
7.41 e17 n/cm2 for 33.1 EFPY (9.64e8 MWh).
For 22 EFPY (6.41e8 MWh), the peak fluence for the RPV inner surface is 5.12e17 n/cm2.
The basis for all fluence values l
used in this report is contained in [4]. Calculations for 1/4T fluence are performed in accordance with RG1.99 [7].
4.2.2 Limiting Beitline Material The limiting beltline material signifies the material that is estimated to receive the J
greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, RG1.99 was applied to compute ART.
J Tables 4-5 and 4-6 list values of beltline ART for 22 and 33.1 EFPY, respectively.
GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-5a: Columbia Beltline Plate and Nozzle ART Values (22 EFPY)
Lower Shell #1 Thickness In inches"' = 9.50 Ratio Peakl Location -
0.428 Thickness In inchesm - 6.1875 Thickness In inchestOt = 6.1875 Lower4ntermediate Shell #2 Ratio Peak/ Location =
1.000 N6 Nozzle Ratio Peak/ Location 0.545 22 EFPY Peak I.D. fluence =
2.19E+17 nlcm'2 22 EFPY Peak 1/4 T fluence =
1.24E+17 nlrcm2 22 EFPY Peak 114 T fluence -
1.24E+17 nlrm^2 22 EFPY Peak I.D. fluenoe =
5.12E+17 nlrm^2 22 EFPY Peak 114 T fluence =
3.53E+17 nlcm^2 22 EFPY Peak 1/4 T fluenre =
3.53E+17 nlrm^A2 22 EFPY Peak I.D. fluence =
2.79E+17 n/rcm2 22 EFPY Peak 114 T fluence =
1.92E+17 n/lm^2 22 EFPY Peak 114 T fluence =
1.92E+17 nlcm^2 Initial 1/4 T 22 EFPY 01 22 EFPY 22 EFPY COMPONENT HEAT OR HEATILOT
%Cu
%Ni CF (1)
RTndt Fluence A RTndt Margin Shift ART
- F nrcmA2
- F
- F
- F
- F PLATES:
Lower Shell Mk 21-1-1 C1272-1 0.15 0.60 110 28 1.24E+17 14 0
7 14 28 56 Mk 21-1-2 C1273-1 0.14 0.60 100 20 1.24E+17 13 0
6 13 25 45 Mk 21-1-3 C1273-2 0.14 0.60 100 4
1.24E+17 13 0
6 13 25 29 Mk 2-1-4 C1272-2 0.15 0.60 110 0
1.24E+17 14 0
7 14 28 28 Lower4ntermnediate Shell Mk 22-1-1 B5301-1 0.13t" 0.50 "I 88
-20 3.53E+17 21 0
11 21 42 22 Mk 22-1-2 C1336-1 0.13 0.50 88
-8 3.53E+ 17 21 0
11 21 42 34 Mk 22-1-3 C1337-1 0.15 0.51 105
.20 3.53E+ 17 25 0
13 25 51 31 Mk 22-1-4 C1337-2 0.15 0.51 105
-20 3.53E+ 17 25 0
13 25 5t 3 1 NOZZLES:
N6 02055W790S'4 )
0.11 0.76 76.4
-20 1.92E+17 1 3 0
6 13 26 6
Notes:
- 1. Determined per RG1.99 Tables 1 and 2.
- 2. Minimum thicknesses per CBIN Drawing #1 Rev. 8 IReference 1 of Appendix Al were used for conservatism. Further, dad thickness was conservatively onitted.
- 3. B5301-1 Is the surveilance plate material; this chemistry represents the average of the Columbia Surveillance Capsule Report 1131 chemistry test results averaged with the baseline CMTR values.
- 4. Chemistry obtained from Columbia CMTRs.
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version I
Table 4-5b: Columbia Beltline Weld ART Values (22 EFPY)
I Lower Shell #1 I =
0.428 Thickness in inches"' = 9 50 Thickness in inchesct = 6.1875 Thickness in inches"' = 6.1875 Ratio Peak/ Location Lower-Intermediate Shell #2 Ratio Peak/ Location =
1.000 Girth Weld Between Shell #1 and Shell #2 Ratio Peakl Localion =
0.428 22 EFPY Peak I. D. fluence =
2.19E+17 22 EFPY Peak 114 T fluency =
1.24E+17 22 EFPY Peak 114 T fluence =
1.24E+17 22 EFPY Peak ID. ftuence =
5.12E+17 22 EFPY Peak 114 T fiuence =
3.53E+17 22 EFPY Peak 1/4 T fluence =
3.53E+17 nlcm12 nlcnm2 nVcm12 n/cm12 n/cmn2 n/cm12 I
22 EFPY Peak I.D. fluence =
22 EFPY Peak 114 T fluence =
22 EFPY Peak 1/4 T fluence =
2.19E+17 n/cnm2 1.51E+17 nlac'2 1.51E+17 n/cmn12 I
Adjusted Initial 1/4 T 22 EFPY 22 EFPY 22 EFPY COMPONENT HEAT OR HEATILOT"
%Cu
%Ni CF('
CF RTndl Fluence A RTndt Margin Shift ART
.F n/cm^2
.F
- F
.F
- F WELDS:
Lower Vertical BA, BB, BD 04P046/D217A27A 0.06 0.90 82
-48 1.24E+17 10 0
5 10 21
-27 BA, BB 07L669/K004A27A 0.03 1.02 41
-50 1.24E+17 5
0 3
5 10
-40 BA, BB, BC.BD 3P4966 / 1214 - 3482 (S) 0.025"' 0.913
34
-30 1.248+17 4
0 2
4 9
-21 BA, BB, BCBD 3P4966 1 1214 - 3482 (T) 0.025)
0.913'"
34
-48 1.24E+17 4
0 2
4 9
-39 BB, BC, BD C3L46C IJ020A27A 0.02 0,87 27
-20 1.24E+17 3
0 2
3 7
-13 BB 08M3651G128A27A 0.02 1.10 27
-48 1.24E+17 3
0 2
3 7
-41 BC 09L853 1A11 1A27A 0.03 0.86 41
-50 1.24E+17 5
0 3
5 10 40 Lower-intermediate Vertical BE, BF, BG, BH 3P4966 / 1214-3481 (S) 0,025(3' 0.913
34
-20 3.53E+17 8
0 4
8 16
-4 BE, BF. BG, BH 3P4966/ 1214 - 3481 (T) 0 025'3' 0 913'3' 34
-6 3.53E+17 8
0 4
8 16 10 BF, BH 04P046 I D217A27A 0 06 0.90 82
-46 3.53E+17 20 0
10 20 40
-8 BF 05P018 I D21 1A27A 0.09 0 90 122
-38 3.53E+17 29 0
15 29 59 21 BG 6240631 C228A27A 0 03 1.00 41
-50 3.53E+17 10 0
5 10 20
-30 BG 6240391D224A27A 007 1.01 95
-36 3.53E+17 23 0
11 23 46 10 BH 6240391 D205A27A 0.10 092 134
-50 3.53E+17 32 0
16 32 65 15 Girth AB 492L4871 /A422B27AF 0.03 0.98 41
-50 1.51E+17 6
0 3
6 12
-38 AB 04T931 1 A423B27AG 0.03 1.00 41
-50 1.51E+17 6
0 3
6 12
-38 AB 5P675610342 - 3447 0.0863' 0.936'3' 108 157.68
-50 1.51E+17 23 0
11 23 45
-5 AB 3P4955/0342-3443(S) 0.027'3' 0.921"'3 37
-16 1.51E+17 5
0 3
5 11
-5 AB 3P4955/0342-3443(T) 0.027'3' 0.921'3' 37
-20 1 51E+17 5
0 3
5 11
-9 Notes:
- 1. Determined per RG1.99 Tables 1 and 2.
- 2. Minimum thicknesses per CBIN Drawing #1 Rev. 8 [Reference 1 of Appendix Al were used for conservatism. Further, dad thickness was conservatively omined.
- 3. Chemistry reflects ISP vessel best estimate data provided by [1]. Adjusted CF for 5P6756 determined as defined in Section 4.2.1.1.
- 4. For weld materials, S = Single Wre, T = Tandem Wire.
-I I
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I
-l GE Nuclear Energy NEDO-331 44 Non-proprietary Version Table 4-6a: Columbia Beltline Plate and Nozzle ART Values (33.1 EFPY)
Lower Shell #1 Ratio Peak/ Location =
0.416 Thickness in inchesm = 9.50 Thickness in Inchesm = 6.1875 Thickness in Inchesm = 6.1875 Lower-Intermediate Shell #2 Ratio Peak/ Location =
1.000 33.1 EFPY Peak l.D. fluence =
3.09E+17 33.1 EFPY Peak 1/4 T fuence =
1.75E+17 33.1 EFPY Peak 1t4 T fluence =
1.75E+17 33.1 EFPY Peak I.D. fluence =
7.41E+17 33.1 EFPYPeak1/4Tftuence 5.11E+17 33.1 EFPYPeakl14Tftuence-5.11E+17 33.1 EFPY Peak I.D. fluence =
4.07E+17 33.1 EFPY Peak 1/4 T fuence 2.81E+17 33.1 EFPY Peak 1/4 T fluence =
2.81E+17 nvrm^2 nvcm^2 nvrcm^2 nvrm^2 nvrm^2 nrcm^2 rVrm^2 nvcm^2 nr m^2 N6 Nozzle Ratio Peak/ Location=
0.549 Initial 114 T 33.1 EFPY a,
CA 33.1 EFPY 33.1 EFPY COMPONENT HEAT OR HEAT/LOT
%Cu
%Ni CF RTndt Fluence A RTndt Margin Shift ART
.F n/cm`2
- F
- F
- F
.F PLATES:
Lower Shell Mk 21-1-1 C1272-1 0.15 0.60 110 28 1.75E+17 17 0
9 17 35 63 Mk 21-1-2 C1273-1 0.14 0.60 100 20 1.75E+17 16 0
8 16 32 52 Mk 21-1-3 C1273-2 0.14 060 100 4
1.75E+17 16 0
8 16 32 36 Mk 21-1-4 C1272-2 0.15 0.60 110 0
1.75E+17 17 0
9 17 35 35 Lower4ntermedIate Shell Mk22-1-1 B5301-1 0.13"'
0.50 88
-20 5.11E+17 26 0
13 26 52 32 Mk 22-1-2 C1336-1 0.13 0.50 88
-8 5.11E+17 26 0
13 26 52 44 Mk22-1-3 C1337-1 0.15 0.51 105
-20 5.11E+17 31 0
16 31 62 42 Mk 22-1-4 C1337-2 0.15 0.51 105
-20 5.11E+17 31 0
16 31 62 42 NOZZLES:
N6 02055W79OS' 4'
0.11 0.76 76.4
-20 2.81E+17 16 0
8 16 32 12 Notes:
- 1. Determined per RG1.99 Tables 1 and 2.
- 2. MKnisum thicknesses per CBIN Draveng i1 Rev. 8 [Reference 1 of Appendix Al were used for conservatism. Further, dad thickness was conservatively omitted.
- 3. B5301-1 Is the surveillance plate matenal; this chemistry represents the average of the Columbia Surveillance Capsule Report 113] chemistry test results averaged with the baseline CMTR values.
- 4. Chemistry obtained from Columbia CMTRs.
GE Nuclear Energy NEDO-33144 Non-proprietary Version I
Table 4-6b: Columbia Beltline Weld ART Values (33.1 EFPY)
I Lower Shell 81 0416 Thickness in inches" 2 ' = 9650 Thickness in inches" 2' = 6.1875 Thickness in inches" 2 ' = 6.1875 Ratio Peakl Location =
Lower-Intermediate Shell #2 Ratio Peaki Location =
1.000 Girth Weld Between Shell #1 and Shell #
33.1 EFPY Peak I1D. ttuence =
3.09E+17 33.1 EFPY Peak 1/4 T ftuence =
1.75E+17 33.1 EFPY Peak 1/4 T fluence =
1-75E+17 33.1 EFPY Peak I.0. fluence =
7.41E+17 33.1 EFPYPeak1/4Tfluence=
5.11E+17 33.1 EFPY Peak 1/4 T ftuence =
5.11E+17 n/cmr2 n/cm12 n/cm12 n/cm12 n/cm12 n/cm12 I
Ratio Peak/ Location =
0 416 33.1 EFPY Peak.D. fluence =
33.1 EFPY Peak 114 T fluence =
33.1 EFPY Peak 114 T ftuence =
3.09E1+17 n/cm'2 2.13E+17 n/cm12 2-13E+17 nIcm12 I
Adjusled Initial 114 T 33.1 EFPY n,
33.1 EFPY 33.1 EFPY COMPONENT HEAT OR HEAT/LOT 141
%CU
%Ni CF CF RTndt Fluence A RTndt Margin Shift ART
_F n/cm^2 F
-F F
F WELDS:
Lower Vertical BA, BB. BD 04P046/ 0217A27A 0 06 0.90 82
-48 1.75E+17 13 0
6 13 26
-22 BA, BB 07L669 / K004A27A 0 03 1.02 41
-50 1.75E+17 6
0 3
6 13
-37 BA, BB, BC0.B 3P4966 1214 - 3482 (S) 0.0250) 0191313) 34
-30 1.75E+17 5
0 3
5 11
-19 BA, BB, BC.BD 3P4966 / 1214 - 3482 (T) 0 025131 0.913'3' 34
-48 1.75E+17 5
0 3
5 11
-37 BB, BC, BD C3L46C/J020A27A 0.02 0.87 27
-20 1.75E+17 4
0 2
4 9
-11 BB 08M365 I G128A27A 0 02 1.10 27
-48 1.75E+17 4
0 2
4 9
-39 BC 09L853 /A111A27A 0.03 0.86 41
-50 1 75E.17 6
0 3
6 13
-37 Lower-intermediate Vertical BE, BF, 8G, BH 3P4966 / 1214 - 3481 (S) 0.025(31 0.913(3) 34
-20 5.11E+17 10 0
5 10 20 0
BE. BF, BG, BH 3P4966/ 1214 -3481 (T) 0.02503) 0.913(3' 34
-6 5.11E+17 10 0
5 10 20 14 BF, BH 04P0461D217A27A 0.06 0.90 82
-48 5.11E+17 24 0
12 24 49 1
BF 05P018 I D21 1A27A 0.09 0.90 122
-38 5.116E+17 36 0
18 36 72 34 BG 6240631 C228A27A 0.03 1.00 41
-50 5.11E+17 12 0
6 12 24
-26 BG 6240391 D224A27A 0.07 1.01 95
-36 5.11E+17 28 0
14 28 56 20 BH 624039 /D205A27A 0.10 0.92 134
-50 5.11E+17 40 0
20 40 79 29 Girth AB 492L4871 IA422B27AF 0.03 0.98 41
-50 2.13E+17 7
0 4
7 15
-35 AB 04T931 IA423B27AG 0.03 1.00 41
-50 2.13E+17 7
0 4
7 15
-35 AB 5P6756/0342-3447 0 08(3) 0936's) 106 157.68
-50 2.13E617 28 0
14 28 56 6
AB 3P4955 /0342 - 3443 (S) 0.02713) 0.921'3) 37
-16 2.13E+17 7
0 3
7 13
-3 AB 3P4955 10342 - 3443 IT) 0027f31 0 921<3' 37
-20 2.13E+17 7
0 3
7 13
-7 Notes:
- 1. Determined per RG1.99 Tables 1 and 2.
- 2. Minimum thicknesses per CBIN Drawing #1 Rev. 8 [Reference 1 of Appendix A] were used for conservatism. Further, clad thickness was conservatively omitted.
- 3. Chemistry reflects ISP vessel besestimate data provided by (1] Adjusted CF for 5P6756 determined as defined in Section 4.2.1.1.
- 4. For weld materials. S = Single Wire, T = Tandem Wire.
I I
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I.
-I
-I
].
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 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 to which a pressure-retaining component may be subjected over its service lifetime.
The ASME Code (Appendix G of Section Xl [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 portions 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 100°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]. The bounding transients used to develop the curves GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 200F/hr or less must be maintained at a
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 l
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 l
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 J
because irradiation effects cause the allowable toughness, Kir, 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 l
is provided in Table 4-7.
1 GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-7: Summary of the 1 OCFR50 Appendix G Requirements
.. 6 -..
e'.A.- ;
m,,.-
p
- .e,-n-Operating Condition and Pressure
.Minimum Temperature Requirement I. 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 + 600F*
- 2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 900F II. 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 + 600F*
- 2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 120'F
- b. Core critical - Curve C
- 1. At < 20% of preservice hydrotest Larger of ASME Limits + 400F 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 + 400F or of pressure a.2 + 400F or the minimum permissible temperature for the inservice system hydrostatic pressure test
- 60'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 10CFR50 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.
[I GE Nuclear Energy NEDO-33144 Non-proprietary Version 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline Regions I
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. Non-beltline components include nozzles (see Appendix E), 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-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The BWR/6 stress analysis bounds for BWR/2 through BWR/5 designs, as will be demonstrated in the following evaluation. The analyses took into account mechanical loading and anticipated thermal J
transients that bound BWR/2 through BWR/5 designs. Transients considered include all normal and upset transients such as 1000F/hr start-up and shutdown, SCRAM, and loss
]
Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used l
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 l
the limiting BWR/6 components:
the feedwater nozzle (FW) and the CRD penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-8 and 4-9.
I,1
GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-8: Applicable BWR/5 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification.
tion
.J 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 Water Level Instrumentation Nozzle Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Table 4-9: 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 Shell**
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, because separate bottom head P-T curves are provided to monitor the bottom head.
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 Columbia 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 GE Nuclear Energy NEDO-33144 Non-proprietary Version l
generic value was adapted to the conditions at Columbia by using plant specific RTNDT values for the reactor pressure vessel (RPV).
The presence of nozzles and CR0 l
penetration holes in 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 l
stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.
))
I An evaluation was performed for the bottom head wall thickness transition discontinuity l
located between the bottom head dollar plate and torus.
Appendix G of this report contains a detailed description of this evaluation. It was concluded that the discontinuity l
is bounded by the bottom head P-T curve developed in the following sections, and no further adjustment was required.
l 4.3.2.1.1 Pressure Test - Non-Betline, Curve A (Using Bottom Head)
In a ((
)) finite element analysis ((
)), the CRD penetration region was l
modeled to compute the local stresses for determination of the stress intensity factor, K1.
The ((
)) evaluation was modified to consider the new requirement for Mm l
as discussed in ASME Code Section Xl Appendix G [6] and shown below. The results of that computation were K, = 143.6 ksi-in"2 for an applied pressure of 1593 psig (1563 psig I
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 840F. ((
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 1]
The limit for the coolant temperature change rate is 20'F/hr 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, tin = 2.83. The resulting value obtained was:
Mm = 1.85 for.fti <2 Mm = 0.926 t-for 2<ft<3.464 = 2.6206 Mm = 3.21 for ti>3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and KIb is calculated from the equation in Paragraph G-2214.2 [6]:
Kim = Mm - cpm = ((
Kib = (2/3) Mm
- Opb = ((
)) ksi-in1 2
)) ksi-in"2 The total K, is therefore:
K, = 1.5 (Kim+ KIb) + Mm * (sm + (2/3) *sb)
= 143.6 ksi-ini' GE Nuclear Energy NEDO-33144 I
Non-proprietary Version l
This equation includes a safety factor of 1.5 on primary stress. 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]:
(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 840F I
I I
The generic curve was generated by scaling 143.6 ksi-in"2 by the nominal pressures and calculating the associated (T - RTNDT) as shown in Table 4-10.
Table 4-10: Pressure Test CRD Penetration K, and (T - RTNDT) as a Function Of Pressure I
I1 I
Nominal Pressure K1 T - RTNDT (psig)
(ksi-in"2)
(OF) 1563 144 84 1400 129 77 1200 111 66 1000 92 52 800 74 33 600 55 3
400 37
-88 I.
I, I,
1.
The highest RTNDT for the bottom head plates and welds is 20'F, as shown in Tables 4-1 and 4-3. ((
f I I
GE Nuclear Energy NEDO-33144 Non-proprietary Version 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 cy (na) "
(4-1)
The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, K, is proportional to RI(t)"2. The generic curve value of R/(t)"2, based on the generic BWR/6 bottom head dimensions, is:
Generic:
R / (t) In2= 138 / (8) In = 49 inch"2 (4-2)
The Columbia-specific bottom head dimensions are R = 130.25 inches and t =7.3125 inches minimum [19], resulting in:
GE Nuclear Energy NEDO-33144 Non-proprietary Version l
I Columbia specific:
R / (t) "/2 = 130.25 / (7.3125)1/2 = 48 inch"2 (4-3)
Since the generic value of R/(t) 1/2 is larger, the generic P-T curve is conservative when applied to the Columbia bottom head.
4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beltline 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. ((
I
]1 j1 1.I.
1.
1]
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 K, value for the core not critical condition is (143.6 / 1.5) 2.0 = 191.5 ksi-in"2.
Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the Kjc equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:
(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02
- 1.
GE Nuclear Energy NEDO-33144 Non-proprietary Version (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102'F The generic curve was generated by scaling 192 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT) as shown in Table 4-11.
Table 4-11: Core Not Critical CRD Penetration Function of Pressure K, and (T-RTNDT)asa Nominal Pressure K --
T - RTNDT (psig)
(ksi-inl/2)
(0F) 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49
-14 The highest and 4-3. ((
RTNDT for the bottom head plates and welds is 20'F, as shown in Tables 4-1 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 Table 4-9 and Appendix A).
With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded.
Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.
GE Nuclear Energy NEDO-33144 1
Non-proprietary Version l
((
l I
I1 I.
I.
I.
I 11 I.
4.3.2.1.3 Pressure Test - Non-Beitline 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"2 for an applied pressure of 1563 psig preservice hydrotest pressure. ((
))
The I
GE Nuclear Energy NEDO-33144 Non-proprietary Version 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 Xl). 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 K, is shown below using the BWR/6, 251-inch dimensions:
Vessel Radius, R, 126.7 inches Vessel Thickness, tv 6.1875 inches Vessel Pressure, Pv 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-175 is 1.4 where:
a = % ( t 2 + ti 2)112
=2.36 inches tr = thickness of nozzle
7.125 inches t,
thickness of vessel
= 6.1875 inches r, = apparent radius of nozzle
= r1 + 0.29 r,=7.09 inches r1 = actual inner radius of nozzle
= 6.0 inches rc = nozzle radius (nozzle corner radius)
= 3.75 inches Thus, a/rn = 2.36 / 7.09 = 0.33.
The value F(a/rr), taken from Figure A5-1 of WRC Bulletin 175 for an a/r, of 0.33, is 1.4.
Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (7a) "2
- F(a/rn):
Nominal K, = 1.5 - 34.97 * (n
- 2.36)1/2
- 1.4 = 200 ksi-in"2 The method to solve for (T - RTNDT) for a specific K, is based on the K," equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:
(T - RTNDT) = In [(KI - 33.2) / 20.734] /0.02 (T - RTNDT) = In [(200 - 33.2) / 20.734] / 0.02 (T-RTNDT) = 104.20F GE Nuclear Energy NEDO-33144 Non-proprietary Version I
J I1 j1 1]
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), ff
))
I1 I
I 1.
-1.I.
))
1.
The highest RTNDT for the feedwater nozzle materials is 0F as shown in Table 4-2.
However, the RTNDT was increased to 340F to consider the stresses in the core spray nozzle together with the initial RTNDT as described below. The generic pressure test P-T curve is applied to the Columbia feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 340F.
I I
GE Nuclear Energy NEDO-33144 Non-proprietary Version
((
JI Second, the P-T curve is dependent on the K, value calculated. The Columbia specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K, are shown below:
Vessel Radius to base metal, R, Vessel Thickness, tv Vessel Pressure, P, 125.7 inches 6.3125 inches 1563 psig Pressure stress: a = PR / t = 1563 psig - 125.7 inches / (6.3125 inches) = 31,124 psi.
The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding CT
= 34.09 ksi. The factor F (alrn) from Figure A5-1 of WRC-175 is determined where:
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version I
a =
tn =
tv=
r=
r =
- r. =
%/ ( tn 2 + tv 2)1/2 thickness of nozzle thickness of vessel apparent radius of nozzle actual inner radius of nozzle nozzle radius (nozzle corner radius)
=2.38 inches
= 7.125 inches
= 6.3125 inches
= ri + 0.29 rc=6.798 inches
= 6.0 inches
= 2.75 inches I
I I
Thus, a/rn = 2.38 / 6.80 = 0.35.
The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.35, is 1.4.
Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 cy (ma) 1/2 F(a/rn):
Nominal K, = 1.5 34.09 (-
2.38)1/2 1.4 = 195.8 ksi-in"2 3]
4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Beltline Curve B (Using Feedwater Nozzle/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 feedwater flow that is colder relative to the 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 400F feedwater injection, which is equivalent to hot standby, as seen in Figure 4-3.
I 1.
1.
I 1
I I
I t
GE Nuclear Energy NEDO-33144 Non-proprietary Version 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 (7Tra)/2 F(a/r,)
(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 NEDO-33144 Non-proprietary Version 1
Finite element analysis of a nozzle corner flaw was performed to determine appropriate values of F(a/rn) for Equation 4-4.
These values are shown in Figure A5-1 of I
WRC Bulletin 175 [15].
1 The stresses used in Equation 4-4 were taken from ((
)) design stress reports for the feedwater nozzle. The stresses considered are primary membrane, apm, and primary I
bending, Cpb.
Secondary membrane, csm, and secondary bending, cYSb, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, K1,:
I Kis = Mm (asm + (2/3)
Gsb)
(4-5)
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. K1p and K15 are added to I
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.
l Once K, was 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 K1, 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 [(K, - 33.2) / 20.734] / 0.02 (4-6)
Example Core Not Critical Heatup/Cooldown Calculation l
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 400F into the vessel while at operating conditions (551.4 0F 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 GE Nuclear Energy NEDO-33144 Non-proprietary Version 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., prpm
= 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:
Opm = 24.84 ksi Osm = 16.19 ksi ay, = 45.0 ksi t, = 6.1875 inches COpb = 0.22 ksi C
0 sb = 19.04 ksi a = 2.36 inches rn = 7.09 inches t, = 7.125 inches In this case the total stress, 60.29 ksi, exceeds the yield stress, ay,, 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 inside surface temperature is used.)
R = fays - apm + ((oftotat - ays) / 30)] / (Ototal - Opm)
(4-7)
For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for crpm. The resulting stresses are:
cpm = 24.84 ksi Osm = 9.44 ksi COpb = 0.13 ksi asb =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'n = 3.072. The resulting value obtained was:
Mm = 1.85 for t-<2 Mm = 0.926 4it for 2< fti:<3.464 = 2.845 Mm = 3.21 for ft >3.464 GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, is therefore, l
F(a/rn) =1.4 1
K1p is calculated from Equation 4-4:
Klp = 2.0 (24.84 + 0.13) - (T 2.36) 112.1.4 1
Klp = 190.4 ksi-in"2 K1, is calculated from Equation 4-5:
Kis = 2.845 (9.44 + 2/3 11.10)
Kis = 47.9 ksi-in12
]
The total K, is, therefore, 238.3 ksi-in"'.
The total K, is substituted into Equation 4-6 to solve for (T - RTNDT):
l (T - RTNDT) = In [(238.3-33.2) / 20.734] /0.02 (T - RTNDT) = 115°F l
The ((
)) curve was generated by scaling the stresses used to determine the K,;
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'F water injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in"2, the pressure is 1050 psig and the hot reactor vessel temperature is 551.40F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by l
(Tsaturation - 40) / (551.4 - 40).
From K, the associated (T - RTNDT) can be calculated as shown in Table 4-13.
l GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 4-13: Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp.
Ki*
(T - RTNDT)
.pi)
(k.
- n.
F....
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 Ki.
The highest non-beltline RTNDT for the feedwater nozzle at Columbia is 0F as shown in Table 4-2. However, the RTNDT was increased to 340F to consider the stresses in the core spray nozzle as previously discussed. The generic curve is applied to the Columbia upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 340 F as discussed in Section 4.3.2.1.3.
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
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 [6]. As the beltline fluence increases with the increase in
]
operating life, the P-T curves shift to a higher temperature.
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 1000F/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. Thermal stresses are calculated including clad thickness I
as defined by the ASME Code. As demonstrated in Tables 4-Sa, 4-5b, 4-6a, and 4-6b, the ART is conservatively calculated using minimum wall thickness excluding clad J
thickness.
An evaluation was performed for the vessel wall thickness transition discontinuity located between the lower and lower-intermediate shells in the beltline region. Appendix G of this J
report contains a detailed description of this evaluation.
It was concluded that the discontinuity is bounded by the beltline P-T curve developed in the following sections, and
]
no further adjustment was required.
4.3.2.2.1 Beltline Region - Pressure Test 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 l
thickness (tmin) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:
J Gm = PR I tMin (4-8) l GE Nuclear Energy NEDO-33144 GE Nuclear Energy ND-34 Non-proprietary Versiori The stress intensity factor, K1m, 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 Kic equation of Paragraph A-4200 in ASME Appendix A [17]
for the pressure test condition:
Kim
- SF = Kic = 20.734 exp[O.02 (T - RTNDT)] + 33.2 (4-9)
This relationship provides values of pressure versus temperature (from K1c and (T-RTNDT),
respectively).
GE's current practice for the pressure test curve is to add a stress intensity factor, Kit, for a coolant heatup/cooldown rate, specified as 20'F/hr for Columbia, 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'F/hr. The K1t calculation for a coolant heatup/cooldown rate of 100'F/hr is described in Section 4.3.2.2.3 below.
4.3.2.2.2 Calculations for the Beltline Region - Pressure Test This sample calculation is for a pressure test pressure of 1020 psig at 33.1 EFPY. The following inputs were used in the beltline limit calculation:
A = 28 +35 =630 F Adjusted RTNDT = Initial RTNDT + Shift A = 28 +
in
=
4-6)
(Based on ART values in Table 4-6)
Vessel Height H = 870.5 inches Bottom of Active Fuel Height B = 216.313 inches Vessel Radius (to base metal)
R = 126.6875 inches Minimum Vessel Thickness (without clad) t = 6.1875 inches GE Nuclear Energy NEDO-33144 Non-proprietary Version l
Pressure is calculated to include hydrostatic pressure for a full vessel:
I P = 1020 psi + (H - B) 0.0361 psi/inch = P psig
= 1020 + (870.5 - 216.313) 0.0361 = 1044 psig (4-10)
I Pressure stress:
a = PR/t
= 1.044 126.6875 / 6.1875 = 21.37 ksi (4-11) l The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]
was based on a thickness of 6.1875 inches (the minimum thickness without cladding);
hence, t1"2 = 2.49. The resulting value obtained was:
I I
Mm = 1.85 for 4t<2 Mm = 0.926 It for 2< I<3.464 = 2.30 Mm = 3.21 for fi >3.464 The stress intensity factor for the pressure stress is Kim = Mm -C. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 200F/hr instead of 1 000F/hr.
l I
I Equation 4-9 can be rearranged, and 1.5 Kim substituted for Kic, to solve for (T-RTNDT).
Using the Kic equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 49.15, and Kit= 2.28 for a 200F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:
(T - RTNDT)
= ln[(1.5-Kim + K,, - 33.2) / 20.734] / 0.02
= ln[(1.5 49.15 + 2.28 - 33.2) / 20.734] / 0.02
= 36.20 F (4-12)
I I
T can be calculated by adding the adjusted RTNDT:
I I
GE Nuclear Energy NEDO-33144 Non-proprietary Version T = 36.2 + 63 = 99.2 0F for P = 1020 psig at 33.1 EFPY 4.3.2.2.3 Beltline Region - Core Not Critical Heatup/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 Xl Appendix G [6]:
Kic = 2.0
- Kim +Kit (4-13) where Kim is primary membrane K due to pressure and K1t 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 AT, is based on one-dimensional heat conduction through an insulated flat plate:
a 2T(x,t) /a X2 = 1 / I (8T(x,t) I at)
(4-14) where T(x,t) is temperature of the plate at depth x and time t, and P is the thermal diffusivity.
The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that aT(x,t) / at = dT(t) / dt = G, where G is the coolant GE Nuclear Energy NEDO-33144 Non-proprietary Version heatup/cooldown rate, normally 1 00F/hr. The differential equation is integrated over x for the following boundary conditions:
l
- 1. Vessel inside surface (x = 0) temperature is the same as coolant temperature, To.
l
- 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:
I T = Gx2 I 23 - GCx I j + To (4-15) l This equation is normalized to plot (T - To) / ATw versus x / C.
l 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 1
appropriate Mt of Figure G-2214-1 of ASME Appendix G [6] to compute K1t for heatup and cooldown.
4 The Mt relationships were derived in the Welding Research Council (WRC) 4 Bulletin 175 [15] for infinitely long cracks of 1/4T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.
4.3.2.2.4 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown This Columbia sample calculation is for a pressure of 1020 psig for 33.1 EFPY. The core not critical heatup/cooldown curve at 1020 psig uses the same Klm calculation 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 condition rather than test condition; the operational condition necessitates the use of a higher safety factor. In addition, there is a K1t term for the thermal stress. The additional inputs used to calculate K1t are:
l GE Nuclear Energy NEDO-33144 Non-proprietary Version Coolant heatup/cooldown rate, normally 1000F/hr G = 100 'F/hr Minimum vessel thickness, including clad thickness C = 0.526 ft (6.1875" + 0.125" = 6.3125")
Thermal diffusivity at 5500F (most conservative value) p = 0.354 ft2/ 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 = GC2/2f3
= 100 - (0.526)2/ (2 - 0.354) = 39 0F (4-16)
The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2914) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kt = Mt A
AT = 11.36, can be calculated.
The conservative value for thermal diffusivity at 5500F is used for all calculations; therefore, K1t is constant for all pressures.
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)
=
In[((2* Klm + Klt)-33.2) /20.734] /0.02
= ln[(2-49.15+ 11.36-33.2)/20.734]/0.02
=
65.2 0F (4-17)
T can be calculated by adding the adjusted RTNDT:
T = 65.2 + 63= 128.2 0F for P = 1020 psig at 33.1 EFPY It is noted that the core not critical beltline curve is bounded by 10CFR50 Appendix G requirements at 1020 psig as can be seen in Figure 5-10.
GE Nuclear Energy NEDO-33144 Non-proprietary Version 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and 1
temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. Similar to the evaluations performed for the bottom head and upper vessel, a BWR/6 finite element analysis [18] was used to model the flange region. The local stresses were computed for determination of the stress intensity factor, K1. Using a 1/4T flaw size and the Kjc formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the 10CFR50 Appendix G requirement of RTNDT + 900F (the largest T-RTNDT for the flange at 1563 psig is 730F). For pressures below 312 psig, the flange curve is bounded by RTNDT + 60 (the largest T - RTNDT for the flange at 312 psig is 540F); therefore, instead of determining a T (temperature) versus l
pressure curve for the flange (i.e., T - RTNDT) the value RTNDT + 60 is used for the closure flange limits.
l In some cases, the results of analysis for other regions exceed these requirements and l
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 Columbia at low pressures.
4 The approach used for Columbia for the bolt-up temperature was based on the J
conservative value of (RTNDT+ 60), or the LST of the bolting materials, whichever is greater.
The 60'F adder is included by GE for two reasons:
- 1) the pre-1971 requirements of the ASME Code Section 1I1, Subsection NA, Appendix G included the 60'F adder, and 2) inclusion of the additional 60'F requirement above the RTNDT provides the additional assurance that a 1/4T flaw size is acceptable. As shown in Tables 4-1, 4-2, and 4-3, the limiting initial RTNDT for the closure flange region is represented by Shell #4 at 200F, and the LST of the closure studs is 100F; therefore, the bolt-up temperature value used is the more conservative value of 80'F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) l for brittle fracture.
I GE Nuclear Energy NEDO-33144 GE Nuclear Energy ND-34 Non-proprietary Version 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 + 90'F) and Curve B temperature no less than (RTNDT + 1200F).
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 680F for the reason discussed below.
The shutdown margin, provided in the Columbia Technical Specification [22], is calculated for a water temperature of 680F. 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 680F limit, further extensive calculations would be required to justify a lower temperature. The 800F limit for the upper vessel and beltline region and the 680F 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 1 of [8] requires that core critical P-T limits be 40'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 400F for pressures above 312 psig.
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
Table 1 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'F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 80'F, based on an RTNDT of 200F. In addition, l
above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160'F or the temperature required for the hydrostatic pressure l
test (Curve A at 1020 psig). The requirement of closure region RTNDT + 160'F causes a temperature shift in Curve C at 312 psig.
l GE Nuclear Energy NEDO-33144 Non-proprietary Version
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)
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 100'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 200F/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 3/4T for a given metal temperature.
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
The following P-T curves were generated for Columbia:
Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 22 and 33.1 effective full power years (EFPY). The composite l
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 22 and l
33.1 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 22 and 33.1 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.
l Using the fluence from Section 4.2.1.2, the P-T curves are upper vessel limited above 910 psig for Curve A and above 790 psig for Curve B for both 22 and 33.1 EFPY.
l Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves is l
presented in Appendix B.
I GE Nuclear Energy NEDO-33144 Non-proprietary Version Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curverve escription Numers for for PresentationPeetaino of th'e P-,T:.: the' P Curves:
u rv e s A
Bottom Head Limits (CRD Nozzle)
Figure 5-1 Tables B-1 & B-3 A
Upper Vessel Limits (FW Nozzle)
Figure 5-2 Tables B-1 & B-3 A
Beltline Limits - 22 EFPY Figure 5-3 Table B-1 A
Beltline Limits - 33.1 EFPY Figure 5-4 Table B-3 A
Bottom Head and Composite Curve A -22 EFPY*
Figure 5-5 Table B-2 A
Bottom Head and Composite Curve A - 33.1 EFPY*
Figure 5-6 Table B-4 B
Bottom Head Limits (CRD Nozzle)
Figure 5-7 Tables B-1 & B-3 B
Upper Vessel Limits (FW Nozzle)
Figure 5-8 Tables B-1 & B-3 B
Beltline Limits - 22 EFPY Figure 5-9 Table B-1 B
Beltline Limits - 33.1 EFPY Figure 5-10 Table B-3 B
Bottom Head and Composite Curve B - 22 EFPY*
Figure 5-11 Table B-2 B
Bottom Head and Composite Curve B - 33.1 EFPY*
Figure 5-12 Table B4 C
Composite Curve C - 22 EFPY**
Figure 5-13 Table B-2 C
Composite Curve C - 33.1 EFPY**
Figure 5-14 Table B-4 The Composite Curve A & B curve is the more limiting of three limits: 1 OCFR50 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.
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version 1400 1300 1200 1100
'a._
um 1000 X
900 LL0 I-w m"
800 C')
w o
700 II-w 600 z
1 500 w
w In 400 300 200 100 0
INITIAL RTndt VALUE IS l340F FOR BOTTOM HEAD I HEATUP/COOLDOWN RATE OF COOLANT
< 20'F/HR I
I ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE I
I 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)
Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve A]
[20 0F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-331 44 Non-proprietary Version 1400 1300 1200 1100 D 1000 I
IL 900 0I---z III WO 800 lo o
700 0
w 0
600 Z
i 500 LU w
gn 400 fLu w
300 1+ I 31 9-10PSIGG A
I l
lI FLANGE REGION 80T3
_<-1 I
INITIAL RTndt VALUE IS 34°F FOR UPPER VESSEL HEATUP/COOLDOWN RATE OF COOLANT
< 20*F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE
-UPPER VESSEL LIMITS (Including Flange and FW Nozzle Limits) 200 100 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]
[201F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version l
I 1400 1300 1200 1100 a.
1000 LUI I
EL 900 0
-J LU Cn 800 LU
=
600 z
X 500 w
U) 400 0~
300 200 100 0
INITIAL RTndt VALUE IS 28 0F FOR BELTLINE I
BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (-F) 22 28 HEATUP/COOLDOWN RATE OF COOLANT s 20'F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE BELTLINE LIMITS l
A I
I I
l I
l l
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
( 0F)
I Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 22 EFPY
[20 0F/hr or less coolant heatup/cooldown]
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 1300 1200 1100
.m 1 000 LU L.
900 0
III-uJ u) 800 U)
LU O
700 I-0 tL 3 600 Z
i3 500 CU 400 w
0U 300 I 1I60 PS11G1 I
I 132PSIG JI I
INITIAL RTndt VALUE IS 28°F FOR BELTLINE BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 33.1 35 HEATUP/COOLDOWN RATE OF COOLANT
< 200F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE
-BELTLINE LIMITS 200 100 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 33.1 EFPY
[20°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 l
Non-proprietary Version I
l 1400 1300 1200 1100 0.
]1 000 w
I X-900 0
-jw u) 800 uJ w
t 600 j
500 us O 400 w
0:
300 200 100 0n INITIAL RTndt VALUES ARE 28 0F FOR BELTLINE, 340F FOR UPPER VESSEL, AND 34 0F FOR BOTTOM HEAD l
BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT( 0F) 22 28 HEATUP/COOLDOWN RATE OF COOLANT
< 20'F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE I
1 l
I l
I 1
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(*F)
I Figure 5-5: Composite Pressure Test P-T Curves [Curve A] up to 22 EFPY
[20°F/hr or less coolant heatup/cooldown]
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 1300 1200 1100 1000 wz II v) 800 w0 700 I
600 Z
I-_
i 500 1I1 K
U) 400 U)w 300 200 100 BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (0F) 33.1 35 HEATUP/COOLDOWN RATE OF COOLANT
< 20°FJHR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-6: Composite Pressure Test P-T Curves [Curve A] up to 33.1 EFPY
[20°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version I
I 1400 1300 1200 1100 0._
c] 1000 I
ao 900 0
I-w CO 800 f)w o
700 w
a:
600 z
i 500 usw in 400 w
300 200 100 0
INITIAL RTndt VALUE IS 34°F FOR BOTTOM HEAD HEATUP/COOLDOWN RATE OF COOLANT
< 100°F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE I
I I
I I
I I1 I1 I1 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( 0F)
I1 Figure 5-7: Bottom Head P-T Curve for Core Not Critical [Curve B]
[100 0F/hr or less coolant heatup/cooldown]
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 1300 1200 1100 a,
In D
1000 aQ 900 0II
-Jj w
all 800 o
700 0
a:
600 z
3 500 u) 400 U'
300 I
I
[INITIALRT'ndtVALUEIS I
l34°F FOR UPPER VESSELl HEATUP/COOLDOWN RATE OF COOLANT c 100*F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE
-UPPER VESSEL LIMITS (Including Flange and FW Nozzle Limits) 200 100 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-8: Upper Vessel P-T Curve for Core Not Critical [Curve B]
[1 OO 0 F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version I
I 1400 1300 1200 1390 PSIGI I
I__
I INITIAL RTndt VALUE IS 280F FOR BELTLINE BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (0F) 22 28 HEATUP/COOLDOWN RATE OF COOLANT
< 100'F/HR 1100 0
I 0.0 I-
-J (n
U) w 0
Q1 Uj z
I-U, U,
EL 1000 900 800 700 600 500 400
~II 1I 10CFR50 BOLTUP
-180 F
1 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE BELTLINE LIMITS 300 200 100 0
I 0
25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
( 0F)
I Figure 5-9: Beltline P-T Curve for Core Not Critical [Curve B] up to 22 EFPY
[100 0F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-331 44 Non-proprietary Version 1400 1300 1200 1100
.F CL 1000 w
0 900 0F--z
[it U) 800 U)
LU o
700 0
Lu w
600 Z
i 500 al gn 400 Lu 300 INITIAL RTndt VALUE IS 280F FOR BELTLINE BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (F) 33.1 35 HEATUP/COOLDOWN RATE OF COOLANT
< 100'F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE
-BELTLINE LIMITS 200 100 0
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-10: Beltline P-T Curve for Core Not Critical [Curve B] up to 33.1 EFPY
[100 0F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version I
1 1400 1300 1200 1100 1000 w
0-900 0
-jw u) 800 w
o 700 (t
600 Z
I-1 500 w
0 400 w
300 200 100 0
[70PIG]
BOTTOM l 600 PSIGl REGION 0
25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE INITIAL RTndt VALUES ARE 280F FOR BELTLINE, 340F FOR UPPER VESSEL, AND 340F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT(-F) 22 28 HEATUP/COOLDOWN RATE OF COOLANT
< 1000F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE UPPER VESSEL AND BELTLINE LI M ITS
BOTTOM HEAD CURVE I1 I1 I1 I1 I1 1
I I
I (OF)
Figure 5-1 1: Composite Core Not Critical P-T Curves [Curve B] up to 22 EFPY
[1 OO 0 F/hr or less coolant heatup/cooldown]
I GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 INITIAL RTndt VALUES ARE 28°F FOR BELTLINE, 1300
'--34°F FOR UPPER VESSEL, 34°F FOR BOTTOM HEAD 1200 BELTLINE CURVES 0
1ADJUSTED AS SHOWN:
1100 EFPY SHIFT (°F) 10.35 PSI 103.1 PS cn
[
3iF 18 33.1 35
_1000-I1 I I
I HEATUP/COOLDOWN
- 0. 900 i
RATE OF COOLANT O
l<
100°F/HR Ljti 1790 PSIGI 800
[140'F_
0 700 t
1600 PSIG I Ix 600 6 8
-_F ACCEPTABLE AREA OF z
.OPERATION TO THE RIGHT OF THIS CURVE
- 5 500 400 IL BOTTOMl;l HEAD 300 I68 F
200 l
-l lUPPER VESSEL 200 i
'/AND BELTLINE l FLANGE l
LIMITS 100 G BOTTOM HEAD 100 L~JCURVE 0
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(°F)
Figure 5-12: Composite Core Not Critical P-T Curves [Curve B] up to 33.1 EFPY
[1 00°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 I
28°F FOR BELTLINE, I
l 34°F FOR UPPERl
- VESSEL, 1200 AND 34°F FOR BOTTOM HEAD 1100 1000 BELTLINE CURVE 1000 ADJUSTED AS SHOWN:
w EFPY SHIFT(°F)
I22 28 0900 w
n800 rIG HEATUPICOOLDOWN i
RATE OF COOLANT_
< 100°F/HR 07 00 T
F_
z..
6 00 500 ACCEPTABLE AREA OF LII OPERATION TO THE 400 RIGHT OF THIS CURVE w
1t 1 12 PIGl_
300 4
200 Minimum Criticality Temperature 80F BELTLINE AND 100
\\
2
.NON-BELTLINE 100
-ILIMITS
\\v[l60 PSIGl 01 0
25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(°F)
Figure 5-13: Composite Core Critical P-T Curves [Curve C] up to 22 EFPY
[1 00°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 28°F FOR BELTLINE, 34°F FOR UPPER
- VESSEL, 1200
_I AND 34°F FOR BOTTOM HEAD 1100-0I 1035 PSIG rI 11 F BELTLINE CURVE 1000 ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 9033.1 35 0
______7.0__PSG Cn 800 79 l
800 9HEATUP/COOLDOWN
>LJRATE OF COOLANT
< 1000F/HR 0
700 I-6 0
______I___
'F600 Z
ACCEPTABLE AREA OF w
OPERATION TO THE D_
RIGHT OF THIS CURVE
'n 400 L
I I
I l~312 PSIGl 300 h
]
200 Minimum Criticality Temperature 80'F l
BELTLINE AND 100 NON-BELTLINE 0
25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 5-14: Composite Core Critical P-T Curves [Curve C] up to 33.1 EFPY
[1 00°F/hr or less coolant heatup/cooldown]
GE Nuclear Energy NEDO-33144 Non-proprietary Version
6.0 REFERENCES
- 1. a) Letter #EN2-PE-04-002, JJ Sisk (Energy Northwest) to JE Larsen (GE), "Contract No. C-31814, Data Package Transmittal No. 002 For: RPV Pressure Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, January 13, 2004.
b) Letter #EN2-PE-04-009, JJ Sisk (Energy Northwest) to JE Larsen (GE), "Contract No. C-31814, Data Package Transmittal No. 003 For: RPV Pressure Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, February 17, 2004.
c) Letter #EN2-PE-04-016, JJ Sisk (Energy Northwest) to JE Larsen (GE), 'Contract No. C-31814, Data Package Transmittal No. 005 For: RPV Pressure Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, March 23, 2004.
- 2.
GE Drawing Number 762E120, "Reactor Vessel Thermal Cycles - Reactor Vessel",
GE-NED, San Jose, CA, Revision 2 (GE Proprietary).
- 3.
GE Drawing Number 166B7120, "Reactor Vessel Nozzle Thermal Cycles - Reactor System", GE-NED, San Jose, CA, Revision 4 (GE Proprietary).
- 4.
Wu, T., "Energy Northwest Columbia Generating Station Neutron Flux Evaluation",
GE-NE-0000-0023-5057-RO, Revision 0, GE Nuclear Energy, San Jose, CA, April 2004 (GE Proprietary).
- 5.
"BWR Vessel and Internals Project BWR Integrated Surveillance Program Implementation Guidelines", BWRVIP-102, EPRI, Palo Alto, CA, June 2002 (EPRI Proprietary).
- 6. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section III or Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.
GE Nuclear Energy NEDO-33144 Non-proprietary Version l
- 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 l
Research Council Bulletin 217, July 1976.
1
- 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, I
- Method, September 1994",
- USNRC, December 16, 1994.
1
"General Electric Company BWR Projects Department QA-Engineered Equipment and Installation", May 1976.
- 13. C. Chu, "Washington Public Power Supply System WNP-2 RPV Surveillance Materials Testing and Analysis", GE-NE, San Jose, CA, March 1997 (GE-NE-B1301809, Revision 1).
- 14. Letter, S.A. Richard, USNRC to J.F. Klapproth, GE-NE, "Safety Evaluation for NEDC-1 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. ((
I
))
I
- 17. "Analysis of Flaws", Appendix A to Section Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.
GE Nuclear Energy NEDO-33144 Non-proprietary Version
- 18. ((
- 19. Bottom Head and Feedwater Nozzle Dimensions:
- a. Drawing # 11, Revision 5, "Bottom Head Dollar Plates", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-021).
- b. Drawing # 12, Revision 7, "Bottom Head Radial Plates", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-022).
- c. Drawing # 56, Revision 2, "N4 Nozzle Forging (Feedwater)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-065).
- d. Drawing # 59, Revision 11, "N4 Nozzle Assembly (Feedwater)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-068).
- 20. ((
))
- 21. "Materials - Properties", Part D to Section II of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.
GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version l
((
I I1 I1 I1 A
A A
A I
J A-2
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1]
A-3
GE Nuclear Energy NEDO-33144
-l Non-proprietary Version 1
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'F. Inconel (or Alloy 600) and stainless steel discontinuities require no fracture toughness evaluations.
I I
I Nozzle or Appurtenance Material Reference Remarks Nozzles made from Alloy 600 and IVK 26-3 N17 - Seal Leak DetectorNozemaefmAly60an Nozzle (attached to Shell Flange) less than 2.5" require no fracture Alloy 600 1, 10 toughness evaluation.
M K 79 N 11 Core Differential Pressure and Liquid Control Nozzle Nozzles made from Alloy 600 and (See Table A-1 for Penetration in less than 2.5" require no fracture the Bottom Head Dollar Plate)
Alloy 600 1, 30, 44 toughness evaluation.
Nozzles less than 2.5" in thickness MK 82 N12 & N13 Instrumentation require no fracture toughness Nozzles SA 508 CL 1 1, 7, 8, 31 evaluation.
Nozzles less than 2.5" in thickness require no fracture toughness MK 85 N14 Instrumentation Nozzle SA 508 CL 1 1, 9, 32, 33 evaluation.
Nozzles less than 2.5" in thickness require no fracture toughness MK 87 N15 Drain Nozzle SA 508 CL 1 1, 34, 44 evaluation.
Components made from Alloy 600 MK 17, MK 20-1, 20-2 Shroud 1, 4, 6, 43, 46, require no fracture toughness Support Alloy 600 47 evaluation.
MK 43 Refueling Containment Skirt Attachment (to Shell Flange) and Not a pressure boundary Refueling Bellows Base component; therefore requires no MK 26 Refueling Bellows Bar SA 516 GR 70 1, 49 fracture toughness evaluation.
Not a pressure boundary 1, 8, 9, 10, 20, component; therefore requires no MK 44 Thermocouple Pad SA 516 GR 70 44, 45, 50, 51 fracture toughness evaluation.
Nozzles made from Alloy 600 and MK 14 CRD Stub Tubes (in Bottom less than 2.5" require no fracture Head Dollar Plate)
Alloy 600 1, 44, 52, 53 toughness evaluation.
Appurtenances made from SA1 82 TP Stainless Steel require no fracture MK 106 Surveillance Brackets F304L 1, 7, 54, 55 toughness evaluation.
I I1 I
I I
I 1
I A-4
GE Nuclear Energy NEDO-33144 1 Non-proprietary Version Nozzle or Appurtenance Material Reference Remarks Appurtenances made from MK 110 Jet Pump Riser Support Stainless Stainless Steel require no fracture Pads Steel 1, 7, 40 toughness evaluation.
Appurtenances made from SA1 82 TP Stainless Steel require no fracture MK 108 Core Spray Brackets F304L 1, 8,41 toughness evaluation.
Appurtenances made from MK 104 Feedwater Sparger SA1 82 TP Stainless Steel require no fracture Brackets F304L 1, 9, 42 toughness evaluation.
MK 100 Steam Dryer Support Nozzles made from Inconel require Bracket Inconel 1, 9, 56, 57 no fracture toughness evaluation.
Appurtenances made from SAI 82 TP Stainless Steel require no fracture MK 98 Guide Rod Bracket F304 1, 39, 58 toughness evaluation Loading only occurs during outages. Not a pressure boundary SA 533 GR B component; therefore requires no MK 40 Top Head Lifting Lugs CL1 1, 11, 59 fracture toughness evaluation.
Not a pressure boundary MK 102 Steam Dryer Hold Down SA 533 GR B component; therefore requires no Bracket CL 1 1, 11, 48, 60 fracture toughness evaluation.
A-5
GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX A
REFERENCES:
-1
- 1. Vessel Drawings Drawing # 1, Revision 8, "Vessel Outline", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-001).
1 Drawing # R17, Revision 2, "Vessel and Attachments Material Identification",
CBI Nuclear Company, Chicago, Illinois (VPF # 3133-531).
1 Drawing # 123, Revision 7, "Vessel Nozzle and Outside Bracket As Built Dimensions", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-484).
1 Drawing # R8, Revision 3, "Nozzles - Heat Number, Lot Number and Material Summary", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-120).
1 QA Records and RPV CMTRs for Columbia (formerly Hanford-2), GE PO #205-AE023, Contract #72-2647, Manufactured by CBI Nuclear Company, Chicago, Illinois.
- 2. Wu, T., "Energy Northwest Columbia Generating Station Neutron Flux Evaluation",
GE-NE-0000-0023-5057-RO, Revision 0, GE Nuclear Energy, San Jose, CA, April 2004 (GE Proprietary).
- 3. a) Letter #EN2-PE-04-002, JJ Sisk (Energy Northwest) to JE Larsen (GE), "Contract 1
No. C-31814, Data Package Transmittal No. 002 For: RPV Pressure Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia I
Generating Station, January 13, 2004.
b) Letter #EN2-PE-04-009, JJ Sisk (Energy Northwest) to JE Larsen (GE), "Contract No. C-31814, Data Package Transmittal No. 003 For: RPV Pressure Temperature l
Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, February 17, 2004.
c) Letter #EN2-PE-04-016, JJ Sisk (Energy Northwest) to JE Larsen (GE), "Contract No. C-31814, Data Package Transmittal No. 005 For: RPV Pressure Temperature l
A-6
GE Nuclear Energy NEDO-33144 Non-proprietary Version Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, March 23, 2004.
- 4. Drawing # 11, Revision 5, "Bottom Head Dollar Plates", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-021).
- 5. Drawing # 12, Revision 7, "Bottom Head Radial Plates", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-022).
- 6. Drawing # 21, Revision 5, "#1 Shell Ring", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-030).
- 7. Drawing # 22, Revision 3, "#2 Shell Ring Assembly", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-031).
- 8. Drawing # 23, Revision 6, "W3 Shell Ring Assembly", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-032).
- 9. Drawing # 24, Revision 2, "#4 Shell Ring Assembly", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-033).
- 10. Drawing # 26, Revision 9, "Shell Flange Forging and Assembly with N17 Nozzle",
CBI Nuclear Company, Chicago, Illinois (VPF # 3133-035).
- 11. Drawing # 32, Revision 5, "Top Head Assembly", CBI Nuclear Company, Chicago, Illinois (VPF #3133-041).
- 12. Drawing # 33, Revision 3, "Top Head Details", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-042).
- 13. Drawing # 46, Revision 4, "N1 Nozzle Forging (Recirculation Outlet)", CBI Nuclear Company, Chicago, Illinois (VPF#3133-055).
A-7
GE Nuclear Energy NEDO-33144 Non-proprietary Version
- 14. Drawing # 48, Revision 4, "N1 Nozzle Assembly (Recirculation Outlet)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-057).
- 15. Drawing # 49, Revision 5, "N2 Nozzle Forging (Recirculation Inlet)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-058).
- 16. Drawing # 52, Revision 11, "N2 Nozzle Assembly (Recirculation Inlet)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-061).
- 17. Drawing # 53, Revision 3, "N3 Nozzle Forging (Steam Outlet)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-062).
- 18. Drawing # 55, Revision 3, "N3 Nozzle Assembly (Steam Outlet)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-064).
- 19. Drawing # 56, Revision 2, "N4 Nozzle Forging (Feedwater)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-065).
- 20. Drawing # 59, Revision 11, "N4 Nozzle Assembly (Feedwater)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-068).
- 21. Drawing # 60, Revision 3, "N5 Nozzle Forging (Core Spray)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-069).
- 22. Drawing # 63, Revision 5, "N5 Nozzle Assembly (Core Spray - Low Pressure)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-072).
- 23. Drawing # 67, Revision 6, "N6 Nozzle Assembly (RHR/LPCI Mode)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-076).
- 24. Drawing # 68, Revision 2, `N7 Nozzle and Weld Neck Flange (Head Spray)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-077).
A-8
GE Nuclear Energy NEDO-33144 Non-proprietary Version
- 25. Drawing # 70, Revision 4, "N8 Nozzle Forging (Vent)", CBI Nuclear Company, Chicago, Illinois (VPF#3133-079).
- 26. Drawing # 71, Revision 9, "N8 Nozzle Assembly (Vent)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-080).
- 27. Drawing # 72, Revision 2, "N9 Nozzle Forging Jet Pump Instrumentation", CBI Nuclear Company, Chicago, Illinois (VPF #3133-081).
- 28. Drawing # 74, Revision 2, "N9 Nozzle Assembly (Jet Pump Instrumentation)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-083).
- 29. Drawing # 75, Revision 4, "N10 Nozzle Forging (CRD Hyd. System Return)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-084).
- 30. Drawing # 81, Revision 6, "Core Differential Pressure and Liquid Control Nozzle Assembly (N11)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-090).
- 31. Drawing # 82, Revision 3, "N12 and N13 Nozzle Forgings (Instrumentation)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-091).
- 32. Drawing # 85, Revision 3, "N14 Nozzle Forgings (Instrumentation)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-094).
- 33. Drawing # 86, Revision 3, "N14 Instrumentation Nozzle Assembly", CBI Nuclear Company, Chicago, Illinois (VPF #3133-095).
- 34. Drawing # 87, Revision 6, "N15 Drain Nozzle Forging and Attachment", CBI Nuclear Company, Chicago, Illinois (VPF#3133-096).
- 35. Drawing # 91, Revision 6, "N16 Nozzle Assembly (Core Spray - High Pressure)",
CBI Nuclear Company, Chicago, Illinois (VPF #3133-100).
A-9
GE Nuclear Energy NEDO-331 44 Non-proprietary Version
- 36. Drawing # 92, Revision 4, "N18 Nozzle and Weld Neck Flange (Spare)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-101).
- 37. Drawing # 93, Revision 6, " N18 Nozzle Assembly (Spare)", CBI Nuclear Company, Chicago, Illinois (VPF #3133-102).
- 38. Drawing # 95, Revision 2, "Stabilizer Bracket Attachment", CBI Nuclear Company, Chicago, Illinois (VPF #3133-104).
- 39. Drawing # 99, Revision 4, "Guide Rod Bracket Attachment", CBI Nuclear Company, Chicago, Illinois (VPF #3133-108).
- 40. Drawing # 110, Revision 6, "Jet Pump Riser Support Pads", CBI Nuclear Company, Chicago, Illinois (VPF#3133-304).
- 41. Drawing # 108, Revision 2, "Core Spray Bracket Forgings", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-352).
- 42. Drawing # 104, Revision 1, "Feedwater Sparger Bracket Forging", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-368).
- 43. Drawing # 20, Revision 2, "Shroud Support Assembly", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-444).
- 44. Drawing # 13, Revision 3, "Bottom Head Assembly", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-023).
- 45. Drawing # 9, Revision 5, "Skirt Knuckle Details", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-019).
- 46. Drawing # 17, Revision 5, "Shroud Support Stubs", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-027).
A-10
GE Nuclear Energy NEDO-33144 Non-proprietary Version
- 47. Drawing # 19, Revision 7, "Shroud Support Assembly Details", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-029).
- 48. Drawing # 103, Revision 3, "Steam Dryer Hold Down Bracket Attachment" CBI Nuclear Company, Chicago, Illinois (VPF # 3133-112).
- 49. Drawing # 42, Revision 2, "Refueling Bellows Support", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-051).
- 50. Drawing # 44, Revision 3, "Thermocouple Pads for Vessel", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-053).
- 51. Drawing # 45, Revision 4, "Thermocouple Pads for Nozzle N4 at 90° and 21 0", CBI Nuclear Company, Chicago, Illinois (VPF # 3033-054).
- 52. Drawing # 15, Revision 2, "Control Rod Drive Stub Assembly", CBI Nuclear Company, Chicago, Illinois (VPF#3133-025).
- 53. Drawing # 16, Revision 2, "Incore Penetration Details", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-026).
- 54. Drawing # 106, Revision 2, 'Surveillance Specimen Bracket", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-284).
- 55. Drawing # 107, Revision 3, "Surveillance Specimen Bracket Attachment", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-285).
- 56. Drawing # 100, Revision 2, "Steam Dryer Support Bracket Forging", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-109).
- 57. Drawing # 101, Revision 2, uSteam Dryer Support Bracket Attachment", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-110).
A-11
GE Nuclear Energy NEDO-331 44 Non-proprietary Version 1
- 58. Drawing # 98, Revision 2, "Guide Rod Bracket Forging", CBI Nuclear Company, l
Chicago, Illinois (VPF # 3133-107).
l
- 59. Drawing # 40, Revision 1, "Top Head Lifting Lugs", CBI Nuclear Company, Chicago, Illinois (VPF # 3133-049).
1
- 60. Drawing # 102, Revision 2, "Steam Dryer Hold Down Bracket", CBI Nuclear I
Company, Chicago, Illinois (VPF # 3133-111).
I A-12 I
GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 1
BOTTOM UPPER 22 EFPY BOTTOM UPPER 22 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVEA CURVEA CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(0F) 0 68.0 80.0 80.0 68.0 80.0 80.0 10 68.0 80.0 80.0 68.0 80.0 80.0 20 68.0 80.0 80.0 68.0 80.0 80.0 30 68.0 80.0 80.0 68.0 80.0 80.0 40 68.0 80.0 80.0 68.0 80.0 80.0 50 68.0 80.0 80.0 68.0 80.0 80.0 60 68.0 80.0 80.0 68.0 80.0 80.0 7 0 68.0 80.0 80.0 68.0 80.0 80.0 80 68.0 80.0 80.0 68.0 80.0 80.0 90 68.0 80.0 80.0 68.0 80.0 80.0 100 68.0 80.0 80.0 68.0 80.0 80.0 110 68.0 80.0 80.0 68.0 80.0 80.0 120 68.0 80.0 80.0 68.0 80.0 80.0 130 68.0 80.0 80.0 68.0 80.0 80.0 140 68.0 80.0 80.0 68.0 80.0 80.0 150 68.0 80.0 80.0 68.0 80.0 80.0 160 68.0 80.0 80.0 68.0 80.0 80.0 170 68.0 80.0 80.0 68.0 80.0 80.0 180 68.0 80.0 80.0 68.0 81.9 80.0 190 68.0 80.0 80.0 68.0 84.2 80.0 200 68.0 80.0 80.0 68.0 86.3 80.0 210 68.0 80.0 80.0 68.0 88.3 80.0 220 68.0 80.0 80.0 68.0 90.3 80.0 230 68.0 80.0 80.0 68.0 92.1 80.0 240 68.0 80.0 80.0 68.0 93.9 80.0 B-2
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9
'BOTTOM UPPER 22 EFPY BOTTOM UPPER 22EFPY PRESSURE (PSIG)
HEAD CURVE A (OF)
VESSEL CURVE A (0F)
BELTLINE CURVE A (0F)
HEAD CURVE B (OF)
VESSEL CURVE B (0F)
BELTLINE CURVE B (0F) 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 490 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 95.6 97.2 98.8 100.3 101.8 103.2 104.5 104.9 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 B-3
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 22 EFPY BOTTOM UPPER 22 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVE A CURVEA CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(°F) 500 68.0 110.0 110.0 68.0 140.0 140.0 510 68.0 110.0 110.0 68.0 140.0 140.0 520 68.0 110.0 110.0 68.0 140.0 140.0 530 68.0 110.0 110.0 68.0 140.0 140.0 540 68.0 110.0 110.0 68.0 140.0 140.0 550 68.0 110.0 110.0 68.0 140.0 140.0 560 68.0 110.0 110.0 68.0 140.0 140.0 570 68.0 110.0 110.0 68.0 140.0 140.0 580 68.0 110.0 110.0 68.0 140.0 140.0 590 68.0 110.0 110.0 68.0 140.0 140.0 600 68.0 110.0 110.0 68.0 140.0 140.0 610 68.0 110.0 110.0 68.7 140.0 140.0 620 68.0 110.0 110.0 70.1 140.0 140.0 630 68.0 110.0 110.0 71.5 140.0 140.0 640 68.0 110.0 110.0 72.9 140.0 140.0 640 68.0 110.0 110.0 72.9 140.0 140.0 650 68.0 110.0 110.0 74.2 140.0 140.0 660 68.0 110.0 110.0 75.5 140.0 140.0 670 68.0 110.0 110.0 76.8 140.0 140.0 680 68.0 110.0 110.0 78.1 140.0 140.0 700 68.0 110.0 110.0 80.4 140.0 140.0 760 68.0 110.0 110.0 87.3 140.0 140.0 700 68.0 110.0 110.0 80.4 140.0 140.0 710 68.0 110.0 110.0 81.6 140.0 140.0 720 68.0 110.0 110.0 82.7 140.0 140.0 730 68.0 110.0 110.0 83.8 140.0 140.0 740 68.0 110.0 110.0 84.0 140.0 140.0 760 68.0 110.0 110.0 87.0 140.0 140.0 B-4j
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 OF/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 22 EFPY BOTTOM UPPER 22 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVEA CURVEA CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(0F)
(OF)
(0F)
(0F) 770 68.0 110.0 110.0 88.0 140.0 140.0 780 68.0 110.0 110.0 89.0 140.0 140.0 790 68.0 110.0 110.0 90.0 140.0 140.0 800 68.0 110.0 110.0 90.9 140.1 140.0 810 68.3 110.0 110.0 91.9 140.5 140.0 820 69.4 110.0 110.0 92.8 140.9 140.0 830 70.5 110.0 110.0 93.7 141.2 140.0 840 71.5 110.0 110.0 94.6 141.6 140.0 850 72.6 110.0 110.0 95.4 141.9 140.0 860 73.6 110.0 110.0 96.3 142.3 140.0 870 74.6 110.0 110.0 97.1 142.6 140.0 880 75.5 110.0 110.0 98.0 143.0 140.0 890 76.5 110.0 110.0 98.8 143.3 140.0 900 77.4 110.0 110.0 99.6 143.7 140.0 910 78.4 110.0 110.0 100.4 144.0 140.0 920 79.3 110.2 110.0 101.1 144.4 140.0 930 80.1 110.9 110.0 101.9 144.7 140.0 940 81.0 111.5 110.0 102.7 145.0 140.0 950 81.9 112.1 110.0 103.4 145.4 140.0 960 82.7 112.7 110.0 104.1 145.7 140.0 970 83.6 113.3 110.0 104.9 146.0 140.0 980 84.4 113.9 110.0 105.6 146.4 140.0 990 85.2 114.5 110.0 106.3 146.7 140.0 1000 86.0 115.1 110.0 107.0 147.0 140.0 1010 86.7 115.7 110.0 107.6 147.3 140.0 1020 87.5 116.2 110.0 108.3 147.6 140.0 1030 88.3 116.8 110.0 109.0 148.0 140.0 B-5
GE Nuclear Energy NEDO-33144 Non-proprietary Version l
TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 j
BOTTOM UPPER 22 EFPY BOTTOM UPPER 22 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) j 1040 89.0 117.4 110.0 109.6 148.3 140.0 1050 89.7 117.9 110.0 110.3 148.6 140.0 l
1060 90.4 118.5 110.0 110.9 148.9 140.0 1064 90.7 118.7 110.0 111.2 149.0 140.0 1070 91.2 119.0 110.0 111.5 149.2 140.0 j
1080 91.9 119.5 110.0 112.2 149.5 140.0 1090 92.6 120.1 110.0 112.8 149.8 140.0 l
1100 93.2 120.6 110.0 113.4 150.1 140.0 1105 93.6 120.8 110.0 113.7 150.3 140.0 1110 93.9 121.1 110.0 114.0 150.4 140.0 l
1120 94.6 121.6 110.0 114.6 150.7 140.0 1130 95.2 122.1 110.0 115.2 151.0 140.0 l
1140 95.9 122.6 110.0 115.7 151.3 140.0 1150 96.5 123.1 110.0 116.3 151.6 140.0 1155 96.8 123.4 110.0 116.6 151.7 140.0 1160 97.1 123.6 110.0 116.9 151.9 140.0 1170 97.8 124.1 110.0 117.4 152.2 140.0 l
1180 98.4 124.6 110.0 118.0 152.5 140.0 1190 99.0 125.1 110.0 118.5 152.7 140.0 1200 99.6 125.5 110.0 119.1 153.0 140.0 1210 100.2 126.0 110.0 119.6 153.3 140.0 1220 100.8 126.5 110.0 120.2 153.6 140.0 l
1230 101.3 126.9 110.0 120.7 153.9 140.0 1240 101.9 127.4 110.0 121.2 154.2 140.0 1250 102.5 127.8 110.0 121.7 154.4 140.0 1260 103.0 128.3 110.0 122.2 154.7 140.0 1270 103.6 128.7 110.0 122.7 155.0 140.0 B-6 l
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-1. Columbia P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 22 EFPY BOTTOM UPPER 22 EFPY' PRESSURE (PSIG)
HEAD CURVE A (0F)
VESSEL CURVE A (OF)
BELTLINE CURVE A (0F)
HEAD CURVE B (OF)
VESSEL CURVE B (0F)
BELTLINE CURVE B (OF) 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 104.1 104.7 105.2 105.7 106.3 106.8 107.3 107.8 108.3 108.8 109.3 109.8 110.3 129.2 129.6 130.0 130.5 130.9 131.3 131.7 132.1 132.6 133.0 133.4 133.8 134.2 110.2 110.8 111.3 111.9 112.5 113.0 113.5 114.1 114.6 115.1 115.7 116.2 116.7 123.2 123.7 124.2 124.7 125.2 125.6 126.1 126.6 127.0 127.5 127.9 128.4 128.8 155.2 155.5 155.8 156.1 156.3 156.6 156.8 157.1 157.4 157.6 157.9 158.1 158.4 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.1 140.5 B-7
I' GE Nuclear Energy NEDO-33144 Non-proprietary Version I
I TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-5, 5-11 & 5-13 I
I BOTTOM HEAD UPPER RPV &
BELTLINE AT 22 EFPY CURVE A
(°F)
BOTTOM HEAD CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE C
(°F)
I PRESSURE (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 CURVE A
(°F)
I 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 81.9 84.2 86.3 88.3 90.3 80.0 80.0 80.0 80.0 80.0 80.0 80.0 81.2 87.2 92.3 96.8 100.9 104.7 108.2 111.4 114.2 116.9 119.5 121.9 124.2 126.3 128.3 130.3 I
I I
I I
I I
I l
B-8 I
I
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 OF/hr for Curve A For Figures 5-5, 5-11 & 5-13 BOTTOM UPPER RPV &
BOTTOM UPPER RPV &
UPPER RPV &7No HEAD BELTLINEAT 22 EFPY CURVE A CURVEA (OF)
(OF)
HEAD BELTLINE AT 22 EFPY CURVE B CURVE B
(°F)
(OF)
BELTLINE AT 22 EFPY CURVE C (0F)
PRESSURE (PSIG) 230 240 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 92.1 93.9 95.6 97.2 98.8 100.3 101.8 103.2 104.5 104.9 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 132.1 133.9 135.6 137.2 138.8 140.3 141.8 143.2 144.5 144.9 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 B-9
I GE Nuclear Energy NEDO-33144 Non-proprietary Version I
TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 CF/hr for Curve A For Figures 5-5, 5-11 & 5-13 I
I I
BOTTOM HEAD UPPER RPV &
BELTLINE AT 22 EFPY CURVE A (OF)
BOTTOM HEAD CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE C
(°F)
PRESSURE (PSIG) 470 480 490 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 CURVE A
(°F)
I I
68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.7 70.1 71.5 72.9 74.2 75.5 76.8 78.1 79.3 80.4 81.6 82.7 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 I
I I
I I
I I
I A
1 B-1 0 I
I
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-5, 5-11 & 5-13 BOTTOM UPPER RPV &
BOTTOM UPPER RPV &
UPPER RPV& 1 HEAD BELTLINEAT 22 EFPY CURVE A CURVEA (0F)
(OF)
HEAD BELTLINE AT 22 EFPY CURVE B CURVE B (0F)
(0F)
BELTLINE AT 22 EFPY CURVE C (0F)
PRESSURE (PSIG) 730 740 750 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.3 69.4 70.5 71.5 72.6 73.6 74.6 75.5 76.5 77.4 78.4 79.3 80.1 81.0 81.9 82.7 83.6 84.4 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.2 110.9 111.5 112.1 112.7 113.3 113.9 83.8 84.9 86.0 87.0 88.0 89.0 90.0 90.9 91.9 92.8 93.7 94.6 95.4 96.3 97.1 98.0 98.8 99.6 100.4 101.1 101.9 102.7 103.4 104.1 104.9 105.6 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.1 140.5 140.9 141.2 141.6 141.9 142.3 142.6 143.0 143.3 143.7 144.0 144.4 144.7 145.0 145.4 145.7 146.0 146.4 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.1 180.5 180.9 181.2 181.6 181.9 182.3 182.6 183.0 183.3 183.7 184.0 184.4 184.7 185.0 185.4 185.7 186.0 186.4 B-11
I GE Nuclear Energy NEDO-33144 Non-proprietary Version I
TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-5, 5-11 & 5-13 I
I BOTTOM HEAD UPPER RPV &
BELTLINE AT 22 EFPY CURVE A
(°F)
BOTTOM HEAD CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE B
(°F)
UPPER RPV &
BELTLINE AT 22 EFPY CURVE C
(°F)
I I
PRESSURE (PSIG) 990 1000 1010 1020 1030 1040 1050 1060 1064 1070 1080 1090 1100 1105 1110 1120 1130 1140 1150 1155 1160 1170 1180 1190 1200 1210 CURVE A
(°F) 85.2 86.0 86.7 87.5 88.3 89.0 89.7 90.4 90.7 91.2 91.9 92.6 93.2 93.6 93.9 94.6 95.2 95.9 96.5 96.8 97.1 97.8 98.4 99.0 99.6 100.2 114.5 115.1 115.7 116.2 116.8 117.4 117.9 118.5 118.7 119.0 119.5 120.1 120.6 120.8 121.1 121.6 122.1 122.6 123.1 123.4 123.6 124.1 124.6 125.1 125.5 126.0 106.3 107.0 107.6 108.3 109.0 109.6 110.3 110.9 111.2 111.5 112.2 112.8 113.4 113.7 114.0 114.6 115.2 115.7 116.3 116.6 116.9 117.4 118.0 118.5 119.1 119.6 146.7 147.0 147.3 147.6 148.0 148.3 148.6 148.9 149.0 149.2 149.5 149.8 150.1 150.3 150.4 150.7 151.0 151.3 151.6 151.7 151.9 152.2 152.5 152.7 153.0 153.3 186.7 187.0 187.3 187.6 188.0 188.3 188.6 188.9 189.0 189.2 189.5 189.8 190.1 190.3 190.4 190.7 191.0 191.3 191.6 191.7 191.9 192.2 192.5 192.7 193.0 193.3 I
I I
I I
I I
I I
I J
B-12 I
I
GE Nuclear Energy NEDO-331 44 Non-proprietary Version TABLE B-2. Columbia Composite P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-5, 5-11 & 5-13 BOTTOM UPPER RPV &
BOTTOM UPPER RPV &
UPPER RPV &
HEAD BELTLINEAT 22 EFPY CURVE A CURVEA (OF)
(0F)
HEAD BELTLINE AT 22 EFPY CURVE B CURVE B (0F)
(OF)
BELTLINE AT 22 EFPY CURVE C (0F)
PRESSURE (PSIG) 1220 1230 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 100.8 101.3 101.9 102.5 103.0 103.6 104.1 104.7 105.2 105.7 106.3 106.8 107.3 107.8 108.3 108.8 109.3 109.8 110.3 126.5 126.9 127.4 127.8 128.3 128.7 129.2 129.6 130.0 130.5 130.9 131.3 131.7 132.1 132.6 133.0 133.4 133.8 134.2 120.2 120.7 121.2 121.7 122.2 122.7 123.2 123.7 124.2 124.7 125.2 125.6 126.1 126.6 127.0 127.5 127.9 128.4 128.8 153.6 153.9 154.2 154.4 154.7 155.0 155.2 155.5 155.8 156.1 156.3 156.6 156.8 157.1 157.4 157.6 157.9 158.1 158.4 193.6 193.9 194.2 194.4 194.7 195.0 195.2 195.5 195.8 196.1 196.3 196.6 196.8 197.1 197.4 197.6 197.9 198.1 198.4 B-1 3
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 1
BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(OF) 0 68.0 80.0 80.0 68.0 80.0 80.0 10 68.0 80.0 80.0 68.0 80.0 80.0 20 68.0 80.0 80.0 68.0 80.0 80.0 30 68.0 80.0 80.0 68.0 80.0 80.0 40 68.0 80.0 80.0 68.0 80.0 80.0 50 68.0 80.0 80.0 68.0 80.0 80.0 60 68.0 80.0 80.0 68.0 80.0 80.0 70 68.0 80.0 80.0 68.0 80.0 80.0 80 68.0 80.0 80.0 68.0 80.0 80.0 90 68.0 80.0 80.0 68.0 80.0 80.0 100 68.0 80.0 80.0 68.0 80.0 80.0 110 68.0 80.0 80.0 68.0 80.0 80.0 1
120 68.0 80.0 80.0 68.0 80.0 80.0 130 68.0 80.0 80.0 68.0 80.0 80.0 140 68.0 80.0 80.0 68.0 80.0 80.0 150 68.0 80.0 80.0 68.0 80.0 80.0 160 68.0 80.0 80.0 68.0 80.0 80.0 170 68.0 80.0 80.0 68.0 80.0 80.0 180 68.0 80.0 80.0 68.0 81.9 80.0 190 68.0 80.0 80.0 68.0 84.2 80.0 200 68.0 80.0 80.0 68.0 86.3 80.0 210 68.0 80.0 80.0 68.0 88.3 80.0 220 68.0 80.0 80.0 68.0 90.3 80.0 230 68.0 80.0 80.0 68.0 92.1 80.0 B-14
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY PRESSURE (PSIG)
HEAD CURVE A (0F)
VESSEL CURVE A (0F)
BELTLINE CURVE A (0F)
HEAD CURVE B (0F)
VESSEL CURVE B (0F)
BELTLINE CURVE B (0F) 240 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 93.9 95.6 97.2 98.8 100.3 101.8 103.2 104.5 104.9 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 80.0 80.0 80.0 80.0 80.0 80.0 81.2 82.5 82.9 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 B-15
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY j
Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 I
BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY l
HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(OF) 490 68.0 110.0 110.0 68.0 140.0 140.0 500 68.0 110.0 110.0 68.0 140.0 140.0 l
510 68.0 110.0 110.0 68.0 140.0 140.0 520 68.0 110.0 110.0 68.0 140.0 140.0 530 68.0 110.0 110.0 68.0 140.0 140.0 540 68.0 110.0 110.0 68.0 140.0 140.0 550 68.0 110.0 110.0 68.0 140.0 140.0 560 68.0 110.0 110.0 68.0 140.0 140.0 570 68.0 110.0 110.0 68.0 140.0 140.0 580 68.0 110.0 110.0 68.0 140.0 140.0 590 68.0 110.0 110.0 68.0 140.0 140.0 600 68.0 110.0 110.0 68.0 140.0 140.0 610 68.0 110.0 110.0 68.7 140.0 140.0 620 68.0 110.0 110.0 70.1 140.0 140.0 630 68.0 110.0 110.0 71.5 140.0 140.0 640 68.0 110.0 110.0 72.9 140.0 140.0 650 68.0 110.0 110.0 74.2 140.0 140.0 660 68.0 110.0 110.0 75.5 140.0 140.0 670 68.0 110.0 110.0 76.8 140.0 140.0 680 68.0 110.0 110.0 78.1 140.0 140.0 690 68.0 110.0 110.0 79.3 140.0 140.0 1
700 68.0 110.0 110.0 80.4 140.0 140.0 710 68.0 110.0 110.0 81.6 140.0 140.0 720 68.0 110.0 110.0 82.7 140.0 140.0l 730 68.0 110.0 110.0 83.8 140.0 140.0 740 68.0 110.0 110.0 84.9 140.0 140.0 750 68.0 110.0 110.0 86.0 140.0 140.0 1
B-16
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY PRESSURE (PSIG)
HEAD CURVE A (0F)
VESSEL CURVE A (0F)
BELTLINE CURVE A (OF)
HEAD CURVE B (OF)
VESSEL CURVE B (0F)
BELTLINE CURVE B (0F) 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 990 1000 1010 1020 68.0 68.0 68.0 68.0 68.0 68.3 69.4 70.5 71.5 72.6 73.6 74.6 75.5 76.5 77.4 78.4 79.3 80.1 81.0 81.9 82.7 83.6 84.4 85.2 86.0 86.7 87.5 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.2 110.9 111.5 112.1 112.7 113.3 113.9 114.5 115.1 115.7 116.2 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 87.0 88.0 89.0 90.0 90.9 91.9 92.8 93.7 94.6 95.4 96.3 97.1 98.0 98.8 99.6 100.4 101.1 101.9 102.7 103.4 104.1 104.9 105.6 106.3 107.0 107.6 108.3 140.0 140.0 140.0 140.0 140.1 140.5 140.9 141.2 141.6 141.9 142.3 142.6 143.0 143.3 143.7 144.0 144.4 144.7 145.0 145.4 145.7 146.0 146.4 146.7 147.0 147.3 147.6 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 B-17
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY 4
Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 l
BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY l
HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)
(OF)
(OF)
(OF)
(OF)
(OF)
(OF) 1030 88.3 116.8 110.0 109.0 148.0 140.0 1040 89.0 117.4 110.0 109.6 148.3 140.0 l
1050 89.7 117.9 110.0 110.3 148.6 140.0 1060 90.4 118.5 110.0 110.9 148.9 140.0 1064 90.7 118.7 110.0 111.2 149.0 140.0 1070 91.2 119.0 110.0 111.5 149.2 140.0 1080 91.9 119.5 110.0 112.2 149.5 140.0 1
1090 92.6 120.1 110.0 112.8 149.8 140.0 1100 93.2 120.6 110.0 113.4 150.1 140.0 1
1105 93.6 120.8 110.0 113.7 150.3 140.0 1110 93.9 121.1 110.0 114.0 150.4 140.0 1120 94.6 121.6 110.0 114.6 150.7 140.0 l
1130 95.2 122.1 110.0 115.2 151.0 140.0 1140 95.9 122.6 110.0 115.7 151.3 140.0 1
1150 96.5 123.1 110.0 116.3 151.6 140.0 1155 96.8 123.4 110.0 116.6 151.7 140.0 1160 97.1 123.6 110.0 116.9 151.9 140.0 1170 97.8 124.1 110.4 117.4 152.2 140.0 1180 98.4 124.6 111.1 118.0 152.5 140.0 1
1190 99.0 125.1 111.7 118.5 152.7 140.0 1200 99.6 125.5 112.4 119.1 153.0 140.0 1210 100.2 126.0 113.0 119.6 153.3 140.0 1220 100.8 126.5 113.6 120.2 153.6 140.0 1230 101.3 126.9 114.2 120.7 153.9 140.0 1
1240 101.9 127.4 114.8 121.2 154.2 140.3 1250 102.5 127.8 115.4 121.7 154.4 140.8 1260 103.0 128.3 116.0 122.2 154.7 141.3 B-18 l
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-3. Columbia P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 BOTTOM UPPER 33.1 EFPY BOTTOM UPPER 33.1 EFPY PRESSURE (PSIG) 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 HEAD CURVE A (0F)
VESSEL CURVE A (0F)
BELTLINE CURVE A (0F)
HEAD CURVE B (0F)
VESSEL CURVE B (0F)
BELTLINE CURVE B
(°F) 103.6 104.1 104.7 105.2 105.7 106.3 106.8 107.3 107.8 108.3 108.8 109.3 109.8 110.3 128.7 129.2 129.6 130.0 130.5 130.9 131.3 131.7 132.1 132.6 133.0 133.4 133.8 134.2 116.6 117.2 117.8 118.3 118.9 119.5 120.0 120.5 121.1 121.6 122.1 122.7 123.2 123.7 122.7 123.2 123.7 124.2 124.7 125.2 125.6 126.1 126.6 127.0 127.5 127.9 128.4 128.8 155.0 155.2 155.5 155.8 156.1 156.3 156.6 156.8 157.1 157.4 157.6 157.9 158.1 158.4 141.8 142.2 142.7 143.2 143.6 144.1 144.5 145.0 145.4 145.8 146.3 146.7 147.1 147.5 B-1 9
I GE Nuclear Energy NEDO-33144 Non-proprietary Version I
I TABLE B-4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-6, 5-12 & 5-14 I
I
.1 BOTTOM HEAD UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE A (OF)
BOTTOM HEAD CURVE B
(°F)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE B
(°F)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE C
(°F)
PRESSURE (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 CURVE A (OF)
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68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 81.9 84.2 86.3 88.3 90.3 80.0 80.0 80.0 80.0 80.0 80.0 80.0 81.2 87.2 92.3 96.8 100.9 104.7 108.2 111.4 114.2 116.9 119.5 121.9 124.2 126.3 128.3 130.3 I
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GE Nuclear Energy NEDO-331 44 Non-proprietary Version TABLE B4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-6, 5-12 & 5-14 BOTTOM UPPER RPV & BOTTOM HEAD BELTLINEAT 33.1 EFPY CURVE A CURVE A (OF)
(OF)
HEAD CURVE B
(°F)
UPPER RPV &
UPPERIRPV &-7 BELTLINE AT BELTLINE AT I 33.1 EFPY 33.1 EFPY I
PRESSURE (PSIG)
CURUVE tB
(°F)
CUKVt U
(°F) 230 240 250 260 270 280 290 300 310 312.5 312.5 320 330 340 350 360 370 380 390 400 410 420 430 440 450 460 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 80.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 92.1 93.9 95.6 97.2 98.8 100.3 101.8 103.2 104.5 104.9 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 132.1 133.9 135.6 137.2 138.8 140.3 141.8 143.2 144.5 144.9 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 B-21
I GE Nuclear Energy NEDO-331 44 Non-proprietary Version I
TABLE B-4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 1 00 'Ffhr for Curves B & C and 20 'F/hr for Curve A For Figures 5-6, 5-12 & 5-14 I
I BOTTOM HEAD U PP ER RPV &
BELTLINE AT 33.1 EFPY CURVE A (OF)
BOTTOM HEAD CURVE B (OF)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE B (OF)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE C (OF)
I PRESSURE (PSIG) 470 480 490 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 700 710 720 CURVE A (OF)
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68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.7 70.1 71.5 72.9 74.2 75.5 76.8 78.1 79.3 80.4 81.6 82.7 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.0 I
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GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 'F/hr for Curve A For Figures 5-6, 5-12 & 5-14 r -'----BOTTOM UPPER RPV &
BOTTOM UPPER RPV'&
UPPER"RPV &
HEAD BELTLINEAT 33.1 EFPY CURVEA CURVEA (0F)
(OF)
HEAD BELTLINE AT 33.1 EFPY CURVE B CURVE B (0F)
(0F)
BELTLINE AT 33.1 EFPY CURVE C (OF)
PRESSURE (PSIG) 730 740 750 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.3 69.4 70.5 71.5 72.6 73.6 74.6 75.5 76.5 77.4 78.4 79.3 80.1 81.0 81.9 82.7 83.6 84.4 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.0 110.2 110.9 111.5 112.1 112.7 113.3 113.9 83.8 84.9 86.0 87.0 88.0 89.0 90.0 90.9 91.9 92.8 93.7 94.6 95.4 96.3 97.1 98.0 98.8 99.6 100.4 101.1 101.9 102.7 103.4 104.1 104.9 105.6 140.0 140.0 140.0 140.0 140.0 140.0 140.0 140.1 140.5 140.9 141.2 141.6 141.9 142.3 142.6 143.0 143.3 143.7 144.0 144.4 144.7 145.0 145.4 145.7 146.0 146.4 180.0 180.0 180.0 180.0 180.0 180.0 180.0 180.1 180.5 180.9 181.2 181.6 181.9 182.3 182.6 183.0 183.3 183.7 184.0 184.4 184.7 185.0 185.4 185.7 186.0 186.4 B-23
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GE Nuclear Energy NEDO-33144 Non-proprietary Version l
TABLE B-4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-6, 5-12 & 5-14 I
I BOTTOM HEAD UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE A (OF)
BOTTOM HEAD CURVE B
(°F)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE B
(°F)
UPPER RPV &
BELTLINE AT 33.1 EFPY CURVE C
(°F)
I I
PRESSURE (PSIG) 990 1000 1010 1020 1030 1040 1050 1060 1064 1070 1080 1090 1100 1105 1110 1120 1130 1140 1150 1155 1160 1170 1180 1190 1200 1210 CURVE A (OF) 85.2 86.0 86.7 87.5 88.3 89.0 89.7 90.4 90.7 91.2 91.9 92.6 93.2 93.6 93.9 94.6 95.2 95.9 96.5 96.8 97.1 97.8 98.4 99.0 99.6 100.2 114.5 115.1 115.7 116.2 116.8 117.4 117.9 118.5 118.7 119.0 119.5 120.1 120.6 120.8 121.1 121.6 122.1 122.6 123.1 123.4 123.6 124.1 124.6 125.1 125.5 126.0 106.3 107.0 107.6 108.3 109.0 109.6 110.3 110.9 111.2 111.5 112.2 112.8 113.4 113.7 114.0 114.6 115.2 115.7 116.3 116.6 116.9 117.4 118.0 118.5 119.1 119.6 146.7 147.0 147.3 147.6 148.0 148.3 148.6 148.9 149.0 149.2 149.5 149.8 150.1 150.3 150.4 150.7 151.0 151.3 151.6 151.7 151.9 152.2 152.5 152.7 153.0 153.3 186.7 187.0 187.3 187.6 188.0 188.3 188.6 188.9 189.0 189.2 189.5 189.8 190.1 190.3 190.4 190.7 191.0 191.3 191.6 191.7 191.9 192.2 192.5 192.7 193.0 193.3 I
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GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE B-4. Columbia Composite P-T Curve Values for 33.1 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-6, 5-12 & 5-14 BOTTOM UPPER RPV &
BOTTOM UPPER RPV &
UPPER RPV &
HEAD BELTLINE AT 33.1 EFPY CURVE A CURVEA (0F)
'(F)
HEAD BELTLINE AT 33.1 EFPY CURVE B CURVE B (0F)
(0F)
BELTLINE AT 33.1 EFPY CURVE C (0F)
PRESSURE (PSIG) 1220 1230 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 100.8 101.3 101.9 102.5 103.0 103.6 104.1 104.7 105.2 105.7 106.3 106.8 107.3 107.8 108.3 108.8 109.3 109.8 110.3 126.5 126.9 127.4 127.8 128.3 128.7 129.2 129.6 130.0 130.5 130.9 131.3 131.7 132.1 132.6 133.0 133.4 133.8 134.2 120.2 120.7 121.2 121.7 122.2 122.7 123.2 123.7 124.2 124.7 125.2 125.6 126.1 126.6 127.0 127.5 127.9 128.4 128.8 153.6 153.9 154.2 154.4 154.7 155.0 155.2 155.5 155.8 156.1 156.3 156.6 156.8 157.1 157.4 157.6 157.9 158.1 158.4 193.6 193.9 194.2 194.4 194.7 195.0 195.2 195.5 195.8 196.1 196.3 196.6 196.8 197.1 197.4 197.6 197.9 198.1 198.4 B-25
GE Nuclear Energy NEDO-331 44 Non-proprietary Version APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1
GE Nuclear Energy NEDO-33144 Non-proprietary Version 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. A discussion of monitoring of vessel temperatures can be found in Section 4 of the pressure-temperature curve report prepared in 1989 [1].
I C-2
GE Nuclear Energy NEDO-33144 Non-proprietary Version C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by <200F 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 HeatuplCooldown 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 201F 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 bolt-up, 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 NEDO-33144 Non-proprietary Version 1
In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:
I Head flange bolt-up Leakage test (Curve A compliance) l Startup (coolant temperature change of less than or equal to 1000F in one hour period heatup) l Shutdown (coolant temperature change of less than or equal to 1000F in one hour period cooldown)
]
Recirculation pump trip, bottom head stratification (Curve B compliance)
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GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX C
REFERENCES:
- 1. T.A. Caine, 'Pressure-Temperature Curves Per Regulatory Guide 1.99, Revision 2 for the Dresden and Quad Cities Nuclear Power Stations", SASR 89-54, Revision 1, August 1989.
C-5
GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX D GE SIL 430 D-1
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version I
September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions 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.
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I Steam i temper from m, line pre TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)
Measurement Use Limitations dome saturation Primary measurement Must convert saturated 3ture as determined above 2120F for Tech steam pressure to 3in steam instrument Spec 100°F/hr heatup temperature.
I ssure and cooldown rate.
I Recirc suction line coolant temperature.
Primary measurement below 2120F for Tech Spec 100°F/hr heatup and cooldown rate.
Must have recirc flow.
Must comply with SIL 251 to avoid vessel stratification.
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Alternate measurement above 2120F.
When above 212cF need to allow for temperature variations (up to 10-150F lower than steam dome saturation temperature) caused primarily by FW flow variations.
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GE Nuclear Energy NEDO-33144 Non-proprietary Version 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 inlet coolant temperature RPV drain line coolant temperature Alternate measurement for Tech Spec 1 OO 0 F/hr cooldown rate when in shutdown cooling mode.
Primary measurement to comply with Tech Spec delta T limit between steam dome saturated temp and drain line coolant temperature.
Must have previously correlated RHR inlet coolant temperature versus RPV coolant temperature.
Must have drain line flow. Otherwise, lower than actual temperature and higher delta T's will be indicated Delta T limit is 1000F for BWR/6s and 1450F for earlier BWRs.
Primary measurement to comply with Tech Spec brittle fracture limits during cooldown.
Alternate information only measurement for bottom head inside/
outside metal surface temperatures.
Must have drain line flow. Use to verify compliance with Tech Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).
Must compensate for outside metal temperature lag during heatup/cooldown.
Should have drain line flow.
D-3
GE Nuclear Energy NEDO-33144 Non-proprietary Version I
TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
-I Measurement Use Limitations l
Closure head flanges outside surface T/Cs Primary measurement for BWR/6s to comply with Tech Spec brittle fracture metal temperature limit for head bolt-up.
Use for metal (not coolant) temperature. Install temporary T/Cs for alternate measurement, if required.
-I I
One of two primary measure-ments for BWR/6s for hydro test.
I RPV flange-to-shell junction outside surface T/Cs Primary measurement for BWRs earlier than 6s to comply with Tech Spec brittle fracture metal temperature limit for head bolt-up.
Use for metal (not coolant) temperature. Response faster than closure head flange T/Cs.
I I
One of two primary measurements for BWRs earlier than 6s for hydro test. Preferred in lieu of closure head flange T/Cs if available.
Use RPV closure head flange outside surface as alternate measurement.
I I
I RPV shell outside surface T/Cs Top head outside surface T/Cs Information only.
Information only.
Slow to respond to RPV coolant changes. Not available on BWR/6s.
Very slow to respond to RPV coolant changes. Not avail-able on BWR/6s.
I I
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D-4 I
GE Nuclear Energy NEDO-33144 Non-proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)
(Typical)
Measurement Bottom head outside surface T/Cs Use Limitations 1 of 2 primary measurements to comply with Tech Spec brittle fracture metal temperature limit for hydro test.
Should verify that vessel stratification is not present for vessel hydro.
(see SIL No. 251).
Primary measurement to comply with Tech Spec brittle fracture metal temperature limits during heatup.
Use during heatup to verify compliance with Tech Spec metal temperature/reactor pressure curves.
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 NEDO-33144 Non-proprietary Version 1
Product
Reference:
B21 Nuclear Boiler Prepared By: A.C. Tsang l
Approved for Issue:
Issued By:
I B.H. Eldridge, Mgr.
D.L. Allred, Manager Service Information Customer Service Information l
and Analysis Notice:
J 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 J
to any plant or facility other than GE BWRs as designed and furnished 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.SlLs are part of GE s I
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 SlLs.
GE J
assumes no responsibility for liability or damage, which may result from the use of information contained in SILs.
D-6
GE Nuclear Energy NEDO-331 44 Non-proprietary Version APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1
GE Nuclear Energy NEDO-33144 Non-proprietary Version I
I 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:
1 "The 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."
I 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 1.0e17 n/cm2. 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.
I I
and are specified I
The following dimensions are obtained from the referenced drawings as the distance above vessel "0":
-I Shell # 2 - Bounding Top of Active Fuel (TAF)*
366.3"
[1]
Shell # 1 - Bottom of Active Fuel (BAF) 216.3"
[1]
Centerline of Recirculation Outlet Nozzle N1 in Shell # 1 172.5"
[2]
Top of Recirculation Outlet Nozzle N1 in Shell # 1 193.5"
[3]
Centerline of Recirculation Inlet Nozzle N2 in Shell # 1 181.0"
[2]
Top of Recirculation Inlet Nozzle N2 in Shell # 1 195.9"
[4]
Centerline of Instrumentation Nozzle N12 in Shell # 2 368.0"
[5]
Centerline of LPCI Nozzle N6 in Shell #2 372.5"
[5]
Bottom of LPCI Nozzle N6 in Shell # 2 355.0"
[6]
- The fuel contained in the Columbia vessel includes rods of varying heights, where 150" is the bounding height. This is discussed in more detail in [1].
I I1 I
I
-I I
From [2], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -23" below BAF and the top of E-2 I
I
GE Nuclear Energy NEDO-33144 Non-proprietary Version the recirculation outlet nozzle is -20" below BAF).
As shown in [5], the N12 Instrumentation Nozzle is contained within the core beltline region; however, this nozzle has a thickness less than 2.5" and therefore requires no fracture toughness. As noted in Table A-2, components having a thickness of less than 2.5" do not require fracture toughness evaluations. The only other nozzle within the BAF-TAF region of the reactor vessel is the N6 LPCI Nozzle, which is included in the beltline evaluation.
Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm2, it can be safely concluded that all nozzles and welds, other than those included in Tables 4-5 and 4-6 and discussed above, are outside the beltline region of the reactor vessel.
Based on the axial fluence profile [1], the RPV fluence drops to less than 1.0e17 n/cm2 at -3" below the BAF and at -7" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 7" above the active fuel region and the additional 3" below the active fuel region. This adjusted beltline region extends from 213" to 372.5" above reactor vessel "0" for 33.1 EFPY.
Based on the above, it is concluded that none of the Columbia reactor vessel plates, nozzles, or welds, other than those included in Tables 4-5 and 4-6, are in the beltline region.
E-3
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
APPENDIX E
REFERENCES:
- 1.
Wu, T, "Energy Northwest Columbia Generating Station Neutron Flux Evaluation", GE-NE-0000-0023-5057-RO, Revision 0, GE-NE, San Jose, CA, I
April 2004 (GE Proprietary).
- 2.
Drawing #21, Revision 5, "#1 Shell Ring", CBI Nuclear Company, GENE, l
San Jose, California (GE VPF #3133-030).
- 3.
Drawing #46, Revision 4, "N1 Nozzle Forging (Recirculation Outlet)", CBI l
Nuclear Company, Chicago, Illinois (GE VPF #3133-055).
- 4.
Drawing #49, Revision 5, "N2 Nozzle Forging (Recirculation Inlet)", CBI l
Nuclear Company, Chicago, Illinois (GE VPF #3133-058).
- 5.
Drawing #22, Revision 3, "#2 Shell Ring Assembly", CBI Nuclear Company, l
Chicago, Illinois (GE VPF #3133-031).
- 6.
Drawing #67, Revision 6, "N6 Nozzle Assembly (RHR/LPCI Mode)", CBI 1 Nuclear Company, Chicago, Illinois (GE VPF #3133-076).
E-4_
GE Nuclear Energy NEDO-33144 Non-proprietary Version APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)
F-1
GE Nuclear Energy NEDO-33144 Non-proprietary Version Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy of the l
beltline materials.
The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be 33.1 EFPY.
Calculations of 33.1 EFPY USE, using l
Regulatory Guide 1.99, Revision 2 [2] methods and BWROG Equivalent Margin Analyses [3, 4] methods are summarized in Tables F-1 and F-2.
A Regulatory l
Guide 1.99, Revision 2 analysis is also provided for those materials where sufficient unirradiated upper shelf energy data was available in order to further demonstrate the l
qualification of these materials.
Unirradiated upper shelf data was not available for all of the material heats in the Columbia beltline region. Therefore, Columbia is evaluated to verify that the BWROG J
EMA is applicable. The USE decrease prediction values from Regulatory Guide 1.99, Revision 2 are used for the beltline components as shown in Tables F-1 and F-2. These calculations are based upon the 33.1 EFPY peak 1/4T fluence as provided in Tables 4-5 and 4-6.
Surveillance capsule data is available for Columbia. Columbia has removed and tested one surveillance capsule that was irradiated in the Columbia vessel [5]. Columbia has J
committed to participate in the BWRVIP Integrated Surveillance Program (ISP).
Information was provided from available surveillance capsule test results [6] for J
application to the Columbia vessel weld for use in determining the upper shelf energy.
Test results from [5] for the plate and weld materials, and [6] for the weld materials were l
applied to the USE EMA evaluation as seen in Tables F-1 and F-2. For Tables F-1 and F-2, unirradiated test results performed at the time that the irradiated surveillance l
capsule was evaluated [5] are calculated based upon the average of all test results where the % shear is 95% or greater.
Based on the results presented in Tables F-1 and F-2, the USE EMA values for the Columbia reactor vessel beltline materials remain within the limits of Regulatory I
Guide 1.99, Revision 2 and 10CFR50 Appendix G for 33.1 EFPY of operation. Further, Table F-3 demonstrates that the upper shelf energy for those materials where sufficient I
unirradiated information is available remain well above 50 ft-lb at 33.1 EFPY.
F-2
GE Nuclear Energy NEDO-33144 Non-proprietary Version Table F-1 Equivalent Margin Analysis Plant Applicability Verification Form for Columbia including Power Uprate Conditions 33.1 EFPY BWR/3-6 PLATE Surveillance Plate USE (Heat B5301-1):
%Cu
=
0.11 Unirradiated USE
=
98.0 ft-lb 1st Capsule Measured USE
=
99.6 ft-lb 1stCapsuleFluence
=
1.55E+17 n/cm2 1st Capsule Measured % Decrease
=
-1.6 1st Capsule RG 1.99 Predicted % Decrease
=
8.0 (Charpy Curves)
(RG 1.99. Rev. 2, Figure 2)
Limiting Beltline Plate USE (Heat C1337-1 and C1337-2):
%Cu
=
0.15 33.1 EFPY1/4TFluence
=
5.11E+17 n/cm2 (Cumulative Energy Provided in Fluence Report)
RG 1.99 Predicted % Decrease
=
12.0 (RG 1.99, Rev. 2, Figure 2)
Adjusted % Decrease
=
N/A (RG 1.99, Rev. 2, Position 2.2) 12.0%
21.5%
Therefore, vessel plates are bounded by Equivalent Margin Analysis F-3
GE Nuclear Energy NEDO-33144 I
Non-proprietary Version I
Table F-2 Equivalent Margin Analysis Plant Applicability Verification Form for Columbia including Power Uprate Conditions 33.1 EFPY BWR/2-6 WELD I
I Surveillance Weld USE (Heat 3P4966):
%Cu Unirradiated USE 1st Capsule Measured USE 1st Capsule Fluence 0.03 98.0 ft-lb 108.0 ft-lb 1.55E+17 n/cm2 I
11
.1 1st Capsule Measured % Decrease
=
-10.2 1 st Capsule RG 1.99 Predicted % Decrease
=
6.0 (Charpy Curves)
(RG 1.99, Rev. 2, Figure 2)
ISP Surveillance Weld USE (Heat 5P6756):
l
%Cu Unirradiated USE River Bend 1830 Capsule Measured USE River Bend 183° Capsule Fluence SSP Capsule F Measured USE SSP Capsule F Fluence SSP Capsule H Measured USE SSP Capsule H Fluence River Bend 183° Capsule Measured % Decrease River Bend 183° Capsule RG 1.99 Predicted % Decrease SSP Capsule F Measured % Decrease SSP Capsule F RG 1.99 Predicted % Decrease SSP Capsule H Measured % Decrease SSP Capsule H RG 1.99 Predicted % Decrease 0.06 104.4 ft-lb 84.4 ft-lb 1.16E+18 n/cm2 79.3 ft-lb 1.94E+18 n/cm2 84.6 ft-lb 1.36E+18 n/cm2 I
-1 19.2 12.5 24.0 14.0 19.0 13.0 (Charpy Curves)
(RG 1.99, Rev. 2, Figure 2)
(Charpy Curves)
(RG 1.99, Rev. 2, Figure 2)
(Charpy Curves)
(RG 1.99, Rev. 2, Figure 2)
I1 I
Limiting Beltline Weld USE (Heat 6240391D205A27A):
%Cu 33.1 EFPY 1/4T Fluence (Cumulative Energy Provided in Fluence Report)
RG 1.99 Predicted % Decrease Adjusted % Decrease
=
0.10
=
5.11E+17 n/cm2 I
=
12.0
=
18.0 (RG 1.99, Rev. 2, Figure 2)
(RG 1.99, Rev. 2, Position 2.2)
I 18.0%
34.0%
Therefore, vessel welds are bounded by Equivalent Margin Analysis F-4 I
I I
I
GE Nuclear Energy NEDO-33144 Non-proprietary Version Table F-3 Upper Shelf Energy Evaluation for Columbia at 33.1 EFPY 33.1 Initial EFPY 33.1 EFPY Transverse 114T Decrease Transverse Material Heat or Heat/Lot USE(1"
%Cu Fluence USEt 21 USEt31 (ft-lb)
(ft-lb)
Plates:
Lower-Intermediate Shell 85301-1 98 0.13 5.11E+17 14 84 Welds:
Vertical:
Lower Shell 3P4966/1214/3482(4) 98 0.025 1.75E+17 6
92 Lower-intermediate Shell 3P4966/121413481(4) 98 0.025 5.11E+17 7.5 91 Girth:
AB 5P6756/0342/3447 Single 91 0.08 2.13E+17 9
83 AB 5P6756/0342/3447 Tandem 97 0.08 2.13E+17 9
88 AB 3P4955/0342/3443 Single 90 0.027 2.13E+17 6.5 84 AB 3P4955/0342/3443 Tandem 95 0.027 2.13E+17 6.5 89 Notes:
- 1. Initial USE and %Cu obtained from [6].
- 2. USE Decrease obtained from RG1.99 Figure 2.
- 3. 33.1 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE 1100)].
- 4. This evaluation includes both Single and Tandem Wire materials.
F-5
GE Nuclear Energy NEDO-33144 Non-proprietary Version 1
APPENDIX F
REFERENCES:
X
- 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.
A
- 3. J.T. Wiggins (NRC) to L.A. England (Gulf States Utilities Co.), "Acceptance for
]
Referencing of Topical Report NEDO-32205, Revision 1, '10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels"', December 8, 1993.
- 4. L.A. England (BWR Owners' Group) to Daniel G. McDonald (USNRC), "BWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin Analysis - Approved Version", BWROG-94037, March 21, 1994.
- 5. CL Chu, "Washington Public Power Supply System WNP-2 RPV Surveillance Materials Testing and Analysis", GE-NE, San Jose, CA, March 1997 (GE-NE-B1301809-01, Revision 0).
- 6. a) Letter #EN2-PE-04-002, JJ Sisk (Energy Northwest) to JE Larsen (GE),
"Contract No. C-31814, Data Package Transmittal No. 002 For: RPV Pressure Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, January 13, 2004.
l b) Letter #EN2-PE-04-009, JJ Sisk (Energy Northwest) to JE Larsen (GE),
"Contract No. C-31814, Data Package Transmittal No. 003 For: RPV Pressure l
Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, February 17, 2004.
l c) Letter #EN2-PE-04-016, JJ Sisk (Energy Northwest) to JE Larsen (GE),
"Contract No. C-31814, Data Package Transmittal No. 005 For: RPV Pressure l
Temperature Curves, Fluence Calculations, Fluence Maps and Associated Services for Columbia Generating Station, March 23, 2004.
l F-6 I
GE Nuclear Energy NEDO-33144 Non-Proprietary Version APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION G-1
i GE Nuclear Energy NEDO-33144 Non-Proprietary Version G.1 Objectives:
I The purpose of the following evaluations is to determine the hydrotest and the heat-up/cool-1 down temperature (T) for the shell thickness transition discontinuity in the beltline and the bottom head and to demonstrate that the temperature is bounded by the beltline hydrotest and l
heat-up/cool-down temperature.
G.2 Methods and Assumptions:
A ANSYS finite element analyses were performed for the thickness discontinuities in the beltline l
and bottom head of the Columbia vessel. The purpose of this evaluation was to determine the RPV discontinuity stresses (hoop and axial) that result from the thickness transition discontinuity J
in the beltline region and the bottom head.
The transition in the beltline is modeled as a transition from 6.1875" minimum thickness to 9.5" minimum thickness [1]. The bottom head J
thickness transition is modeled as 6 3/16" to 8" from the radial plates to the dollar plates, respectively.
For conservatism, the minimum thickness of the thinner plate to the maximum thickness of the thicker plate was modeled in both the beltline and bottom head analysis. This captures the greatest possible discontinuity and the greatest stress concentration at the discontinuity.
Three load cases were evaluated for the beltline and bottom head shell discontinuity:
I
- 1) hydrostatic test pressure at 1020 psig, 2) a cool-down transient of 100F/hr, starting at 5520F and decreasing to 700F on the inside surface wall [2] and with an initial operating pressure of 1
1050 psig, and 3) a heat-up transient of 1000F/hr, starting at 70'F and increasing to 5520F on the inside surface wall [2] and with a final operating pressure of 1050 psig for the beltline region l
and 1075 psig for the bottom head region. For both transient cases it was assumed that the outside RPV wall surface is insulated with a heat transfer coefficient of 0.2 BTU/hr-ft2 'F [3] and that the ambient temperature is 100'F.
These are the bounding beltline transients of those described in Region B of the thermal cycle diagram [2] at temperatures for which brittle fracture could occur. Additionally, the bottom head was analyzed for Case 4) a turbine generator trip of
-200°F/hour with a corresponding step increase from 450°F to 544°F, as a bounding transient j
associated with Region C of the thermal cycle diagram [2]. Material properties were used from the Code of construction for the RPV Materials: Shell and Bottom Head Plate Materials are
]
G-2
GE Nuclear Energy NEDO-33144 Non-Proprietary Version ASME SA533, Grade B, Class 1, low alloy steel (LAS) and Bottom Head Skirt Materials are ASME SA516 Grade 70 [4].
The operating pressure used for the cooldown transient introduces conservatism as the peak thermal stresses occur at the end of the cooldown, where the pressure is less than 1050 psig.
Methods consistent with those described in Section 4.3 were used to calculate the T-RTNDT for the shell discontinuity for a hydrotest pressure of 1020 psig and the two transient cases. The adjusted reference temperature values shown in Table 4-6 were added to the T-RTNDT to determine the temperature "T". The value of 'T' was compared to that of the beltline region for the same condition as described in Sections 4.3.2.2.1 for the hydrotest pressure case and 4.3.2.2.4 for the transient cases.
The Control Rod Drive Penetrations in the bottom head were not evaluated as a part of this analysis. The stub tubes provide sufficient stiffness that the deletion of these penetrations from this analysis is acceptable.
It is demonstrated in this analysis that Curve A (Figures 5-1 and 5-4) bounds the temperatures found for the hydrostatic test pressure Temperatures from the FEA analysis. It is also shown that Curve B (Figures 5-7 and 5-10) bounds the temperatures found for Transient Pressures from the stresses obtained in the FEA analysis. Therefore, the transition discontinuity stresses in the beltline and bottom head are bounded by the Curves.
The locations of maximum stress were evaluated in the beltline shell and bottom head evaluation; these locations are displayed in Figures G-1 and G-2, respectively.
The methods of ASME Code Section Xl, Appendix G [5] are used to calculate the pressure test and thermal limits. The membrane and bending stress were determined from the finite element analysis and are shown below. The hoop stresses were more limiting than the axial stresses; the hoop stresses are provided in Tables G-1, G-2, G-3 and G4 of this appendix.
The stress intensity factors, Klm and Klb, are calculated using 1998 ASME Code with Addenda through 2000 Section Xl Appendix A 17] and Appendix G [5], as shown in Section 4.3.2.2.2 of G-3
GE Nuclear Energy NEDO-33144 I
Non-Proprietary Version this report. Therefore, Kim= Mm*am and Kib = Mb*Cyb. The values of Mm and Mb were determined from the ASME Code Appendix G [5]. The stress intensity is based on a 1/4 T radial flaw with a six-to-one aspect ratio (length of 1.5T). The flaw is oriented normal to the maximum stress direction, in this case a vertically oriented flaw since the hoop stress was limiting.
The calculated value of Kim + Klb is multiplied by a safety factor (SF) (1.5 for pressure test and 2.0 for the transient cases), per ASME Appendix G [5] for comparison with KIR, the material fracture toughness expressed as KIc.
The relationship between Kjc and temperature relative to reference temperature (T - RTNDT) is provided in ASME Code Section Xl Appendix A [7] Paragraph A-4200, represented by the relationship (K, units ksi-in 05):
Kjc = 33.2 + 20.734 exp [0.02 (T - RTNDT) ]; therefore, T-RTNDT = In [ (Kic-33.2) / 20.734 ] / 0.02, where Kjc = SF * (Kim + Kib) for pressure test and Kic = (SF
- Kip) + Kis for transient case.
This relationship is derived in the Welding Research Council (WRC) Bulletin 175 [8] as the lower bound of all dynamic fracture toughness data. This relationship provides values of pressure versus temperature (from KIR and (T - RTNDT), respectively).
The RTNDT is added to the (T-RTNDT) to determine the hydrotest, heat-up, and cool-down temperatures.
I I
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I j1 I
I I
I I
I I
Analysis Information:
Beltline Thin Section Thickness tmin = 6.1875 inch
\\l(t) = 2.49 inch05 Thick Section Thickness tmax= 9.75 inch q(t) = 3.12 inch05 Bottom Head Thin Section Thickness tmin = 6.1875 inch q(t) = 2.49 inch05 1
Thick Section Thickness tmax = 8 inch l
4(t) = 2.83 inch05 G-4
1' 1' '"
l-,
[****
1-*I I-IFi.
17 1.
1 -i F "-I K-I I'
a-1.*
- 1.
r -
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version Analvsis and Results for the Hydrotest Pressure (Case 1):
Primary Primary Membrane Bending Pb Km = Mm*Pm Mb = 213 T - RTNDT Pressure Location Surface Pm (psi)
(psi)
Mm (psi in 12)
Mm Kb K!
(7F) 1000 1
Inside 20,570 659 1000 1
Outside 20,570
-659 1020 1
Inside 20,981 672 2.30 48,257 1.53 1,030 49.29 33.76 1020 1
Outside 20,981
-672 2.30 48,257 1.53
-1,030 47.23 29.82 1000 2
Inside 18,690
-924 1000 2
Outside 18,690 924 1020 2
Inside 19,064
-942 2.30 43,847 1.53
-1,445 42.40 19.14 1020 2
Outside 19,064 942 2.30 43,847 1.53 1,445 45.29 25.80 1000 3
Inside 12,760 365 1000 3
Outside 12,760
-365 1020 3
Inside 13,015 372 2.85 37,093 1.90 707 37.80 6.26 1020 3
Outside 13,015
-372 2.85 37,093 1.90
-707 36.39 1.53 Table G-1: Analysis Results for Case 1 for Beltline Shell Discontinuity G-5
GE Nuclear Energy NEDO-33144 Non-Proprietary Version Primary Primary Km =
Membrane Bending Mm*Pm Mb = 2/3 T - RTNDT Pressure Location Surface Pm (psi)
Pb (psi)
Mm (psi in 112)
Mm Kb KI (OF) 1000 1
Inside 9,242
-1,285 2.30 21,257 1.53
-1970 28.93 1000 1
Outside 9,242 1285 2.30 21,257 1.53 1970 34.84 -126.84 1020 1
Inside 9,427
-1,311 2.30 21,682 1.53
-2010 29.51 1020 1
Outside 9,427 1,311 2.30 21,682 1.53 2010 35.54 -109.14 1000 2
Inside 7,617 699 2.30 17,519 1.53 1072 27.89 1000 2
Outside 7,617
-699 2.30 17,519 1.53
-1072 24.67 1020 2
Inside 7,769 713 2.30 17,869 1.53 1094 28.45 1020 2
Outside 7,769
-713 2.30 17,869 1.53
-1094 25.16 1000 3
Inside 7,647
-57 2.30 17,588 1.53
-88 26.25 1000 3
Outside 7,647 57 2.30 17,588 1.53 88 26.51 1020 3
Inside 7,800
-58 2.30 17,940 1.53
-90 26.78 1020 3
Outside 7,800 58 2.30 17,940 1.53 90 27.04 1000 4
Inside 8,405 1,524 2.30 19,332 1.53 2337 32.50 1000 4
Outside 8,405
-1,524 2.30 19,332 1.53
-2337 25.49 1020 4
Inside 8,573 1,554 2.30 19,718 1.53 2384 33.15 1020 4
Outside 8,573
-1,554 2.30 19,718 1.53
-2384 26.00 Table G-2: Analysis Results for Case 1 for Bottom Head Discontinuities Note that the axial stress is approximately 1/2 of the hoop stress, hoop stresses presented in Tables.
- Transients that did not have sufficient stresses to cause a meaningful T-RTNDT are omitted for clarity (KITOTAL<3 3.2 psi in12)
G-6 I
I I
I I
I I
GE Nuclear Energy NEDO-33144 Non-Proprietary Version G.3 Results and Conclusions for Hydrostatic Pressure Case The results of this analysis prove that Curve A (Figures 5-1 and 5-4) remains bounding for the bottom head and beltline shell discontinuities, respectively.
Beltline The maximum Columbia plant-specific T-RTNDT calculated with the linearized stresses from the Finite Element Analysis (FEA) for the beltline thickness discontinuity is 33.760F as shown in Table 4-1. The limiting beltline weld material RTNDT (ART) at the region of the discontinuity is 340F (see Table 4-6b), so T = 67.760F. The limiting beltline plate RTNDT (ART) at the region of the discontinuity is 630F (see Table 4-6a), so T = 96.760F.
At 1020 psig, the T-RTNDT for the beltline region Curve A is 36.20F at 1020 psig (see Section 4.3.2.2.2), and T = 99.20F (see Section 4.3.2.2.2).
However, Curve A is limited by 10CFR 50 Appendix G requirements of the Boltup Temperature + 900F, which gives a value of 110F (as shown in Tables B-1 and B-3).
Because the beltline region pressure test temperature 'T' of 110'F is greater that the T=96.760F, obtained with the FEA analysis result, the thickness discontinuity remains bounded by the beltline curve.
Bottom Head The maximum T-RTNDT calculated with the Finite Element Analysis results for the bottom head region is -109.4°F, as shown in the Table G-2. The maximum RTNDT for the bottom head is 20°F for the plates (see Table 4-1) and 10°F for the welds (from Table 4-3). Thus a value of T = -89.4 0F is obtained from the linearized stresses obtained in the FEA analysis.
From Section 4.3.2.1.1, the T used in the analysis is 87.50F at 1020 psig. This value bounds the maximum value obtained using the linearized stresses from the FEA analysis.
G-7
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version Hoop Stress Pressure Primary Primary Secondary Secondary M
Kip K,
Case (pi)Location Surface MembranePm Bending membrane Bending Sb Mm23Mm(s 112) (pi1n/2)
KTOTAL T-RTNDT (psi)
Pb (psi)
Sm (psi)
(psi)
(p
) (p Pressure 1000 1
Inside 20,570 659 Pressure 1000 1
Outside 20,570
-659 Heat Up 1050 1
Inside 21,599 692 159
-5,474 2.30 1.53 50,737
-8,027 93,448 53.33 Heat Up 1050 1
Outside 21,599
-692 159 5,474 2.30 1.53 48,616 8,760 105,992 62.79 Cool Down 1050 1
Inside 21,599 692
-120 4,643 2.30 1.53 50,737 6,843 108,317 64.36 Cool Down 1050 1
Outside 21,599
-692
-120
-4,643 2.30 1.53 48,616
-7,395 89,836 50.24 Pressure 1000 2
Inside 18,690
-924 Pressure 1000 2
Outside 18,690 924 Heat Up 1050 2
Inside 19,625
-970
-641
-8,621 2.30 1.53 43,649
-14,694 72,604 32.10 Heat Up 1050 2
Outside 19,625 970
-641 8,621 2.30 1.53 46,624 11,744 104,992 62.10 Cool Down 1050 2
Inside 19,625
-970 589 7,475 2.30 1.53 43,649 12,815 100,113 58.58 Cool Down 1050 2
Outside 19,625 970 589
-7,475 2.30 1.53 46,624
-10,108 83,140 43.95 Pressure 1000 3
Inside 12,760 365 Pressure 1000 3
Outside 12,760
-365 Heat Up 1050 3
Inside 13,398 383
-449
-13,290 2.85 1.90 38,912
-26,531 51,293
-6.81 Heat Up 1050 3
Outside 13,398
-383
-449 13,290 2.85 1.90 37,457 23,971 98,884 57.65 Cool Down 1050 3
Inside 13,398 383 371 11,690 2.85 1.90 38,912 23,269 101,093 59.31 Cool Down 1050 3
Outside 13,398
-383 371
-11,690 2.85 1.90 37,457
-21,153 53,760
-0.42 Table G-3 : Beltline Analysis and Results for Cool-down (CD - Case 2) and Heat-up (HU - Case 3):
G-8
F-
[-
r--"
n-r--
rF---
F-F-
r-r-
r--
r--
r-Fr-r-
n GE Nuclear Energy NEDO-33144 Non-Proprietary Version Primary Primary Secondary Secondary Case Pressure Pressure Heat Up Heat Up Cooldown Cooldown Pressure Pressure Heat Up Heat Up Cooldown Cooldown Pressure Pressure Heat Up Heat Up Cooldown Cooldown Pressure (psig) 1000 1000 1050 1050 1050 1050 1000 1000 1050 1050 1050 1050 1000 1000 1050 1050 1050 1050 Location Surface Membrane Bending membrane Pm (psi)
Pb (Psi)
Sm (Psi) 1 Inside 9,242
-1,285 1
Outside 9,242 1,285 1
Inside 9,704
-1,349
-6,096 1
Outside 9,704 1,349
-6,096 1
Inside 9,704
-1,349 2,858 1
Outside 9,704 1,349 2,858 Bending Sb (psi)
-9,870 9,870 13,330
-13,330
-864 864 9,311
-9,311 Mb=
Mm 2/3 Mm 2
2 2
2 2
2 3
3 3
3 3
3 Inside Outside Inside Outside Inside Outside Inside Outside Inside Outside Inside Outside 7,617 7,617 7,998 7,998 7,998 7,998 7,647 7,647 8,029 8,029 8,029 8,029 699
-699 734
-734 734
-734 2.30 2.30 2.30 2.30 2.30 2.30 2.30 2.30 1.53 1.53 1.53 1.53 1.53 1.53 1.53 1.53 Kip (PSI in 1'2) 20,251 24,388 20,251 24,388 19,521 17,269 19,521 17,269 Kis (PSi KITOTAL inl12)
-29,155 11,346 1,113 49,890 27,013 67,514
-13,866 34,911
-22,588 16,454
-19,939 14,599 23,136 62,179
-5,417 29,121
-10.85 25.19
-124.75 16.74 T-RTNDT
-9,245
-9,245 3,852 3,852
-57 57 423
-416 423
-416
-3,357
-3,357 1,887 1,887
-11,840 11,840 13,290
-13,290 2.30 2.30 2.30 2.30 1.53 1.53 1.53 1.53 19,116 17,830 19,116 17,830
-25,876 12,355 10,434 46,093 24,718 62,949
-16,038 19,621
-23.76 18.05 Pressure Pressure Heat Up Heat Up Cooldown Cooldown 1000 1000 1050 1050 1050 1050 4
4 4
4 4
4 Inside Outside Inside Outside Inside Outside 8,405 8,405 8,825 8,825 8,825 8,825 1,524
-1,524 1,600
-1,600 1,600
-1,600
-1,540
-1,540 2,023 2,023
-10,660 10,660 14,700
-14,700 2.30 2.30 2.30 2.30 1.53 1.53 1.53 1.53 22,752 17,844 22,752 17,844
-19,887 25,616 12,803 48,492 27,193 72,696
-17,887 17,802
-15.22 32.22 Tablea G-4: Bottom Head Analvsis and Results for Heat-un (HU - Case 31 and Turbine Trip (TG -Case 4)
Note: The Cases for Heat Up and Cooldown Provided in Table 4 represent the limiting of the cases, the Heat Up Case provided is the Startup of the Reactor, and the Cooldown Case is the -200¶ Decrease in internal temperature of the reactor during a turbine trip transient.
^ Transients that did not have sufficient stresses to cause a meaningful T-RTNDT are omitted for clarity (KiToTAL< 3 3.2 psi into)
G-9
GE Nuclear Energy NEDO-33144 Non-Proprietary Version LOC3L FigureG-1 Loationsof EvauationDisconinuitisinBtle G-10
~.__
~
~
~
~
~
I.
~.
L
-~m
17-1,---
If If_
I I ---
1' 1'
I- '-'
I '-
I '---
l i
I --
I ---- 1--
I '-'-
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version 3
Location 2 Figure G-2: Locations of Evaluation Discontinuities in Bottom Head G-11
GE Nuclear Energy NEDO-33144 Non-Proprietary Version G.4 Results and Conclusions for Transient Cases:
The results of the discontinuity analysis show that the linearized stresses in the beltline and bottom head are bounded by Curve B for the Beltline and Bottom Head, Figures 5-10 and 5-7, respectively.
BeItline The maximum Columbia plant-specific T-RTNDT for the thickness discontinuity in the beltline is 64.40F (Table G-3). The limiting beltline material RTNDT in the region of the discontinuity is 630F (see Table 4-6a), so T = 127.40F.
At 1050 psig, the beltline is limited by the beltline Curve B (Figure 5-10). The T-RTNDT l
for the beltline region is 65.20F at 1050 psig, and T = 128.20F (see Section 4.3.2.2.4).
Because the beltline region pressure test temperature "T" of 129.60F bounds the limiting thickness discontinuity for the case with the limiting ART value (T = 102.40F for the weld J
material in the region of the discontinuity), the thickness discontinuity remains bounded by the beltline curve.
Bottom Head The maximum Columbia plant-specific T-RTNDT for the thickness discontinuity in the l
bottom head is 32.220F.
The maximum RTNDT for the bottom head is 200F (see Table 4-1) for the plates and 10F for the welds (from Table 4-3).
This yields a l
maximum value of T= 52.22°F.
From Section 4.3.2.1.2, the T used in the analysis is 110.30F at 1050 psig. This value bounds the maximum value obtained using the linearized stresses from the FEA l
analysis.
The results of the discontinuity analysis show that the Curves are bounding.
G-12
GE Nuclear Energy NEDO-33144 Non-Proprietary Version Appendix G
References:
8,
'Vessel Outline",
(GE VPF #3133-001, Rev. 5).
b) CBI Nuclear Company Drawing #11, Rev. 5, "Bottom Head Dollar Plates",
(GE VPF #3133-021, Rev. 5).
c) CBI Nuclear Company Drawing #12, Rev. 7, "Bottom Head Radial Plates", (GE VPF #3133-022, Rev. 5).
d) CBI Nuclear Company Drawing #6, Rev. 5, "Seam Details (Shell)",
(GE VPF #3133-016, Rev. 5).
- 2. GE Drawing Number 762E120, "Reactor Vessel Thermal Cycles", GE-NED, San Jose, CA, Revision 0 (GE Proprietary).
- 3. "Reactor Vessel Purchase Specification, Reactor Pressure Vessel", (21A9431, Revision 0), December 1971.
- 4. RPV QC Surveillance and Records Summary, Hanford II GE PO# 205-AE023, "General Electric Company BWR Projects Department QA-Engineered Equipment and Installation", May 1976.
- 5. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section III or Xl of the ASME Boiler and Pressure Vessel Code, 1998 Edition with Addenda through 2000.
- 6. Not used.
- 7. "Analysis of Flaws", Appendix A to Section Xl of the ASME Boiler and Pressure Vessel Code, 1998 Edition with Addenda through 2000.
- 8. "PVRC Recommendations on Toughness Requirements for Ferritic Materials",
Welding Research Council Bulletin 175, August 1972.
G-13
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version APPENDIX H CORE NOT CRITICAL BOTTOM HEAD (CRD PENETRATION)
EXAMPLE CALCULATION H-1
GE Nuclear Energy NEDO-33144 Non-Proprietary VersionI I
I i
I i1 I
I1 I
I H-2
GE Nuclear Energy NEDO-33144 Non-Proprietary Version H-3
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version I
I I
I II
-I H-4
GE Nuclear Energy NEDO-331 44 Non-Proprietary Version H-5
L L
H,E 7Ud
, ergyl L
f5 CJrtni" A.4venue L
Jo!s.ol)s.
CA 95125 L
L L
L L
L L
L L
L