Text

Revised Non-Proprietary GE Report GE-NE-A22-00100-08-01a-R2, Pressure-Temperature Curves for Duane Arnold Energy Center
	ML032390387
Person / Time
Site:	Duane Arnold
Issue date:	08/18/2003
From:	Branlund B, O'Connor M, Tilly L; General Electric Co
To:	; Office of Nuclear Reactor Regulation
References
	DRF 0000-0018-2033, NG-03-0608, TAC MB8750 GE-NE-A22-00100-08-01a-R2
	Download: ML032390387 (144)
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Attachment 2 to NG-03-0608 August 18, 2003 Revised Non-Proprietary GE Report-GE-NE-A22-00100-08-Ola-R2

w. GE Nuclear Energy Engineering and Technology GE-NE-A22-00100-08-Ola-R2 General Electric Company Revision 2 175 Curtner Avenue DRF 0000-0018-2033 San Jose, CA 95125 Class I August 2003 Non-PopretaryVenlon Pressure-Temperature Curves For Duane Arnold Energy Center Prepared by: 9U(C O'Connor M. C. OConnor, Mechanical Engineer Structural Mechanics and Materials Verified by: LI 2i-V L.J. illy, Senior Engineer Structural Mechanics and Materials Approved by: qlg Branfund B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-A22001 O-OM8a-R2 Non-Proprietary Version REPORT REVISION STATUS Revision Purpose I The report was revised to incorporate a Regulatory Guide 1.190 compliant fluence.

2 The report was revised to correct proprietary markings and to clarify fluence information, N2 nozzle beitline calculations, and Figure 4-2 and the associated discussion.

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GE Nuclear Energy GE-NE-A22-001000-a-R2 Non-Proprietary Version 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 Nuclear Management Company and GE, PO# 17168 Duane Arnold Energy Center Asset Enhancement Program, effective December 12, 2002, 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 NuclearManagement Company, 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 lability as to the completeness, accuracy, or usefulness of the Information contained in this document, or that Its use may not Infringe privately owned rights.

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-GE Nuclear Energy GE-NE-A22-00100-08-0le-R2 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 behiine. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 (1I. The P-T curve methodology includes the following: ) The ncorporation of ASME Code Case N-640. 2) The use of the Mm calculation In the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of madmum stress. ASME Code Case N-640 allows the use of K.c of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNwT. This report incorporates a fluence calculated in accordance with the GE Ucensing Topical Report NEDC-32983P, which has been approved by the NRC In SER [143, and is in compliance with Regulatory Guide 1.190.

CONCLUSIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [21:

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 100F/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

.V.

GE Nuctear Energy GPE-NE-A22-00100-08-01a-R2 Non-Proprietary Version nozzle thermal cycle diagrams 3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20"F/hr or less must be maintained at all times.

The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, K,, at 14T to be less than that at 314T for a given metal temperature.

Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beltline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beftline (at 25 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions. A composite P-T curve was also generated for the Core Critical condition at 25 EFPY.

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GE Nuclear Energy GS-NE-A22-00100-08-01a-R2 Non-Proprietary Verslon TABLE OF CONTENTS

1.0 INTRODUCTION

1 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITIL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 13 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 18

6.0 CONCLUSION

S AND RECOMMENDATIONS 51

6.0 REFERENCES

66

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GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 40 APPENDIX E DETERMINATION OF BELTUNE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EVALUATION FOR UPPER SHELF ENERGY EQUIVALENT MARGIN ANALYSIS APPENDIX G FLUENCE EVALUATION

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GE Nuclear Energy G-NE-A22-0010s00-s-R2 Non-Proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE DUANE ARNOLD RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2: CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 30 FIGURE 4-3: FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 35 FIGURE 5-1: B(TrOM HEAD P-T CURVE OR PRESSURE EST [CURVE Al 20OFHR OR LESS COOLANTHEATUP/COOLDOWN] 54 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A 120F/HR OR LESS COOLANT HEATUPCOLDOWN] 55 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE Al UP TO 25 EFPY 120 0 F/HR OR LESS COOLANT REAUPICOOLDOWNJ 56 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE Al UP TO 32 EFPY 120 0F/HR OR LESS COOLANT HEATUP/COOLDOWNI 57 0 F/HR OR FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE Bl f0 LESS COOLANT HEATUP/COOLDOWNJ 58 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B1 1100F/HR OR LESS COOLANT HEATUP/COOLDOWNJ 59 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL ICURVE Bl UP TO 25 EFPY

[1000F/HR OR LESS COOLANT HEATUPCOOLDOWN] 60 FIGURE 54: BELTLINE P-T CURVES FOR CORE NOT CITICAL [CURVE B] UP TO 32 EFPY

[100FIHR OR LESS COOLANT HEATUP/COOLDOWN 61 FIGURE -9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C1UP TO 25 EFPY 100 0F/HR OR LESS COOLANT HEATUP/COOLDOWNI 62 FIGURE 5-10: COMPOSITE PRESSURE TEST P-T CURVES [CURVE Al UP TO 32 EFF'Y [200FR OR LESS COOLANT HEATUPICOOLDOWN] 63 FIGURE 5-1 1: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE Bl UP TO 32 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 FIGURE 5-12: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C1 UP TO 32 EFPY [100°F/HR OR LESS COOLANT HEATUP/0OOLDOWN1 65

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GE Nuclear Energy GE-NE-A22-00100-08 01a-R2 Non-Proprietary Version TABLE OF TABLES TABLE 4-1: RT, VALUES FOR DUANE ARNOLD VESSEL MATERIALS 11 TABLE 4-2: RTr VALUES FOR DUANE ARNOLD NOZZLE, WELD, & BOLTING MATERIALS 12 TABLE 4-3: DUANE ARNOLD BELTLINE ART VALUES (25 EFFY) 16 TABLE 4-4: DUANE ARNOLD BELTLINE ART VALUES (32 EFPY) 17 TABLE 4-5:

SUMMARY

OF THE 10CFR50 APPENDIX 0 REQUIREMENTS 20 TABLE 4.6: APPLICABLE BWRI4 DISCONINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 22 TABLE 4-7: APPLICABLE BWR14 DISCONTIN COMPONENTS FOR USE WITH CRD (BOTIOM HEAD) CURVES A&B 23 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 53

GE Nuclear Energy GE-NE-A22 00108001a-R2 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 rradiation embrittlement effects in the betline.

Complete P-T curves were developed for 25 and 32 effective full power years (EFPY).

The P-T curves are provided In Section 5.0 and a tabulation of the curves Is Included In Appendix B. The P-T curves incorporate a fluence calculated In accordance with the GE Ucensing Topical Report NEDC-32983P, which has been approved by the NRC in SER 14], and Is In compliance with Regulatory Guide 1.190. This fluence Is discussed In Section 4.2.1.2 and Appendix G.

The methodology used to generate the P-T curves in this report Is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves In 2000 (1]. The P-T curve methodology includes the following: 1) The ncorporation of ASME Code Case N-640 [4]. 2) The use of the Ma calculation in the 1995 ASME Code paragraph G-2214.1 6] for a postulated defect normal to the direction of maxImum stress. ASME Code Case N-640 allows the use of Ktc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTWT. P-T curves are developed using geometry of the RPV shells and discontinuifies, the initial RTNDT of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.

The initial RTI,= is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section il of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the nitial RTNDT values are tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTwr is documented In Section 4.1.

Adjusted Reference Temperature (ART) Is the reference temperature when Including Irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 7] provides the methods for calculating ART. The value of ART is a function of RPV 114T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 25 and 32 EFPY are included in Section 4.2. The 32 EFPY peak ID fluence value of

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GE Nuclear Energy Git-NE-A22001M008-018-R2 Non-Proprietary Version 4.17 x 1016n cr 2 used In this report is discussed in Section 4.2.1.2 (also, see Appendix G). Beltline chemistry values are discussed in Section 4.21.1.

Comprehensive documentation of the RPV discontinuities that are considered in this report Is Included In Appendix A. This appendix also Includes a table that documents which non-beline discontinuity curves are used to protect the discontinuities.

Guidelines and requirements for operating and temperature monitoring are ncluded in Appendix C. GE SIL 430, a GE service nformation letter regarding Reactor Pressure Vessel Temperature Monitoring Is Included In Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the N16 Instrumentation and N2 Recirculation Inlet nozzles) are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

A discussion of the effect of a change In irradiation on the embrittlement of the RPV materials due to the increased flux associated with an increase in core thermal power from 1658 MWO, to 1912 MWjh, which is 115% of current rated thermal power, Is ncluded In Appendix G.

1 GE Nuclear Energy GE-NE-A22-00100 08-01a-R2 Non-Proprietary Version 2.0 SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report Is similar to the methodology used to generate the P-T curves in 2000 [11. The P-T curves In this report incorporate a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER 141, and is in compliance with Regulatory Guide 1.190. This fluence Is discussed In Section 42.1.2 and Appendix G. A detailed description of the P-T curve bases is included in Section 4.3.

The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTwT. Other features presented are:

Generation of separate curves for the upper vessel in addition to those generated for the beltiTne, 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 [81 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 M7.

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-beitline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable Duane Arnold vessel components. The non-beldine 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 bettline region. This refinement could minimize heating requirements prior to pressure testing, Operating and temperature monitoring GE Nuclear Energy GE-NE-A22-00100- a-Ola-R2 Non-Proprietary Version requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the N16 Instrumentation and N2 Recirculation Inlet nozzles) are outside the beltline region.

Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

A discussion of the effect of a change in irradiation on the embrittlement of the RPV niaterials due to the increased flux associated with an increase in core thermal power from 1658 MWth to 1912 MW,, which is 115% of current rated thermal power, Is Included in Appendix G. [

D GE Nuclear Energy GE-NE-A220100-08-01a-R2 Non-Proprietary Version 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

For end-of-license (32 EFPY) fluence an 80% capacity factor is used to determine the EFPY for a 40-year plant life. The 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (refueling outages, etc.

account for -20% of the year).

The shutdown margin is calculated for a water temperature of 6WF, as defined in the Duane Arnold Technical Specification, Section 1.1.

For the N2 Recirculation Inlet and N16 Instrumentation nozzles, the chemistry value used In the analysis was assumed to be 0.18% copper, this Is explained in Tables 4-3 and 44.

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GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTwT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section 1I1, 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 ftlb.

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 RTmT value are significantly different For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section III, Subsection NB-2300 are as follows:

a Test specimens shall be transversely oriented (normal to the rolling direction) CVN specimens.

b. RTNDT Is defined as the higher of the dropwelght NDT or 600F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion Is met.

c. Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest RTNoT of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever Is greater.

10CFR50 Appendix G (5] states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented In an approved manner. GE developed methods for analytically GE Nuclear Energy GE-NE-A22 00100-08-01a-R2 Non-Proprietary Version converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data in WRC Bulletin21719] and from data collected to respond to NRC questions on FSAR submittals In the late 1970s. In 194, these methods of estimating RTNDT were submitted for generic approval by the BWR Owners' Group 10], and approved by the NRC for generic use 111.

4.1.2 Values of Initial RTmT and Lowest Service Temperature (LST)

To establish the nitial RTNDT temperatures for the Duane Arnold vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDr calculations for vessel plate, weld, HAZ, and forging, and bolting material LST are summarized In the remainder of this section.

For vessel plate material, the first step in calculating RTwT Is to establish the 50 ft-lb transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs 112]). For Duane Amold CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 2F per ft-lb energy difference from 50 ft-lb.

For example, for the Duane Arnold beltline plate heat C6794-2 in the upper shell course, the lowest Charpy energy and test temperature from the CMTRs is 33.0 ft-lb. at 1 0F.

The estimated 50 ft-lb longitudinal test temperature Is:

TOL = 10F + (50 - 33) fl-lb.

2Fft-lb.I = 440F 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, TsOT = 440F + 30°F = 74eF GE Nuclear Energy GE-NE-A22-0010D00al-R2 Non-Proprietary Version The nitial RTT is the greater of nil-ductility transition temperature (NDT) or (Tan 600 F).

Dropweight testing to establish NDT for plate material was listed in the CMTR; the NDT for the case above was 10F. Thus, the Irtial RTNEJ for plate heat C6794-2 was 140F.

For the Duane Arnold beltline weld heat 432Z0471 (contained in the lower 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 30°F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat 432ZO471 has a lowest Charpy energy of 100 ft-lb at 10OF as recorded in weld qualification records. Therefore, TsT = 100F - 600 F = -500 F.

The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TT-609F). For Duane Amold, the dropwelght testing to establish NDT was not recorded for most weld materials. - GE procedure requires that, when no NDT is available for the weld, the resulting RTNDr should be -60rF or higher. The value of (TwT - 06 F) In this example is -50eF; therefore, the intial RTW was -50 F.

For the vessel HAZ material, the RTIDT 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 dosure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the feedwater nozzle at Duane Arnold (Heat Q2Q6V), the NDT is 400F and the lowest CVN data is 87 f-b at 40'F. The corresponding value of (TeT- 600F) is:

(Tsor - 600F) = 4 0F + 300F- 60F = 100F.

Therefore, the initial RTwT is the greater of nil-ductility transition temperature (NDT) or (Tear 600F), which is 400F.

.GE Nuclear Energy GE-ME-A22-00100-08-Ola-R2 Non-Proprietary Version In the bottom head region of the vessel, the vessel plate method Is applied for estimating RTNDT. For the center plate of Duane Arnold (Heat B03903), the NDT is 40°F and the lowest CVN data was 71 ft-b at 40*F. The corresponding value of (ThT - 600F) was:

(TwT - 60F) =40°F + 30 0F- DF =10F.

Therefore, the initial RTN=, being the greater of (TOT - 60 0F) or the NDT, was 400F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported Is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-b test temperature plus 600F (as discussed in Section 4.3.2.3) is the LST for the boling materials, Charpy data for the Duane Arnold closure studs do not meet the 45 ft-lb, 25 MLE requirement at 100F, but the CVN energy was greater than 30 ft-lb.

Therefore, the LST for he bolting material Is 700 F. The highest RTNoT in the closure flange region is 140F, for the vessel upper shell (Shell Ring #4) materials. Thus, the higher of the LST and the RTm.T 600 F Is 740 F, the boltup limit In the closure flange region.

The Initial RTNDT values for the Duane Arnold reactor vessel (refer to Figure 4-1 for the Duane Arnold schematic) materials are listed in Tables 4-1 and 4-2. This tabulation includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves.

GE Nuclear Energy GE-NE-A2200100-08-Ola-R2 Non-Proprietary Version BNLOSURE P-0 CLOSURE RANGE REGION UPPER SHELL LD

. .. f.

4 tI

'UP PER INTERMEDIATE EU f LOWER INTERMEDIATESHELL LO'S ITUDINAL SEAM WELD 12 CORE SELUINE PLATE 1-2 HEAT 3D436-2 0* PLATE11 HEAT B0673-1 RECON E~~oMR LOWE LONGfIODINAL SEAM WELDE1 SHEL GHeLD D1 S0 - aRCUMFERENTIAL GIRTH WELD DE 4

FLATE 1 18 HE!AT C64392 PLATE 1-HEAT 30402-1 0 "WELDo2 SOTTOM HEAD J 1

- R------- ' '> ' '

Notes: (1)Refer to Tables 41 nd 4-2 for reactor vessel components and their heat identficatons (2) See Appendix E for the definion of the beline region.

Figure 4-1: Schematic of the Duane Arnold RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Table 4-1: RTNDT Values for Duane Arnold Vessel Materials TEST Tm c ARPY KNERGY (T=T40) DROP Tw Long (Fr4.8) rF) NOT rF 4, lLATES lORGINGS: _l l epHad &Flange __

dlanger(Veo Pee 126 BYA242 10 L 114 96 67 -20 10 10 Flane P 127 30CM44 10 L 69 99 O -20 10 10 opHoodDoePlcoe4 .0 40 L -6 149 111 10 40 40 THead Sde Pts_

Pb2 1ru 1.30 C6491-aA 10 L 77 84 107 -20 10 10 Peoes 131 ru 13 C69148 10 L 131 145 132 -20 10 10 6~whe _ _ _

Rcg44 Pke t-24 C6704-2 10 L 37 44 33 14 10 14 Rb £4 Pce -2 C7090-1 10 L 81 6 7 -20 10 10 nt. -hell _ =_=

Rha 83 Pi 1-22 10402-2 40 L 78 88 S 1a 40 40 Rbgt3 Plow 1 _23 _ C641-1 40 _ 103 101 _1S 10 40 40 LD . hea .______

62 ee 1o-20 80436-2 40 L 57 C4 62 10 -30 10 higi2 Pece 12 W731 40 L 39 104 121 10 - 10 g Pce 1-4_ C6439 40 L 36 48 43 38 40 40 t1 Pce 119 5041 40 L 83 86 72 10 40 40 otm Head _ .

CnerPis _ __ _ _ .__

ee 50390-3 40 L 90 T7 71 10 40 40 i 1Amd 1-28 B40043 40 L 2 l ea 10 401 40 P 1-3 C6491-2A 40 L 63 ea 74 10 40 40 5 Pt __

Pie*os 14b1u 14 80400.1 40 L 33 40 41 44 40 44 1-9 firm-?u 3400-2 40 L 01 7 72 10 40 40 8WPORT SKIRT __=_=_=_

PCMK1-1017 0390-1 40 L 96 149 111 10 40 40 5ABLIR UACKETS _

MRS" 80390-1 40 L 6 149 Ill 10 40 40 PCMK 542 C4629-2 10 L 79 8 49 -18 10 10 SFROUO SWPORT _ _ ___

CUK 48-1-1 4 L51590-2 = = = =40 40 PCIK 48-1-3 LS1S80-1A 40 40 H 42 L5189 =___ 40 -40 PCK( 9-1 L61377 - - 40 40 GE Nuclear Energy GE-NE-A22-0100-08-0la-R2 Non-Proprietary Version Table 4-2: RTNDT Values for Duane Arnold Nozzle, Weld, & Bolting Materials TEST TsosCAYENRYTw6 CROPRTC COMPONENT HEAT TEMP. ARP -L Y l OI W l)T, NOZZLES, __ ===_=_=

N1 Rtir. Odlel Nob _ Q17W 2 40 L as BO 3 1l 40 40 N2 RecInbt Nm~le 0 40 j J2 54 13 40 hf3 Stam OuLW MM _ .9 OD J 10 . 40 40 NSCamn Sy NzlQ 40 L go 14 103 10 40 40 N6 Hed 8e AIndtumentanti Nozzle C2Q1VW 40 L 120 3 10 40 40 NVent Nozzle Pbea 39-L Q2QYW . _ 1L. .747S.240475 1Q_ _4Q 4 N8 Jet Pim h,,trmeian Noe .- 29WL. 40 L _M _L IQ 14 _ 40 N9 CRD Hyd Sys Rtn Nozzhe 1VW 40 .JdL AL 4 t4 40 40 N10 Core Of POessue NozIz E20YW 40 L 38.0 20L. 40 50 Nil kmtumentaleon Nozzle Q :159W40 . L 44 3A 40 40 N12 Wtmiffn oui E20 A L. _ flB _L _

N1Uoaon~Jlzzhl QIQMWJ 4 .L --. _. 4Q N14 Sea Lek Detection Nozzle !!Q 40 IL 33 40 2 70 40 70 lo le_ ._

QLQ2W 40 AL 18 19 I Is 4 4 74 N16 lnmmentalon Nonzle 02O5'V 40 L 44 42 36 38 40 40 WELDS: _ _

Shel Rkm " S2ZmL. 10 We 3L 4 A06 0 40 Se RhQ R2 43245" 10 llh W 94 .6 -0 _ 60 Ginh Welds ___ _ __

Rhn 1 to Rmq 2 07UD 10 nha 60 50 54 60 _ 80 STUDS: 1504 () 10 e 4 48 70 OK Piee 23-1 16966 10 l 0 66 68 10 OK NUTS and WASHERS, ____ _ _ _ _

Pi.ce 70-1 and 70.2 8230 ( 10 a 40 S 38 70 OK BUSHtNGS: Sm ) I 10 nh 40 37 3s 70 OK Piece 24-1 ?? 10 DM 504760 la OK (a) Thee mtedefl h e t LST keased from 10 to 70 F because they do no meet the reuirement for 45 ft-bs. and 25 I&LE.

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2

- Non-Proprietary Version 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BEL TUNE The adjusted reference temperature (ART) of the limiting belfline material Is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (Rev 2) provides the methods for determining the ART. The Rev 2 methods for determining the limiting material and austing the P-T curves using ART are discussed In this section. As discussed In Appendix E, the beitline contains the N2 Recircutation Inlet nozzle and the N16 Instrumentation nozzle, which represents a slight extension beyond the core region. This is determined by the location on the vessel where the fluence exceeds 1x10V n/ck. An evaluation of ART for all betline plates, the N2 and N16 nozzles, and several beitline welds was made and summarized In Table 4-3 for 25 EFPY and Table 4-4 for 32 EFPY.

4.2.1 Regulatory Guide 1.99, Revision 2 (Rev 2) Methods The value of ART is computed by adding the SHIFT term for a given value of effective full power years (EFPY) to the initial RTw,. For Rev 2, the SHIFT equation consists of two terms:

SHIFT = ARTNDT + Margin where, &RTwT= [CF*Tf2 010 l-2 Margin 2(CT2 + rA )0 CF = chemistry factor fcom Tables 1or 2 of Rev. 2 f sT fluence l O'9 Margin - 2(cl2 + a02)03 J= standard deviation on initial RTNDT, which is taken to be 0°F.

ca standard deviation on ARTT, 28-F for welds and 17F for base material, except that qA need not exceed 0.50 times the ART= value.

ART = Initial RTNDo + SHIFT GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version The margin term ,& has constant values in Rev 2 of 170F for plate and 280 F for weld.

However, iA,need not be greater than 0.5 ARTHDr. 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 a, s taken to be 0F for the vessel plate and weld materials.

4.2.1.1 Chemistry The vessel beltline copper and nickel valuer (except for the N2 and N16 nozzles) were obtained from [5]. For the N2 and N16 nozzles, a bounding value of 0.18% was assumed for copper (see Tables 4-3 and 4-4), and the nickel values for N16 and N2 of 0.85% and 0.84%, respectively, were obtained from a Certified Material Test Report 1121.

The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively. For Plate Heat 80673-1, the CF was adjusted using Section 2.1 of Rev. 2; a detailed description of the adjustnent is included in 5). The margin term, a&, has constant values in Rev 2 of 170F for plate and 280 F for weld. For Plate Heat B0673-1, the margin term was halved, consistent with the guidance in Section 2.1 of Rev. 2.

However, a, need not be greater than 0.5*&RT.T. Since the GE1BWROG 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 S60flb level, the value of a Is taken to be 0 0F for the vessel plate materials and girth weld.

4.2.1.2 Fluence A Duane Arnold flux for the vessel ID wall was calculated in accordance with the method described in GE licensing Topical Report NEDC-32983P, which has been approved by the NRC [14], and is in compliance with Regulatory Guide 1.190. Peak fluxes have been determined for both the current rated power 1658 MW and EPU power of 1912 MWt.

The peak RPV ID fluence used in the P-T curve evaluation is 4.17e18 ncm2, as discussed in detail in Appendix G. (( ii for the entire plant life. This fluence applies to the lower-intermediate plates and longitudinal welds. The fluence is adjusted for the lower plates and longitudinal welds and the girth GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version weld based upon an attenuation factor of 1.18; hence, the peak ID surface fluence for these components is 3.55e18 n/cm2 . Similarly, the fluence is adjusted for the N2 nozzle based upon an attenuation factor of 5.46; hence the peak ID surface fluence used for this component Is 7.64e17 ncm2 . The same method Is applied to the N16 nozzle, which has an attenuation factor of 3.7, resulting Ina peak ID surface fluence of 1.13e18 n/Cm2 .

4.2.2 Limiting Beftline Matenial The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTMDT. Using initial RTwT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART. For Duane Arnold, the N2 Recirculation Inlet nozzle Is the limiting material for the beitline region as discussed in Section 4.3.2.22 and Appendix E Table 4-3 lists values of beltline ART for 25 EFPY and Table 4-4 lists the values for 32 EFPY. Sections 43.2.2.2 and 4.3.2.2.3 provide a discussion of the limiting material.

GE Nuclear Energy GE-NE-A2200100-08-Ola-R2 Non-Proprietary Version Table 4-3: Duane Arnold Beltine ART Values (25 EFPY)

L9 U11l.9 WN.Mkka 34L150WtN "

188418. *44 8458.

25 WIS&IMhU. 3*343 lUAOI.. 4.440h40 A _4h8m* 5... 2.3711z

)awrli .tA4 *w.

El wM%*%8. m?88-. 3.73841A;

11. . 4A44 NICE85 212331 13.. E$I-bo L 14F346 P1134-440 Inch" tdkvo 4u 844 3*VV~8.W.1811 AW43t 25 W3VF*t.llT-m. 434 SLATIM~~~~~~~~~~~~~~~~~~

ME 44 IF 4 2 231 13W I"P us v23851 80 I119 UM4 an .4 am I31II to U 173il ad.3 "543 ME4 I IIE13424 8.8 U4 us1 55 3*4 484 OA 17 J9.# 1USE 1I III 43ZIIniVt3 8.3 tkI It 3SN S In SUBM3 ILI3 Si4 1:13 43-7 341 8111'SW 96" 0 4,0 Ut 3 88 1.in A"8 81.8 In 153 PA4 Si 33.33 4~~~1318471LATA SM11 on1 of 1314 lb,

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IA. hi" 14 am a1w NW

?*411" 3j-A

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3+/- M" W"

~~~~~011 lets LA UAVA *M 11A 3 se ~~I". so 35 "31 h+/- 4 been 422300 Ld.4IxPWA WE *A a3 .13 tE* 4. 133 NA 3AI 11JI 1M242LWANUA2A "I on3 II do3 1mF 1 1%I as u 1.4 u 2is ana3111? 141 1 41 171* I Du32 4 Nu 1 "13n me MAW~IRAWV 4.3 an . 3 13 WE 71is WA. 6.6 31 MA* 12 3 M 41"AdIA1110AW GM .3I "' - 0 .J I18 2..4... 1A III. MA S ~ li 44W CrNi.S.i tI~* VA 33131am am. 34 Sp 31 . 27t0h1 2&4uf W AMb 03W4o a8hwnndw~gsEI.*1~h~mm 9~1,eb.ae d3E ~n. h1.SS*t0kSE144311

-17 -

GE Nucliear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) IOCFR50 Appendix G 181 specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over Its seice lifetime. The ASME Code (Appendix G of Section Xl of the ASME Code [6D forms the basis for the requirements of IOCFR5O 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 ctincal 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 [21:

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNDT. The remaining portion of the vessel (i.e., upper vessel, lower vessel) includes shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.

For e core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1000 Fhr 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 GE Nuclear Energy GE-NE-A220100 01a-R2 Non-Proprietary Version nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20F1hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 114T and 3/4T locations. When combining pressure and thermal stresses, it Is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 34T 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 waf during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location Is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 114T location. This approach Is conservative because irradiation effects cause the allowable toughness, Kf, at 14T to be less than that at 34T 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 heatuplcooldown oune limits.

The applicable temperature Is the greater of the 10CFR5O Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-5:

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version Table 4-5: Summary of the I OCFR50 Appendix G Requirements Operating Condition and Pressure Minimum Temperature Requirement

1. Hydrostatic Pressure Test & Leak Test (Core Is Not Critical) - Curve A

1. At < 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region Initial RTND + 60F*

2. At > 20% of preservice hydrotest jLarger of ASME Limits or of highest pressure closure flange region Initial RTNwT + 90F II. Normal operation (heatup and cooldown),

including anticipated operational occurrences

a. Core not critical - Curve B 1.At I 20% of preserve hydrotest Larger of ASME Limits or of highest pressure closure flange region Initial RTwT + 60°F*

2. At> 20% of preservice hydrotest Larger of ASME imits or of highest pressure closure flange region initial RT + 120 0F

b. Core critical - Curve C

1. At c 20% of preservice hydrotest Larger of ASME Umits + 40°F or of a.1 pressure, with the water level within the normal range for power operation

2. At > 20% of preservice hydrotest Larger of ASME Umits + 40°F or of pressure a2 + 40OF 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-beitline regions). The closure flange region limits are controdling at lower pressures primarily because of IOCFR50 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.

[1 GE Nudear Energy GE-NE-A2200100-08-01a-R2 Non-Proprietary Version 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Belfine Regions Non-beltline regions are defined as the vessel locations that are remote frorn the active fuel and where the neutron fluence is not sufficient (11.0e17 nicm2 ) to cause ary signricant shift of RTNwT (see Appendix E). Non-beltline components include nozzles, the closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations.

Detaied stress analyses of the non-beltline components were performed for the BWRl6 specifically for the purpose of fracture toughness analysis. The analyses took Into account all mechanical loading and anticipated thermal transients. Transients considered include 100°F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or aw, loss of recirculation pump flow, and all transients Involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code 16] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T- RTwT). Plots were developed for the limiting BWR/6 components: the feedwater nozzle (FVV) and the CRD penetration (bottom head). All other components in the non-bettline regions are categorized under one of these two components as described In Tables 4-6 and 4-7.

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Table 4-6: Applicable BWR/4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B DiscontinuityIdentificationt FW Nozzle CRD HYD System RetuR Core Spray Nozzle Recirculatlon Inlet Nozzle Steam Outlet Nozzle Maln Closure Flange Support Skit Stabilzer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water nterface Jet Pump Instrunmentation Nozzle Shell CRD and Bottom Head (Curve B only)

Top Head Nozzles (Curve B only)

Recirculation Outlet Nozzle (Curve B only)

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version Table 4-7: Applicable BWR/4 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuityldentification CRD and Bottom Head Top Head Nozzle Recirculation Outlet Nozzle Sheir Support Skirtr Shroud Support Attachment*

Core AP and Liquid Control Nozzle*

These dlscontinulites are added to the bottom head curve discontinuity list to assure that the entire bottom head Is covered, since separate bottom head P-T curves are provided to monitor the bottom head The P-T curves for the non-beltline region were conservatively developed for a large BWR/6 (rnominal inside diameter of 251 Inches). The analysis is considered appropriate for Duane Arnold as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at Duane Arnold by using plant specific RTtm values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.

GE Nuclear Energy GE-NE-A22-00100 0-Ola-R2 Non-Proprietary Version D

4.3.2.1.1 PressureTest - Non-Betllne, Curve A (Using Bottom Head)

In a [ I]finite element analysis [ D. the CRD penetration region was modeled to compute the local stresses for determination of the stress Intensity factor. K.

The (( J evaluation was modified to consider the new requirement for M,

- as discussed in ASME Code Section Xl Appendix G 16] and shown below. The results of-that computation were K; 143.6 ksi-inm2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T- RTNT) was 841F. ((

The limit for the coolant temperature change rate Is 2001F7hr or less.

[t D

GE Nuclear Energy GBE-NE-A22 00100 08 Ola-R2 Non-Proprietary Version The value of Mm for an Inside axial postulated surface flaw from Paragraph G-2214.1 161 was based on a thickness of 8.0 Inches; hence, tn = 2.83. The resulting value obtained was:

Mm 1.85 for -ftc2 Mm = 0.926 4t for 254t3.464 = 2.6206 M, = 3.21 for f>3.464 Klm Is calculated from the equation In Paragraph G-2214.1 [6] and Kb Is calculated from the equation in Paragraph G-2214.2 6]:

Kim M Gpm (( ))ksi-ln"2 Kb = (3) Mm c-,a U [ ksi-ln10 The total Kg is therefore:

f = 1.5 (m+ Kb) + Mm (am + (213) 'ao)= 143.6 ksl-Wn1 This equation includes a safety factor of 1.6 on primary stress. The method to solve for (T - RTN,) for a specific K1Is based on the Kl,equation of Paragraph A-4200 In ASME Appendix A [171:

(T - RTwT) = In (K - 33.2)120.734]I 0.02 (T- RTWT) =In (144 - 33.2) 20.734JI 0.02 (T- RT T)= 84F The generic curve was generated by scaling 143.6 ksl-in"2 by the nominal pressures and calculating the associated (T- RTT):

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Piroprietary Version Pressure Test CRD Penetration K}and (T - RTNDT) as a Function Of Pressure Nominal Pressure T - RTNDT (psig) (ksi-in (F -

1563 144 84 1400 129 77 1200 111 66 1000 92 52 800 74 33 600 55 3 400 37 -88 The highest RTwT for the botom head plates and welds is 44 0F, as shown in Tables 4-1 and 4-2. [

GBE Nuclear Energy GE-NE-A22 00100 08-0le-R2 Non-Proprietary Version Second, the P-T curve is dependent on the calculated 1 value, and the 14 value s proportional to the stress and the crack depth as showin below 14a:

~(a 8)" (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, 8, is tV4. Thus, Y is proportional to R/(t)'2. The generic curve value of R/(t)" 2, based on the generic BWRJ3 bottom head dimensions, is:

Generic. R / (t"2= 138 (8)12 9 inch (4-2)

The Duane Arnold specific bottom head dimensions are R = 92.69 inches and t =6.5 inches minimum 19], resulting in:

Duane Arnold specific: R (t)"2 = 92.69/ (6.5)12 = 36 inch" 2 (4-3)

Since the generic value of R/(t)"'2 Is larger, the generic P-T curve Is conservative when applied to the Duane Arnold bottom head.

4.2.1.2 Core Not CriticalHeatupfCooldown - Non-Beltine 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. ff GE Nuclear Energy GE-NE-A22-O1D-08-Ola-R2 Non-Proprietary Version 1

The calculated value of 1 for pressure test s multiplied by a safety factor (SF) of 1.5, per ASME Appendix G 6] for comparison with KR, 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"'.

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Therefore, the method to solve for (T - RTmT) for a specific K Is based on the N, equation of Paragraph A-4200 InASME Appendix A [171 for the core not critical curve:

(T - RTCT) = In [(K - 33.2)/ 20.734]/ 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.7341/0.02 (T- RTwT) = 102'F The generic curve was generated by scaling 192 ksi-ln"2 by the nominal pressures and calculating the associated (T - RTwT):

Core Not Critical CRD Penetration Kg and (T - RTWT) as a Function of Pressure Nominal Pressure T - RTj,=

(psg) 2 (ksi-i ) (OF) 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 67 600 74 33 400 49 -14 The highest RT for the bottom head plates and welds is 440F. as shown in Tables 4-1 and 4-2. [

As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Tables 4-6, 4-7, and Appendix A). Wth 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 GE-NE-A22-0100-08-O1a-R2 Non-Proprietary Version

[I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I 4.3.21.3 Pressure Test - Non-Belftlne Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress Intensity factor, 1K, for the feedwater nozzle was computed using the methods from WRC 175 15] together with the nozzle dimension for a generic 251-nch BWR/6 feedwater nozzle. The result of that computation was K 200 kslifn"2 for an applied pressure of 1563 psig preservice hydrotest pressure.

GE Nuclear Energy Gt-Ne-A22oo100 08-01a-R2 Non-Proprietary Version 11 The respective law depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1I4T through the corner thickness.

To evaluate the results, Is calculated for the upper vessel nominal stress, P, according to the methods in ASME Code Appendix G (Section III or Xi). The result is compared to that determined by CEIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of 1 Is shown below using the BWR/6, 251-inch dimensions:

Vessel Radius, R, 126.7 Inches Vessel Thickness, t 6.1875 Inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psi 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 ksl. The factor F (air) from Figure AS-1 of WRC-175 Is 1.4 where:

a= l(2+ t=2 )"2 2.36 Inches t = thickness of nozzle = 7.125 inches t = thickness of vessel = 6.1875 Inches r = apparent radius of nozzle = ri + 0.29 r= 7.09 Inches r = actual inner radius of nozzle = 6.0 inches r = nozzle radius (nozzle comer radius) = 3.75 Inches Thus, aern = 2.36 / 7.09 = 0.33. The value FJr,3, taken from Figure A5-1 of WRC Bulletin 175 for an ar, of 0.33, Is 1.4. Including the safety factor of 1.5, the stress intensity factor. K, is 1.5 a (a)'2, F(arj:

Nominal 1.5- 34.97 - (z - 2.36)"2 1.4 = 200 ksi-in1 2 14=

The method to solve for (T - RTwT) for a specific is based on the N, equation of Paragraph A-4200 in ASME Appendix A 1171 for the pressure test condition (T - RTwT) = In [(1 - 33.2) 20.734 / 0.02 GE Nuclear Energy GBE-NE-A22-00100-08-0la-R2 Non-Proprietary Version (T - RTwT) = In [(200 - 33.2) 20.7341/ 0.02 (T - RTwT) = 104.2F

((

The generic pressure test P-T curve was generated by scaling 200 ksi-in'2 by the nominal pressures and calculating the associated (T- RTNf),

D if 11 The highest RTNDT for the feedwater nozzle materials is 400F as described below. The generic pressure test P-T curve is applied to the Duane Arnold feedwater nozzle curve by shifting the P vs. (r - RTNDT) values above to reflect the RTNDT value of 400F.

I GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version B

Second, the P-T curve Is dependent on the 14 value calculated. The Duane Arnold specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and 1 are shown below:

Vessel Radius, R, 92.69 Inches Vessel Thickness, t, 4.469 Inches Vessel Pressure, P. 1563 pslg Pressure stress: a = PR I t = 1563 psig 92.69 inches / (4.469 Inches) = 32,418 psi.

The Dead weight and thennal RFE stress of 2.967 ksi is conservatively added yielding a = 35.4 siW.The factor F (tr.) from Figure AS-1 of WRC-175 s determined where:

e Y(t 2 + 2)112 =1.78 Inches t, = thickness of nozzle = 5.56 Inches t = thickness of vessel = 4.469 Inches r = apparent radius of nozzle = + 0.29 r.=6.0 inches GE Nuclear Energy GS-NE-A22-00100-08-01a-R2 Non-Proprietary Version r, = actual Inner radius of nozzle = 5.375 inches r = nozzle radius (nozzle comer radius) = 2.25 Inches Thus, a/r = 1.78 6.0 = 0.3. The value F(a1r). taken from Figure AS-1 of WRC Bulletin 175 for an a/rn of 0.30, Is 1.5. Including the safebty factor of 1.5, the stress intensity factor, K4, is 1.5 (7w)" F(air):

Nominal 1 = 1.5

35.4 (,* 1.78)11* 1.5 = 188 ksl-in12 4.3.2.14 Core Not CriticalHeatuplCooldown - Non-Beftline Curve B (Using FeedwaterNozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beltine components for fracture toughness analyses because e stress conditions are the most severe experienced In the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.

Stresses were taken from a g Dfinite element analysis done specifically for the purpose of fracture toughness analysis (( D. Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40F feedwater injection, which is equivalent to hot standby, see Figure 4-3.

The non-belline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (VRC)

Bulletin 175 [15].

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version The stress intensity factor for a nozzle flaw under primary stress conditions (p) Is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

KIp = SF v-(ra) F(a/r,) (4-4) where SF Is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(alr,) is the shape correction factor.

Finite element analysis of a nozzle corner flaw was performed to determine appropriate values of F(e/r,) for Equation 44. These values are shown In Figure AS1 of WRC Bulletin 175 [151.

GE Nuciear Energy GE-NE-A22-001000-a-R2 Non-Proprietary Version The stresses used In Equation 4-4 were taken from (( fl design stress reports for the feedwater nozzle. The stresses considered are primary membrane, pm, and primary bending, apb. Secondary membrane, a., and secondary bending, a, stresses are included in the total 14 by using ASME Appendix G [6] methods for secondary portion, 1,:

Ni= Min (arm (2/3) °Ub) (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 pnmary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. Kp and 14 are added to obtain the total value of stress Intensity factor, 4. A safety factor of 2.0 Is applied to primary stresses for core not critical heatupfcooldown conditions.

Once K was calculated, the following relationship was used to determine (T - RTNDT).

The method to solve for (T - RTNDT) for a specific is based on the K equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltine components was then used to establish the P-T curves.

(T - RTwT) = In [( -33.2) 20.734]I 0.02 (4-6)

Example of Core Not Critical Heatup/Cooldown Calculation for Feedwater NozzlelUpper Vessel Reglon The non-beltline core not critical heatup/cooldown curve was based on the [ D feedwater nozzle [ Danalysis, where feedwater injection of 40"F into the vessel while at operaUng conditions (551.40F and 1050 pslg) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle comer stresses were obtained from finite element analysis ff E. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 Inches was used In the evaluation.

However, a thickness of 7.5 inches Is not conservative for the pressure stress evaluation. Therefore, tf pressure stress (aim) was adjusted for the actual E E vessel thickness of 6.1875 inches (i.e., a., - 20.49 ksl was revised to 20.49 ksi -

I .-

GE Nuclear Energy Gg-Ng-A22-001 000-a-R2 Non-Proprietary Version 7.5 inchesl6.1875 Inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:

op = 24.84 ksi a,, = 16.19 ksi =

= 45.0 ksi t, = 6.1875 inches opb = 0.22 ksi c. = 19.04 ksi a = 2.36 inches r. = 7.09 inches t = 7.125 Inches In this case the total stress, 60.29 kWl, exceeds the yield stress, a,,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root Is the Inside surface temperature.)

R ci),y - cp. + ((aiotra - ayj / 30)]i (%ow - °r (4-7)

For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for p,. The resulting stresses are:

Opm = 24.84 ksi a,= 9.44 ksi apb 0.13 ksi Jab I1.10 MI The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [8]

was based on the 4a thickness; hence, t = 3.072. The resulting value obtained was:

Mm = 1.85 for I/t52 Ulnae 0.926 for 2<<3.4= 2.845 m= 3.21 for 43.464 The value F(a1rQ), taken from Figure AS-1 of WRC Bulletin 175 for an ar, of 0.33, Is therefore, F (air/) =1.4 GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version KNp is calculated from Equation 4-4:

Kp = 2.0 (24.84 + 0.13) . 2.36)' 2 1.4 Kjp 190.4 ksi-in12 K1, Is calculated from Equation 4-5:

K, =2845 (9.44+2/ 11.10)

K, = 47.9 ksl-ln' 2 The total K Is,therefore, 238.3 ksi-inln.

The total V.is substituted into Equation 4-6 to solve for (T - RTNwo):

(T - RTD) = In [(238.3- 33.2) / 20.734] / 0.02 (T - RTmTD) = 1 SOF The [ Dcurve was generated by scaling the stresses used to determine the K1 ;

this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case tat yielded a K. value of 238 ksi-in"2, the pressure is 1050 psig and the hot reactor vessel temperature Is 5S1.4F. Since e reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by - 40) / (551.4 -40). From 14 the associated (T - RTmT) can be calculated:

Core Not Critical Feedwater Nozzle K and (T - RTp=)

as a Function of Pressure Nominal Pressure Saturation Temp. :R Kr (T RTmyr)

Si() - (F) (ksI-ln") (F) 1563 604 0.23 303 128 1400 588 0.34 283 124 1200 557 0.48 257 119 1050 551 0.58 23 115 GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version 1000 546 0.62 232 113 800 520 0.79 206 106 600 l 489 1.0 181 I 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 1.

The highest non-beltline RTNmT for the feedwater nozzle at Duane Arnold is 40'F as shown in Tables 4-1 and 4-2 and previously discussed. The generic curve is applied to the Duane Arnold upper vessel by shifting the P vs. (T - RTp=) values above to reflect the RT= value of 400 F as discussed in Section 4.3.2.1.3.

11 11 4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beiltine region are determined according to the ASME Code. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.

The stress Intensity factors (), calculated for the beltline region according to ASME Code Appendix G procedures (63, were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 100'F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTwcT values for the P-T limits.

GE Nuclear Energy GE-NE-A2200100-08-0la-Rt2 Non-Proprietary Version The Recirculation Inlet (N2) nozzle is located In the extended beitline region, and is the most limiting material for the vessel. The stress intensity factors (Ke) for the Pressure Test and Core Not Critical HeatupCooldown conditions were calculated in the same manner as the feedwater nozzle (see Sections 4.3.2.1.3 and 4.3.2.1.4, respectively),

except that the RTN= was adjusted to account for the effects of Irradiation In accordance with Section 4.2. For the N2 nozzle, the following dimensions were used to calculate the factor F (air") in accordance with WRC-175:

8= Y(tu2 + t) 1' 2 . =1.69inches t = thickness of nozzle = 5.25 Inches t, = thickness of vessel = 5.4375* inches r = apparent radius of nozzle = r + 0.29 r= 6.4125 inches ri = actual Inner radius of nozzle = 5.6875 inches r, = nozzle radius (nozzle corner radius) = 2.5 Inches

Includes cladding thickness Thus, eir. = 1.89 / 6.4125 - 0.295. The value Fair.), taken from Figure AS-1 of WRC Bulletin 175 for an airs of 0.295, is 1.5. These values are used In the stress ntensity factor equations described InSections 4.3.2.1.3 and 4.3.2.1.4.

4.3.2.2.1 Betluine Region - PressureTest The methods of ASME Code Section VU, Appendix G [6] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (t,O ratio of 15, is treated as a thin-walled cylinder. The maximum stress Is the hoop stress, given as:

PR / t, (4-8)

The stress intensity factor, Kim, is calculated using Paragraph G-2214.1 of the ASME Code.

GE Nuclear Energy GE-NE-A22-001O08Ole0-R2 Non-Proprietary Version The calculated value of KYm for pressure test Is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with Klc, the material fracture toughness. A safety factor of 2.0 Is used for the core not critical and core critical conditions.

The relationship between l,- and temperature relative to reference temperature (T - RTNr) is based on the K, equation of Paragraph A-4200 in ASME Appendix A 1171 for the pressure test condition:

Km SF = K = 20.734 exp[O.02 (T - RTNDT )1+ 33.2 (4-9)

This relationship provides values of pressure versus temperature (from KR and (T-RTmT), respectively).

GE's current practice for the pressure test curve is to add a stress intensity factor, Kt, for a coolant heatup/cooldown rate of 2 0 0F/hr to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatupfcooldown rate of 100Fihr. The 11 calculation for a coolant heatup/cooldown rate of 100Fihr Is described inSection 4.32.2.3 below.

-41 -

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version 4.3.2.2.2 Calculations for the Beltine Region - Pressure Test This sample calculation Is for a pressure test pressure of 1035 psig at 32 EFPY for Me limiting beltline plate material. The following inputs were used in the beitline limit calculation:

Adjusted RTNT = Initial RTNOT + Shift A 10+ 130.5= 140.50F (Based on ART values In Table 4-4)

Vessel Height H = 796.94 inches Bottom of Active Fuel Height B = 201 inches Vessel Radius (to inside of clad) R = 92.69 inches Minimum Vessel Thickness (without clad) t = 4.469 Inches Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1035 psi (H - B) 0.0361 psilinch = P pslg (4-10)

= 1035 + (796.94 - 201) 0.0361 = 1057 psig Pressure stress:

a PRA (4-11)

= 1.057 92.69 /4.469 = 21.9 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on a thickness of 4.469 inches (the minimum thickness without cladding);

hence, tn = 2.114. The resulting value obtained was:

Mm = 1.85 for 4t.2 M, - 0.926 4t for 2c1i3.464 = 1.958 M,, = 321 for 4t>3.464 GE Nuclear Energy GE, NE-A22-0010001a-R2 Non-Proprietary Version The stress intensity factor for the pressure stress is Ktm = Mm a. The stress intensity factor for the thermal stress, K4, Is calculated as described in Section 4.3.2.2.4 except that the value of GN is 206F/hr instead of 100 Flhr.

Equation 4-9 can be rearranged, and 1.5 K, substituted for Klc, to solve for (T - RTNDT).

Using the K, equation of Paragraph A-4200 in ASME Appendix A 117], Ki. = 42.88, and K4,= 1.08 for a 2 0°FJhr coolant heatuplcooldown rate with a vessel thickness, t that Includes cladding:

(T - RTmDT) = In[(1.5 m+ K - 33.2) / 20.7343 10.02 (4-12)

= In[(1.5 42.88 + 1.08 - 33.2) 20.734] / 0.02

= 220F T can be calculated by adding the adjusted RTNDT:

T = 22 + 140.5 = 162.65F for P = 1035 psig For Duane Arnold, the N2 Recirculation Inlet nozzle is the limiting material for the beltiine region for 32 EFPY. The betline pressure test P-T curves provided in Section 5.0 of this report are calculated in the same manner as the Feedwater Nozzle pressure test P-T curves, using the N2-specific geometry, as described In Section 4.3.2.1.3. The Initial RTNDT for the N2 Recirculation Inlet nozzle materials Is 40"F as shown InTable 4-2. The generic pressure test P-T curve Is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T - RTwT) values in Section 4.3.2.1.3 to reflect the ART value of 119.2'F. Similarly, the generic pressure test P-T curve is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T-RTNDT) values in Section 4.3.2.1.3 to reflect the 25 EFPY ART value of 113.60F.

GE Nuclear Energy GE-NE-A22-00100-8-01a-IR2 Non-Proprietary Version 4.3.2.2.3 Beltline Region - Core Not Crtical Heatup/Cooldown The betline curves for core not critical heatup/cooldown conditions are Influenced by pressure stresses and thermal stresses, according to the relationship In ASME Section Xi Appendix G 16]:

c = 20 mKm (413) where Km is primary membrane K due to pressure and Kt 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 intenstty factor Is computed by multiplying the coefficient Ml from Figure G-2214-1 of ASME Appendix G [61 by the through-wal temperature gradient AT,,

given that the temperature gradient has a through-wall shape similar to that shown In Figure G-2214-2 of ASME Appendix G 63. The relationship used to compute the through-wallA T. Is based on one-dimensional heat conduction through an insulated flat plate:

b2T(x,t)/ax2 = I /p ((x,/at) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that T(xt) / t = dT(t) I dt = G, where G is the coolant heatup/cooldown rate, normally 100TF/hr. The differential equation is integrated over x for the following boundary conditions:

GE Nuclear Energy Oti-NE-A22-00100 08-01a-R2 Non-Proprietary Version

1. Vessel inside surface (x = 0) temperature Is the same as coolant temperature, To.

2. Vessel outside surface (x - C) is perfectly Insulated; the thermal gradient dTldx - 0.

The integrated solution results in the following relationship for wall temperature:

T = Gx2 12p - GCx f *P+ To (4-15)

This equation Is normalized to plot (T - To) I AT, versus x / C.

The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G 6]. Therefore, AT, calculated from Equation 4-15 is used with the appropriate Ml of Figure G-2214-1 of ASME Appendix G 6] to compute Kt for heatup and cooldown.

The M relationships were derived in the Welding Research Council (WRC)

Bulletin 175 [15] for infinitely long cracks of 1/4T and 1T. For the flat plate geometry and radial thermal gradient, orientation of the crack Is not important.

For Duane Arnold, the N2 Recirculation Inlet nozzle Is the limiting material for the beltline region for 32 EFPY. The beltline core not critical P-T curves provided in Section 5.0 of this report are calculated In the same manner as the Feedwater Nozzle core not critical P-T curves, using the N2-specific geometry, as described in Section 4.3.2.1.4. The Initial RTmT for the N2 Recirculation Inlet nozzle materials Is 40§F as shown In Table 4-

2. The generic core not critical P-T curve is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T- RTwT) values In Section 4.3.2.1.4 to reflect the ART value of 119.2*F. Similarly, the generic core not critical P-T curve Is applied to the Duane Arnold N2 Nozzle curve by shifting the P vs. (T-RTNN,) values in Section 4.3.2.1.4 to reflect the 25 EFPY ART value of 11 3.6°F.

GE Nuclear Energy GE-NE-A22 001080&0a-R2 Non-Proprietary Version Example of Core Not Critical Bettline Calculation Using N2 Recirculation Inlet Nozzle at 1050 psig and 32 EFPY For the N2 Recirculation Inlet nozzle curve, the primary membrane stresses are scaled using the plant specific N2 nozzle geometry. The secondary thermal stresses for the FW nozzle are conservatively used for the N2 nozzle. These stresses are then adjusted for stresses above yield. From these stresses, 14 can be determined. The stresses are scaled for the various pressures and temperatures, similar to the scaling used for the FW nozzle core not critical curve in Section 4.3.2.1.4. The primary stresses are scaled by the nominal pressures, while the secondary stresses are scaled by the temperature difference of the cold FW nozzle (400F) water Injected into the hot reactor vessel nozzle.

The base case Is a pressure of 1050 psig and reactor vessel temperature Is 551.4*F; this yields a K value of 305.6 ksi-in"7. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tw,,tJ, - 40'F) I (551.4 0F - 400F). From 1A, the associated T-RTwr can be calculated.

FW Nozzle t, = 6.1875 inches N2 Nozzle tv = 525 Inches, however, t = 4.875 inches is conservatively used F(e~rr) = 1.5 The FW nozzle stresses are used for the N2 nozzle; only the primary membrane stress Is scaled for the plant specific vessel thickness, t,. At a pressure of 1050 psig and a temperature of 651.4"F, the stresses are:

op= 24.84 ksl - (6.1875 inches / 4.875 Inches) = 31.53 ksl apb -0.22 ksl aO 16.19 ksi

°.b 19.04 ks1 K is calculated:

t12 = (4a)'2 = (4* 1.89)112 = 2.75 MmE= 0.926 2.75 = 2.546 GE Nuclear Energy GFE-NE-A22-00100-08-01a-R2 Non-Proprietary Version Kp is calculated using Equation (4-4) as shown In Section 4.3.2.1.4:

Kp = 2.0* (31.53 + 0.22) (Tir 1.89)'2 1.5 = 232.1 ki-in'1t KN s calculated using Equation (4-5) as shown In Section 4.3.2.1.4:

lt = 2.4 * (16.19 + 23 19.04) = 73.5 ksl-In1 '

r = KP + 5 KI = 232.1 ksi-in1r2 + 73.5 ksi-in1' = 305.6 ksiIn"'2 T-RTmT Is further calculated:

T-RTwT = In[(305.6- 33.2)120.734] / 0.02 = 128.8°F T can be calculated by adding the N2 nozzle adjusted RTNDT of 119.2eF from Table 4-4:

T = 128.8F + 119.2F = 248*F for P = 1050 psig 4.3.2.2.4 Calculationsfor the Beftline Region Core Not Crical Heatup/Cooldown This sample calculation is for a pressure of 1035 psig for 32 EFPY. The core not critical heatuplcooldown curve at 1035 psig uses the same Km as the pressure test curve, but with a safety factor of 2.0 Instead of 1.5. The Increased safety factor Is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. In addition, there is a Kf term for the thermal stress.

The additional Inputs used to calculate K are:

Coolant heatupcooldown rate, normally 100"F/hr G = 100 F/hr Minimum vessel thickness, Including clad thickness C = 0.372 ft (4.469 inches)

(the maxdmum vessel thickness Is conservatively used)

Thermal diffusivity at 550*F (most conservative value) P = 0.354 ft2/ hr 211 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 / 2 (4-16)

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GE Nuclear Energy GFE-NE-A22-00100-08-01a-R2 Non-Proprietary Version

= 100* (0.372)2/ (2- 0.354) = 9.5OF The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of M (=0.25) can be interpolated from ASME Appendix G, Figure G-2214-2 []. Thus the thermal stress intensity factor, Kt = Ml

AT = 5.42, can be calculated. Km 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 - RTNwT) = Inl((2 Km + Kt) - 33.2) / 20.734 /0.02 (417)

= InR(2 42.88+5.52-33.2)/20.734] /0.02

= 51.4 F T can be calculated by adding the adjusted RTar:

T = 51.4 + 140.5 = 191.9 F for P = 1035 psig 4.3.2.3 CLOSURE FLANGE REGION 10CFR5O Appendix G 8] sets several minimum requirements for pressure and temperature Inaddition to those outlined In the ASME Code, based on the closure flange region RTwr. In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves.

However, some closure flange requirements do Impact the curves, as is true. with Duane Arnold at low pressures.

The approach used for Duane Arnold for the bolt-up temperature was based on a conservative value of (RTwT+ 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 III, Subsection NA, Appendix G included the 60OF adder, and 2) inclusion of the additional 60°F requirement above the RTNDT GE Nuclear Energy GE-NE-A22-00100 084la-R2 Non-Proprietary Version provides the additional assurance that a flaw size between 0.1 and 0.24 Inches is acceptable. As shown in Tables 4-1 and 4-2, the upper vessel shell plate material at 14§F represent the limiting Initial RTNDT for the closure flange region, and the LST of the closure studs is 7F; therefore, the bolt-up temperature value used is 74"F. This conservatism Is appropriate because bolt-up Is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

IOCFR5O Appendix G paragraph IV.A.2 [B] 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 RTwT + 90F) and Curve B temperature no less than (RTwT + 1200F).

For pressures below 20% of pre-service hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at RTwT 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 Duane Arnold Technical Specification, is calculated for a water temperature of 68 0F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68'F limit, further extensive calculations would be required to justify a lower temperature. The 740F 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 [81 do not apply, and there are no limits on the vessel temperatures.

GE Nuclear Energy GE-NE-A22-001OD-08-Ola-R2 Non-Proprietary Version 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF IOCFR50, APPENDIX G Curve C, the core critical operation curve, Is generated from the requirements of IOCFR60 Appendix G [8, Table 1. Table 1 of 81 requires that core critical P-T limits be 40F 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.

Table I of IOCFR50 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 74*F, based on an RTwT of 14*F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTgD=

of the closure region 160°F or the temperature required for the hydrostatic pressure test (Curve A at 1035 psig). Due to the presence of the N2 nozzle discontinuity, the requirement of closure region RTmT. 160°F does not cause a temperature shift in Curve C at 312 pslg.

GE Nulear Energy GE-NE-A22 0010D08-01a-R2 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 [2J:

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 heatp and cooldown temperature rate of 100FJhr 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 [31. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20F/hr or less must be maintained at all times.

The P-T curves apply for both heatup/cooldown and for both the 114T and 314T 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 rradiation effects cause the allowable toughness, K, at 1/4T to be less than that at 314T for a given metal temperature.

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GE Nuclear Energy GE-NE-A2200 D08-Oa-R2 Non-Proprietary Version The fllowing P-T curves were generated for Duane Arnold.

e Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Umits (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, beitline (at 25 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

A composite P-T curve was also generated for the Core Critical condition at 25 and 32 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.

Using the fluence defined in Section 4.2.1.2 and Appendix G, the P-T curves are beltline (N2 Recirculation Inlet nozzle) limited above 240 and 230 psig for curve A for 25 and 32 EFPY, respectively, and above 30 psig for curve B for both 25 and 32 EFPY.

Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves Is presented in Appendix B.

GE Nuclear Energy GE-NE-A220100 Ola-R2 Non-Proprietary Version Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curve Curve Description Numbers for for Presentation of Presentation of

_______ ____ . 0 f P,,f:.' r' r.' . .L i ' '; the P-T Curves the P-T Curves Curve A Bottom Head Limits (CRD Nozzle) Figure 5-1 B-I1 & B-3 Upper Vessel Umits (FW Nozzle) Figure 5-2 B-1 &8-3 Belfine Umits for 25 EFPY Figure 5-3 B-3 Belfine Umits for 32 EFPY Figure 5-4 B-1 Curve B I I Bottom Head Umits (CRD Nozzle) Figure 5-5 B-I & -3

__ Upper Vessel Umits (FW Nozzle) Figure 56 B1- & B-3 Betline Limits for 25 EFPY Figure 5-7 B-3 Betline Umits for 32 EFPY Figure 5-8 B-1 Curve C

________ Composite Curve for 25 EFPY" Figure 5-9 B-4 A. B. & C Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure 5-10 B-2 for 32 EFPY*

Bottom Head and Composite Curve B Figure 5-11 B-2 for 32 EFPY*

Composite Curve C for 32 EFPr.* IFigure 5-12 B-2

The Composite Curve A & B curve Is the more limiting of three limits: IOCFR50 Bolt-up Umrits, Upper Vessel Umits (FW Nozzle), and Beltline Umits. A separate Bottom Head Umits (CRD Nozzle) curve Ls Individually Included on this figure.

The Composite Curve C curve Is the more limiting of four limits: IOCFR50 Bolt-up Limits, Bottom Head Umits (CRD Nozzle), Upper Vessel Umits (FW Nozzle), and Beltline Umits.

GE Nuclear Energy GE-NE-A2200100-08-Ola-R2 Non-Proprietary Version 1400 1300 1200 1100 I

a 1000 800 I

0 iI I 800 700 600 400 IL 300 200 100 0 . i i . II I I _li_ I I ! i. I 0 25 50 75 100 125 150 175 200 225 250 275 300 MIMMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve Al

[20F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A2200100 01a-R2 Non-Proprietary Version 1400 1300 1200 1100 I-a 1000 lINITIAL RTndt VALUE IS I 40'F FOR UPPER VESSELI HEATUPICOOLDOWN RATE OF COOLANT a 500 M 2F/ R i

00

-lii-~~~~~~~~~~~

I.- - I-300 UPPES VESSEL UMITS (Including 200 Flange and FW Noze Umit) 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUI REACTOR VESSEL METAL TEMPERATURE (IF)

Figure 52: Upper Vessel P-T Curve for Pressure Test Curve Al

[200FIhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A22-00100 Ola-R2 Non-Proprietary Version 1400 1300 INTIAL RTndt VALUE IS 40*F FOR BELTLINE (N2 NOZZLE) 1200 1100

- 1000 p ELTLINE CURVE

& 900 ADJUSTED AS SHOWN:

EFPY SHIFT rF) 25 73.6 i00 p 700 HEITUPIOO LDOWN RATE OF COOLANT 20 FHR M 1 600 m 500 U0 400 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 25 EFPY

[Flhr or less coolant heatup/cooldown]

GE Nudear Energy GE-NE-A2200100 Ola-R2 Non-Proprietary Version 1400 1300 IMTiAL RTndt VALUE IS 40F FOR BELTLINE (N2 NOZZLE) 1200 1100 1000 3 IBEITLINECURVE 0- 900 ADJUSTED AS SHOWN:

EFPY SHIFT (F) i 32 79.2 0600 w 700 HEATUPCOOLDOWN RATE OF COOLANT

_120*FM"R 600 3 500 I 4D 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 UINMMUM REACTOR VESSEL METAL TEMPERATURE rF)

Figure 5-4: Belfine P-T Curve for Pressure Test [Curve Al up to 32 EFPY 200 Flhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A22-001O00l0a-R2 Non-Proprietary Version 1400 1300 1200 1100 I 1000 I-800

~700 6W Q

1 600 I 40D 300 200 100 0 I1 I! 11 -- I- I I1--I I 0 25 50 75 100 125 150 175 200 225 250 275 300 MIMMUM REACTOR VESSEL METAL TEMPERATURE cF)

Figure 5-5: Bottom Head P-T Curve for Core Not Critical Curve B]

11000Flhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A2200100-0-Ola-R2 Non-Proprietary Version 1400 1300 1200 1100

£ i 1000 NITIAL RTnXdtVALUE IS I4rF FOR UPPER VESSEL I J

I1z oo HEATUPICOOLDOWN RATE OF COOLANT o 700 I OOF/HR 1600 1 00 1 400 300 "UPPER VESSEL UMITS (Inducing 200 Flan and FW Nzle Umft) 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE Fn Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B]

[100OF/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A22001OD-08-Ola-R2 Non-Proprietary Version 1400 1300 INITIAL RTndt VLE IS 40*F FOR BELTUNE (N NOZZLE) 1200 1100 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F) 1000 25 73.6 900 I

800 HEATUP/COOLDOWN RATE OF COOLANT 700 1WOFMR I 600 500 w

400 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTORVESSEL METAL TEMPERATURE F)

Figure 5-7: Beltline P-T Curve for Core Not Critical [Curve B] up to 25 EFPY

[100 0F/hr or less coolant heatup/cooldownj GE Nuclear Energy GE-NE-A2200100 Ola-R2 Non-Proprietary Version 1400 1300 INITIAL RTndt VALUE IS 409 F FOR BELTUNE (N2 NOZZLE) 1200 1100 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F)

I 1000 32 79.2 0 900 HEAPXCOOLNT 800 RATE OF COOLANT o 700 s ltFA4R I

, 500 400 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MIIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 5-8: Beitline P-T Curves for Core Not Critical [Curve B] up to 32 EFPY

[100 0F/hr or less coolant heatupcooldown]

GE Nudear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version 1400 INITIAL RTndt VAWES ARE 1300 40F FOR BELTUNE, (N2 NOZZLE) 40F FOR UPPER VESSEL, 1200 AND 4PF FOR BOTTOM HEAD 1100 BELTUN£ CURVE ADJUSTED AS SHOWN:

I 1000 EFPY SHIFT ( F) 2S 73.6 F 900

¢ 800 9 700 I 600 rHETUPICOOLAN PATE OF COOLANT I Soo s 100rFIHR w 400 300 BELTUNE AND NONBELTLINE 200 L MIT8 100

. 0 t

0 25 50 75 100 125 150 175 20 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-9: Composite Core Critical P-T Curves [Curve Cl up to 25 EFPY l100 0Flhr or less coolant heatuplcooldown]

GE Nuclear Energy GE-NE-A22-O100-08-01a-R2 Non-Proprietary Version 1400 INITIAL RTndt VAWES ARE 1300 40F FOR BELTUNE, (N2 NOZZLE) 40F FOR UPPER VESSEL, 1200 AND 49*F FOR BOTTOM HEAD 1100 BELTUNE CURVES ADJUSTED AS SHOWN:

tZ 1000 tFPY SHIFT (F) 32 79.2 s e00

@e t00 6ON 2 700 HEATUP£CLOhN RATE OF COOLANT e j2rF/HR

500 u

e 400

-UPPER VESSEL 300 AND BELTLINE UMITS BOTTOM HEAD -

200 CURVE 1W0 0 4 0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTORVESSELMETALTEMPERATURE rF)

Figure 5-10 Composite Pressure Test P-T Curves [Curve Al up to 32 EFPY

[20 0Flhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-A2200100 Ola-R2 Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 40F FOR BELTUNE, (N2 NOZZLE) 40F FOR UPPER VESSEL, 1200 AND 49'F FOR BOTTOM HEAD 1100 BELTLINE CURVES ADJUSTED AS SHOWN:

1000 EFPY SHIFT (F) 32 79.2 0t i 900 w

0:

o 700 HEATUP/COOLDOWN z (6 RATE OF COOLANT s i000F1HR g 500 lU 400

-- UPPER VESSEL 300 AND BELTUNE UMITS I I BOTTOM HEAD .

200 CURVE 100 0

0 25 50 75 100 125 10 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE rn :

Figure 5-11: Composite Core Not Critical P-T Curves [Curve BJ up to 32 EFPY

[100°FAir or less coolant heatWcooldown]

GE Nuclear Energy GE-NE-A22-00100-O0-Ola-R2 NonProprietary Version 1400 INITIAL RTndt VALUES ARE 1300 40-F FOR BELTUNE, (N2 NOZZLE) 40 F FOR UPPER VESSEL, 1200 AND I 49'F FOR BOTTOM HEAD 1100 BELTUNE CURVE ADJUSTED AS SHOWN:

0 1000 EFPY SHIFT fF) 32 79.2 s 900 i

I 800 o 700 z

HEATUPCOOLDON t RATE OF COOLANT 1400 s 1006FIHR 400 300

-BELTLINE AND NON-BELTUNE 200 UIMiTS 100 0

0 25 50 75 100 125 150 175 20 225 250275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-12: Composite Core Critical P-T Curves [Curve C up to 32 EFPY

[100 0F/hr or less coolant heatup/cooldownj GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version

6.0 REFERENCES

1. B. D. Frew, "Pressure-Temperature Curves for Duane Arnold Energy Center,'

GE-NE, San Jose, CA, September 2000, (GE-NE-A22-00100-08-01 Revision 0) (GE Proprietary Information).

2. GE Drawing Number 729E762, Reactor Thermal Cycles - Reactor Vessel,'

GE-NE, San Jose, CA, Revision 0. Duane Arnold RPV Thermal Cycle Diagram.

DAEC MDL Document Number APED-A41-003 Revision 0 (GE Proprietary Information).

3. GE Drawing Number 135B99, Nozzle Thermal Cycles - Reactor Vessel,' GE-APED, San Jose, CA, Shi Revision I Sh 2-8 Revision 0. Duane Arnold Nozzle Thermal Cycle Diagram. DAEC MDL Document Number APED-1B11-003,

<1> Revision <2>-<8> Revision 0 (GE Proprietary Information).

4. nAltemative Reference Fracture Toughness for Development of P-T Umit Curves Section Xi, Division 1 Code Case N-640 of the ASME Boiler & Pressure Vessel Code, Approval Date February 26, 1999.

6. L J. Tilly, Duane Arnold RPV Surveillance Materials Testing and Analysis' GE-NE, San Jose, CA, July 1997, (GE-NE-B1 100716-01, Revision 0).

6. 'Fracture Toughness Criteria for Protection Against Failure,' Appendix G to Section III or YJ of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

7. 'Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

8. "Fracture Toughness Requirements,' Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

9. Hodge, J. M., Properties of Heavy Section Nuclear Reactor Steels," Welding Research Council Bulletin 217, July 1976.

GE Nuclear Energy GE-NE-A22 00100OB001a-R2 Non-Proprietary Version

10. GE Nuclear Energy, NEDC-32399-P, Basis for GE RTm Estimation Method,"

Report for BWR Owners' Group, San Jose, California, September 1994 (GE Proprietary Information).

11. Letter from B. Sheron to R.A. PinellioSafety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994, USNRC, December 16, 1994.

12. OA Records RPV CMTR's: Duane Arnold - QA Records & RPV CMTR's Duane Arnold GE PO# 205-H1289, Mfg by CBI), General Electric Company Atomic Power Equipment Department (APED) Quality Control - Procured Equipment RPV QC", Project Duane Arnold, Purchase Order 205-11289, Vendor. Chicago Brdge

& Iron Co, Location: Birmingham, Alabama.

13. Not Used

14. Letter, S.A. Richards, USNRC to J.F. 10approth, GE-NE, Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)", MFN 01-050, September 14,2001.

15. ."PVRC Recommendations on Toughness Requirements for Ferritic Materials,"

Wetding Research Council Bulletin 175, August 1972.

16. f

17. "Analysis of Flaws," Appendix A to Section XJ of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

18. [1 D1 GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version

19. Bottom Head and Feedwater Nozzle Dimensions:

a) 'General Plan 183 BWR Nuclear Reactor Vessel for Iowa Electic Ught and Power', CBI Nuclear Company (GE VPF 2655-18-10). DAEC MDL Document Number APED-1i1-2655-018, Revision 10.

b) Feedwater Nozzle Mark N4 A/D," CBI Nuclear Company (GE VPF 2655-99-7).

DAEC MDL Document Number APED-811-2655-099, Revision 8.

20.((

21. "Materials - Properties," Part D to Section 11of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

GE Nuclear Energy GE-NE-A2200100-08-Ola-R2 Non-Proprietary Version APPENDIX A Description of Discontinulties A-1

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version I. I t 9.

9. 9.

t I-

9. I-

4. 9.

4. I-D A-2

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version 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-ductle failure is not required for portions of nozzles and appurtenances having a thickness of 2.M or less provided the lowest service temperature is not lower than RTm plus 60 0F. Nozzles and appurtenances made from AHoy 600 (Inconel) do not require fracture toughness analysis. Components that do not require a fracture toughness evaluation are listed below.

Nozzle or Appurtenance Nozzle or Material Reference Remarks Identification MK 48-1-1,2,4 Shroud Support Aloy 600 1 2, 3 5 Nozzles or appurtenances made 48-1-3 Attachment to RPV SB-O168 from AJloy 600 (Inconel) require no 48-2 Bottom Head fracture toughness evaluatlon.

9-1 MK 74,75, Insulation Brackets - SA-240 TP 304 L 1,2, 3 & 5 Nozzles or appurtenances made 77-84 Lower-ntermediate from stainless steel require no Shells and Bottom Head fracture toughness evaluation.

MK 101-128 Control Rod Drive Stub Ay 600 2,3 & 5 Nozzles or appurtenances made Tubes - Bottom Head SB-167 from Alloy 600 (Inconel) require no fracture toughness evaluation.

N13, N14 High and Low Pressure . Alloy 600 1, 3 & 5 Not a pressure boundary Seal Leak Detection - SB-166 component therefore, requires no Flange fracture toughness evaluation.

MK 210 Top Head Ufing Lugs 6A-533 GR B CL I 1, 2, 3 & 5 Not a pressure boundary component and loads only occur on Whis component when the reactor Is shutdown during an outage.

Therefore, no fracture toughness evaluation Is required.

- The higlVIOW pressure leak detector, and te seal leak detectcOr are tme same nozzle; these nozzles are the closure flange leak detection nozzles.

A-3

GE Nuclear Energy GE-NE-A22 00100 08 01a-R2 Non-Proprietary Version APPENDIX A

REFERENCES:

1. GE Drawing # 197R608, Revision 9, Reactor Assembly, Nuclear Boiler,"

GE-NE, San Jose, CA. DAEC MDL Document Number APED-B11-084<1>

Revision 9 and APED-B 11-084<2> Revision 8.

2. Certified Stress Report Stress Report, 183, BWR Vessel, Duane Arnold Energy Center, Iowa Electric Lght and Power Co. VPF # 2655-330-1. DAEC MDL Document Number APED-B11-232 Revision 1.

3. QA Records & RPV CMTR's: Duane Arnold - QA Records & RPV CMTR's Duane Arnold GE PO# 205-H1289, Mfg by CBI)" General Electric Company Atomic Power Equipment Department (APED) Quallty Control - Procured Equipment, RPV QC' Project Duane Arnold, Purchase Order 205-H1289, Vendor: Chicago Bridge & Iron Co, Location: Birmingham, Alabama.

4. "General Plan 183 BWR Nuclear Reactor Vessel for Iowa Electrk Light and Power', CBI Nuclear Company (GE VPF 2655-18-10). DAEC MDL Document Number APED-81 1-2655-018 Revision 10.

5. Chicago Bridge & Iron Co, Vessel & Attachment Mat'l Identificationr, (GE-NE VPF# 2655-322(1)-l).

6.. Chicago Bridge and Iron Co., alnstumentation Nozzles Mark N11 A/B and N16 A/B, (GE-NE VPF 2655-109-6). DAEC MDL Document Number APED B11-2655-109, Revision 5.

7. Chicago Bridge & Iron Co, Recirculation Inlet Nozzle MK N2 A/H',

(GE-NE VPF# 2655-96-).

A-4

GE Nudear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version APPENDIX B Pressure-Temperature Curve Data Tabulation B-1

GE Nuclear Energy Gt-NE-A22-00100-O01a-R2 Non-Proprietary Version TABLE B-1. Duane Arnold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Fihr for Curves B & Cand 20 F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 54 REA 0 68.0 74.0 74.0 68.0 74.0 74.0 10 68.0 74.0 74.0 68.0 74.0 74.0 20 68.0 74.0 74.0 68.0 74.0 74.0 30 68.0 74.0 74.0 68.0 74.0 74.0 40 68.0 74.0 74.0 68.0 74.0 74.3 20 68.0 74.0 74.0 68.0 74.0 102.8 60 68.0 74.0 74.0 68.0 74.0 113.6 70 68.0 74.0 74.0 68.0 74.0 122.2 50 68.0 74.0 74.0 68.0 74.0 129.2 90 68.0 74.0 74.0 68.0 74.0 135.1 100 68.0 74.0 74.0 68.0 74.0 140.3 110 68.0 74.0 74.0 68.0 74.0 144.9 120 68.0 74.0 74.0 68.0 74.0 149.2 130 68.0 74.0 74.0 68.0 74.2 153.1 140 68.0 74.0 74.0 68.0 7.4 146.7 120 68.0 74.0 74.0 68.0 70.2 149.2 160 68.0 74.0 74.0 68.0 72.9 162.8 170 68.0 74.0 74.0 68.0 75.4 165.7 150 68.0 74.0 74.0 68.0 87.9 168.3 190 68.0 74.0 74.0 68.0 92.2 170.8 200 68.0 74.0 74.0 68.0 92.3 173.1 210 68.0 74.0 74.0 68.0 94.3 175.3 220 68.0 74.0 74.0 68.0 96.3 177.5 230 68.0 74.0 74.0 68.0 98.1 179.5 B-2

GE Nuclear Energy GE-NE-A22-00100 0&0la-R2 Non-Proprietary Version TABLE B-1. Duane Amold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Fihr for Curves B & C and 20 Fihr for Curve A For Figures 5-1, 5-2, 5-4, 55, 5-6, & 54 240 68.0 74.0 78.3 68.0 99.9 181.4 250 68.0 74.0 84.1 68.0 101.6 183.3 260 68.0 74.0 89.2 68.0 103.2 185.0 270 68.0 74.0 93.9 68.0 104.8 186.7 280 68.0 74.0 982 68.0 106.3 188.4 290 68.0 74.0 102.1 68.0 107.8 190.0 300 68.0 74.0 105.8 68.0 109.2 191.5 310 68.0 74.0 109.2 68.0 110.5 192.9 3125 68.0 74.0 110.0 68.0 110.9 1933 312.5 68.0 104.0 110.0 68.0 134.0 193.3 320 68.0 104.0 112.4 68.0 134.0 194.4 330 68.0 104.0 115.4 68.0 134.0 195.8 340 68.0 104.0 118.2 68.0 134.0 197.1 350 68.0 104.0 120.8 68.0 134.0 198.4 360 68.0 104.0 123.4 68.0 134.0 199.7 370 68.0 104.0 125.8 68.0 134.0 200.9 380 68.0 104.0 128.1 68.0 134.0 202.1 390 68.0 104.0 130.3 68.0 134.0 203.2 400 68.0 104.0 132.4 68.0 134.0 204.4 410 68.0 104.0 134.4 68.0 134.0 205.5 420 68.0 104.0 136.4 68.0 134.0 206.6 430 68.0 104.0 138.3 68.0 134.0 207.6 440 68.0 104.0 140.1 68.0 134.0 208.7 450 68.0 104.0 141.8 68.0 134.0 209.7 460 68.0 104.0 143.5 68.0 134.0 210.6 470 68.0 104.0 145.1 68.0 134.0 211.6 B-3

GE Nuclear Energy GE-NE -A22-00100-08 Ola-R2 Non-Proprietary Version TABLE S-1. Duane Arnold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A For Figures 5-1,-2,5-4. -5,5-6, & 8 480 68.0 104.0 146.7 68.0 134.0 212.6 490 68.0 104.0 148.2 -68.0 134.0 213.5 500 68.0 104.0 149.7 68.0 134.0 214.4 510 68.0 104.0 151.1 68.0 134.0 215.3 520 68.0 104.0 152.6 68.2 134.0 216.2 530 68.0 104.0 153.9 70.2 134.0 217.0 540 68.0 104.0 155.2 72.1 134.0 217.9 550 68.0 104.0 156.5 73.9 134.6 218.7 560 68.0 104.0 157.8 75.7 135.4 219.5 570 68.0 104.0 159.0 77.4 136.1 220.3 580 68.0 104.0 160.2 79.0 136.9 221.1 590 68.0 104.0 161.4 80.6 137.6 221.8 600 68.0 104.0 162.5 82.2 138.1 222.6 610 68.0 104.0 163.7 83.7 138.6 223.3 620 68.0 104.0 164.7 85.1 139.0 224.1 630 68.0 104.0 165.8 86.5 139.4 224.8 640 68.0 104.0 166.9 87.9 139.8 225.5 650 68.0 104.0 167.9 89.2 140.2 226.2 660 68.0 104.0 168.9 90.5 140.7 226.9 670 68.0 104.0 169.9 91.8 141.1 227.6 680 68.0 104.0 170.8 93.1 141.5 228.2 690 68.0 104.0 171.8 94.3 141.9 228.9 700 69.2 104.0 172.7 95.4 142.3 229.6 710 70.7 104.0 173.6 96.6 142.7 230.2 720 72.1 104.0 174.5 97.7 143.1 230.8 730 73.5 104.0 175.4 98.8 143.5 231.4 B-4

GE Nuclear Energy GrE-NE-A2 00100-08-0la-R2 Non-Proprietary Version TABLE B-1. Duane Arnold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A For Figures S-1, 6-2, 5-4, 5-5,5, & 54 740 748 140 1763 99.9 143.9 232.1 750 76.1 104.0 177.1 101.0 144.2 232.7 760 77.4 104.8 178.0 102.0 144.6 233.3 770 78.6 105.6 178.8 103.0 145.0 233.9 780 79.8 106.3 179.6 104.0 145.4 234.5 790 81.0 107.1 180.4 105.0 145.8 235.0 800 82.2 107.9 181.2 105.9 146.1 235.6 810 83.3 108.6 182.0 106.9 146.5 2362 820 84.4 109.4 182.7 107.8 146.9 236.7 830 85.5 110.1 183.5 108.7 147.2 237.3 840 86.5 110.8 184.2 109.6 147.6 237.8 850 87.6 111.5 184.9 1.OA 147.9 238.4 860 88.6 112.2 185.7 111.3 148.3 238.9 870 89.6 112.9 186.4 112.1 148.6 239.4 880 90.5 113.6 187.1 113.0 149.0 239.9 890 91.5 1143 187.8 113.8 149.3 240.5 900 92.4 114.9 188.4 114.6 149.7 241.0 910 93.4 115.6 189.1 115.4 150.0 241.5 920 94.3 116.2 189.8 116.1 150.4 242.0 930 95.1 116.9 190.4 116.9 150.7 242.5 940 96.0 117.5 191.1 117.7 151.0 242.9 950 96.9 118.1 191.7 118A 151.4 243.4 960 97.7 118.7 192.3 119.1 151.7 243.9 970 98.6 119.3 192.9 119.9 152.0 244.4 980 99.4 119.9 193.6 120.6 152.4 244.8 990 100.2 120.5 194.2 121.3 152.7 245.3 B-5

GE Nuclear Energy GE--NE-A22-00100-08-0la-R2 Non-Proprietary Version TABLE B-1. Duane Arnold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Fhr for Curves B& Cand 20 Fhr for Curve A For Figures 5-1, 5-2,6-4, 5 6, & 5-1000 101.0 121.1 194.8 122.0 153.0 245.8 1010 101.7 121.7 195.3 122.6 153.3 246.2 1020 102.5 122.2 195.9 123.3 153.6 246.7 1030 103.3 122.8 196.5 124.0 154.0 247.1 1040 104.0 123.4 197.1 124.6 154.3 247.5 1050 104.7 123.9 197.6 125.3 154.6 248.0 1060 105.4 124.5 198.2 125.9 154.9 248.4 1070 106.2 125.0 198.7 126.5 155.2 248.8 1080 106.9 125.5 199.3 127.2 155.5 249.3 1090 107.6 126.1 199.8 127.8 155.8 249.7 1100 108.2 126.6 200.4 128.4 156.1 250.1 1110 108.9 127.1 200.9 129.0 156.4 250.5 1120 109.6 127.6 201.4 129.6 156.7 250.9 1130 110.2 128.1 201.9 130.2 157.0 251.3 1140 110.9 128.6 202.4 130.7 157.3 251.7 1150 1115 129.1 202.9 131.3 157.6 252.1 1160 112.1 129.6 203.5 131.9 157.9 252.5 1170 112.8 130.1 203.9 132.4 158.2 252.9 1180 113.4 130.6 204.4 133.0 158.5 253.3 1190 114.0 131.1 204.9 133.5 158.7 253.6 1200 114.6 131.5 205.4 134.1 159.0 253.9 1210 115.2 132.0 205.9 134.6 159.3 254.1 1220 115.8 132.5 206.4 135.2 159.6 254.4 1230 116.3 132.9 206.8 135.7 139.9 254.7 1240 116.9 133.4 207.3 136.2 160.2 254.9 1250 117.5 133.8 207.8 136.7 160.4 255.2 B-6

GE Nuclear Energy GE-NE-A2200100-08-0la-R2 Non-Proprietary Version TABLE 8-1. Duane Amold P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Fihr for Curves B& Cand 20 Fihr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, -6, & 5-8 1260 118.0 134.3 208.2 137.2 160.7 255.4 1270 118.6 134.7 208.7 137.7 161.0 255.7 1280 119.1 135.2 209.1 138.2 161.2 255.9 1290 119.7 135.6 209.6 138.7 161.5 256.2 1300 120.2 136.0 210.0 139.2 161.8 256.4 1310 120.7 136.5 210.4 139.7 162.1 256.7 1320 121.3 136.9 210.9 140.2 162.3 256.9 1330 121.8 137.3 211.3 140.6 162.6 257.2 1340 122.3 137.7 211.7 141.1 162.8 257.4 1350 122.8 138.1 212.1 141.6 163.1 257.6 1360 123.3 138.6 212.6 142.0 163.4 257.9 1370 123.8 139.0 213,0 142.5 163.6 258.1 1380 124.3 139.4 213.4 142.9 163.9 258.4 1390 124.8 139.8 213.8 143.4 164.1 258.6 1400 125.3 140.2 214.2 143.8 164.4 258.8 B-7

GE Nuclear Energy GE-NE-A22-001000-a-R2 Non-Proprietary Version TABLE 5-2. Duane Arnold ComposIte P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 FIhr for Curves B & C and 20 *F/hr for Curve A For Figures 5-10, 5-11 and 5-12 0 68.0 74.0 68.0 74.0 74.0 10 68.0 74.0 68.0 74.0 74.0 20 680 74.0 68.0 74.0 74.0 30 68.0 74.0 68.0 74.0 105.9 40 68.0 74.0 68.0 88.3 128.3 50 68.0 74.0 68.0 102.8 142.8.

60 68.0 74.0 68.0 113.6 153.6 70 68.0 74.0 68.0 122.2 162.2 80 68.0 74.0 68.0 129.2 169.2 90 68.0 74.0 68.0 135.1 175.1 100 68.0 74.0 68.0 140.3 180.3 110 68.0 74.0 68.0 144.9 184.9 120 68.0 74.0 68.0 149.2 189.2 130 68.0 74.0 68.0 153.1 193.1 140 68.0 74.0 68.0 156.7 196.7 150 68.0 74.0 68.0 159.8 199.8 160 68.0 74.0 68.0 162.8 202.8 170 68.0 74.0 68.0 165.7 205.7 180 68.0 74.0 68.0 168.3 208.3 190 68.0 74.0 68.0 170.8 210.8 200 68.0 74.0 68.0 173.1 213.1 D-8

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version TABLE B-2. Duane Arnold Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 20 Flhr for Curve A For Figures 5-10. -11 and 5-12 r-~UMUWV 91P F~RV IN t~~~$~ eLLNA HK'BLTIE&QW1N 210 68.0 74.0 68.0 175.3 217.3 220 68.0 74.0 68.0 177.5 217.5 230 68.0 74.0 68.0 179.5 219.5 240 68.0 78.3 68.0 181.4 221.4 250 68.0 84.1 68.0 183.3 223.3 260 68.0 89.2 68.0 185.0 225.0 270 68.0 93.9 68.0 186.7 226.7 280 68.0 98.2 68.0 188.4 228.4 290 68.0 102.1 68.0 190.0 230.0 300 68.0 105.8 68.0 191.5 231.5 310 68.0 109.2 68.0 192.9 232.9 312.5 68.0 110.0 68.0 193.3 233.3 312.5 68.0 110.0 68.0 193.3 233.3 320 68.0 112.4 68.0 194.4 234.4 330 68.0 115.4 68.0 195.8 235.8 340 68.0 118.2 68.0 197.1 237.1 350 68.0 120.8 68.0 198.4 238.4 360 68.0 123.4 68.0 199.7 239.7 370 68.0 125.8 68.0 200.9 240.9 380 68.0 128.1 68.0 202.1 242.1 390 68.0 130.3 68.0 203.2 243.2 400 68.0 132.4 68.0 204.4 244.4 B-9

GS Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version TABLE B-2. Duane Arnold Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 F/hrfor Curves B & C and 20 PF/hr for Curve A For Figures 5-10,-11 and 5-12 410 68.0 134.4 68.0 205.5 245.5 420 68.0 136.4 68.0 206.6 246.6 430 68.0 138.3 68.0 207.6 247.6 440 68.0 140.1 68.0 208.7 248.7 450 68.0 141.8 68.0 209.7 249.7 460 68.0 143.5 68.0 210.6 250.6 470 68.0 145.1 68.0 211.6 251.6 480 68.0 146.7 68.0 212.6 252.6 490 68.0 148.2 68.0 213.5 253.5 500 68.0 149.7 68.0 214.4 254.4 510 68.0 151.1 68.0 215.3 255.3 520 68.0 152.6 68.2 216.2 256.2 530 68.0 153.9 70.2 217.0 257.0 540 68.0 155.2 72.1 217.9 257.9 550 68.0 156.5 73.9 218.7 258.7 560 68.0 137.8 75.7 219.5 259.5 570 68.0 159.0 77.4 220.3 260.3 580 68.0 160.2 79.0 221.1 261.1 590 68.0 161.4 80.6 221.8 261.8 600 68.0 162.5 82.2 222.6 262.6 610 68.0 163.7 83.7 223.3 263.3 620 68.0 164.7 85.1 224.1 264.1 B-10

GE Nuclear Energy GE-NE-A22-0010-08-O01a-R2 Non-Proprietary Version TABLE B-2. Duane Arnold Composte P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Fhr for Curves B & Cand 20 Fihr for Curve A For Figures 5-10. -11 and 5-12 630 68.0 165.8 86.5 224.8 264.

640 68.0 166.9 87.9 225.5 265.5 650 68.0 167.9 89.2 226.2 266.2 660 68.0 168.9 90.5 226.9 266.9 670 68.0 169.9 91.8 227.6 267.6 680 68.0 170.8 93.1 228.2 268.2 690 68.0 171.8 94.3 228.9 268.9 700 69.2 172.7 95.4 229.6 269.6 710 70.7 173.6 96.6 230.2 270.2 720 72.1 174.5 97.7 230.8 270.8 730 73.5 175.4 98.8 231.4 271.4 740 74.8 176.3 99.9 232.1 272.1 750 76.1 177.1 101.0 232.7 272.7 760 77.4 178.0 102.0 233.3 273.3 770 78.6 178.8 103.0 233.9 273.9 780 79.8 179.6 104.0 234.5 274.5 790 81.0 180.4 105.0 235.0 275.0 800 82.2 181.2 105.9 235.6 275.6 810 83.3 182.0 106.9 236.2 276.2 820 84.4 182.7 107.8 236.7 276.7 830 85.5 183.5 108.7 237.3 277.3 840 86.5 184.2 109.6 237.8 277.8 B-11

GE Nuclear Enery GFE-NE-A22-00100-08-0la-R2 Non-Proprietary Version TABLE B-2. Duane Arnold Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 F/hr for Curves B &C and 20 'F/hr for Curve A For Figures 6-10, -11 and 5-12

~

N<Q ~ ~ 3 EdP', :Y: - . 32 EFP F PRSSRECURVA.CRV CURV5E B CUV B CUV 850 87.6 184.9 110.4 238.4 278.4 860 88.6 185.7 111.3 238.9 278.9 870 89.6 186.4 112.1 239.4 279.4 880 90.5 187.1 113.0 239.9 279.9 890 91.5 187.8 113.8 240.5 280.5 900 92.4 188.4 114.6 241.0 281.0 910 93.4 189.1 115.4 241.5 281.5 920 94.3 189.8 116.1 242.0 282.0 930 95.1 190.4 116.9 242.5 282.5 940 96.0 191.1 117.7 242.9 282.9 950 96.9 191.7 118.4 243.4 283.4 960 97.7 192.3 119.1 243.9 283.9 970 98.6 192.9 119.9 244.4 284.4 980 99.4 193.6 120.6 244.8 284.8 990 100.2 194.2 121.3 245.3 285.3 1000 101.0 194.8 122.0 245.8 285.8 1010 101.7 195.3 122.6 246.2 286.2 1020 102.5 195.9 123.3 246.7 286.7 1030 103.3 196.5 124.0 247.1 287.1 1040 104.0 197.1 124.6 247.5 287.5 1050 104.7 197.6 125.3 248.0 288.0 1060 105.4 198.2 125.9 248.4 288.4 B-12

GE Nud~ear Energy Gt-NE--A22-00100-08-01a-R2 Non-Proprietary Version TABLE 6-2. Duane Arnold Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 Flhr for Curves B & Cand 20 'Fihr for Curve A For Figures 6-10. 5-1I and 6-12 1070 106.2 198.7 126.5 2488 288.8 1080 106.9 199.3 1272 249.3 289.3 1090 107.6 199.8 127.8 249.7 289.7 1100 108.2 200.4 128.4 250.1 290.1 1110 108.9 200.9 129.0 250.5 290.5 1120 109.6 201.4 129.6 250.9 290.9 1130 110.2 201.9 130.2 251.3 291.3 1140 110.9 202.4 130.7 251.7 291.7 1150 111.5 202.9 131.3 252.1 292.1 1160 112.1 203.5 131.9 252.5 292.5 1170 112.8 203.9 132.4 232.9 292.9 1180 113.4 204.4 133.0 253.3 293.3 1190 114.0 204.9 133.5 253.6 293.6 1200 114.6 205.4 134.1 233.9 293.9 1210 113.2 205.9 134.6 254.1 294.1 1220 115.8 206.4 135.2 254.4 294.4 1230 116.3 206.8 135.7 254.7 294.7 1240 116.9 207.3 136.2 254.9 294.9 1250 117.5 207.8 136.7 255.2 295.2 1260 118.0 208.2 137.2 255.4 295.4 1270 118.6 208.7 137.7 255.7 295.7 1280 119.1 209,1 138.2 255.9 295.9

GE Nuclear Energy GE-NE-A2240010008g01a-IR2 Non-Proprietary Version TABLE -2. Duane Arnold Composite P-T Curve Values or 32 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A ForFlguresS-10,-I11 and 5-12 1290 119.7 209.6 138.7 256.2 296.2 1300 120.2 210.0 139.2 256.4 296.4 1310 120.7 210.4 139.7 256.7 296.7 1320 121.3 210.9 140.2 256.9 296.9 1330 121.8 211.3 140.6 257.2 297.2 1340 122.3 211.7 141.1 257.4 297.4 1350 122.8 212.1 141.6 257.6 297.6 1360 123.3 212.6 142.0 257.9 297.9 1370 123.8 213.0 142.5 258.1 298.1 1380 124.3 213.4 142.9 258.4 298.4 1390 124.8 213.8 143.4 258.6 298.6 1400 125.3 214.2 143.8 258.8 298.8 B-14

GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version TABLE B-3. Duane Arnold P-T Curve Values for25 EFPY Required Coolant Temperatures at 100 7Flhr for Curves B & C and 20 F/hr for Curve A For Figures 5-1, 5-2,6-3, 5-5, 5-6, & 5-7 0 68.0 74.0 74.0 68.0 74.0 74.0 10 68.0 74.0 74.0 68.0 74.0 74.0 20 68.0 74.0 74.0 68.0 74.0 74.0 30 68.0 74.0 74.0 68.0 74.0 74.0 40 68.0 74.0 74.0 68.0 74.0 92.7 50 68.0 74.0 74.0 68.0 74.0 97.2 60 68.0 74.0 74.0 68.0 74.0 108.0 70 68.0 74.0 74.0 68.0 74.0 116.6 s0 68.0 74.0 74.0 68.0 74.0 123.6 90 68.0 74.0 74.0 68.0 74.0 129.5 100 68.0 74.0 74.0 68.0 74.0 134.7 1H0 68.0 74.0 74.0 68.0 74.0 139.3 120 68.0 74.0 74.0 68.0 74.0 143.6 130 68.0 74.0 74.0 68.0 74.2 147.5 140 68.0 74.0 74.0 68.0 77.4 151.1 150 68.0 74.0 74.0 68.0 80.2 154.2 160 68.0 74.0 74.0 68.0 82.9 157.2 170 6t.0 74.0. 74.0 68.0 85.5 160.1 180 68.0 74.0 74.0 68.0 87.9 162.7 190 68.0 74.0 74.0 68.0 90.2 165.2 200 68.0 74.0 74.0 68.0 92.3 167.5 210 68.0 74.0 74.0 68.0 94.3 169.7 220 68.0 74.0 74.0 68.0 96.3 171.9 230 68.0 74.0 74.0 68.0 98.1 173.9 B-15

GE Nuclear Energy GE-NF-422-00100-08-01a-R2 Non-Proprietary Version TABLE B-3. Duane Arnold P-T Curve Values for25 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B &C and 20 F/hr for Curve A For Figures -1, 5-2, 5-3, 5-5,5-6, & 5-7 VSUR EA E EL 240 68.0 74.0 74.0 68.0 99.9 175.8 250 68.0 74.0 78.5 68.0 fM.6 177.7 260 68.0 74.0 83.6 68.0 103.2 179.4 270 68.0 74.0 88.3 68.0 104.8 181.1 280 68.0 74.0 92.6 68.0 106.3 182.8 290 68.0 74.0 96.5 68.0 107.8 184.4 300 68.0 74.0 100.2 68.0 109.2 185.9 310 68.0 74.0 103.6 68.0 110.5 187.3 312.5 68.0 74.0 104.4 68.0 110.9 187.7 312.5 68.0 104.0 104.4 68.0 134.0 187.7 320 68.0 104.0 106.8 68.0 134.0 188.8 330 68.0 104.0 109.8 68.0 134.0 1902 340 68.0 104.0 112.6 68.0 134.0 191.5 350 68.0 104.0 115.2 68.0 134.0 192.8 360 68.0 104.0 117.8 68.0 134.0 194.1 370 68.0 104.0 120.2 68.0 134.0 195.3 380 68.0 104.0 122.5 68.0 134.0 196.5 390 68.0 104.0 124.7 68.0 134.0 197.6 400 68.0 104.0 126.8 68.0 134.0 198.8 410 68.0 104.0 128.8 68.0 134.0 199.9 420 68.0 104.0 130.8 68.0 134.0 201.0 430 68.0 104.0 132.7 68.0 134.0 202.0 440 68.0 104.0 134.5 68.0 134.0 203.1 450 68.0 104.0 136.2 68.0 134.0 204.1 460 68.0 104.0 137.9 68.0 134.0 205.0 470 68.0 104.0 139.5 68.0 134.0 206.0 B-16

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version TABLE B-3. Duane Arnold P-T Curve Vahues for 25 EFPY Required Coolant Temperatures at 100 FIhr for Curves B & C and 20 F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 480 68.0 104.0 141.1 68.0 134.0 207.0 490 68.0 104.0 142.6 68.0 134.0 207.9 500 68.0 104.0 144.1 68.0 134.0 208.8 510 68.0 104.0 145.5 68.0 134.0 209.7 520 68.0 104.0 147.0 68.2 134.0 210.6 530 68.0 104.0 148.3 70.2 134.0 211.4 540 68.0 104.0 149.6 72.1 134.0 212.3 550 68.0 104.0 150.9 73.9 134.6 213.1 560 68.0 104.0 152.2 75.7 135.4 213.9 570 68.0 104.0 153.4 77.4 136.1 214.7 580 68.0 104.0 154.6 79.0 136.9 215.5 590 68.0 104.0 155.8 80.6 137.6 216.2 600 68.0 104.0 156.9 822 138.1 217.0 610 68.0 104.0 158.1 83.7 138.6 217.7 620 68.0 104.0 159.1 RS.1 139.0 218.S 630 68.0 104.0 160.2 86.5 139.4 219.2 640 68.0 104.0 161.3 87.9 139.8 219.9 650 68.0 104.0 162.3 892 140.2 220.6 660 68.0 104.0 163.3 90.5 140.7 221.3 670 68.0 104.0 164.3 91.8 141.1 222.0 680 68.0 104.0 165.2 93.1 141.5 222.6 690 68.0 104.0 166.2 94.3 141.9 223.3 700 69.2 104.0 167.1 95.4 142.3 224.0 710 70.7 104.0 168.0 96.6 142.7 224.6 720 72.1 104.0 168.9 97.7 143.1 225.2 730 73.5 104.0 169.8 98.8 143.5 225.8 B-17

GE Nuclear Energy GE-NE-A22-00100-08 01a-R2 Non-Proprietary Version TABLE B-3. Duane Amold P-T Curve Values for25 EFPY Required Coolant Temperatures at 100 Flhr for Curves 8 & C and 20 F/hr for Curve A For Figures -1,5-2, -3.5-5, , & 5-7 740 74.8 104.0 170.7 99.9 143.9 226.5 750 76.1 104.0 171.5 101.0 1442 227.1 760 77.4 104.8 172.4 102.0 144.6 227.7 770 78.6 105.6 173.2 103.0 145.0 228.3 780 79.8 106.3 174.0 104.0 145.4 228.9 790 81.0 107.1 174.8 105.0 145.8 229.4 800 82.2 107.9 175.6 105.9 146.1 230.0 810 83.3 108.6 176.4 106.9 146.5 230.6 820 84.4 109.4 177.1 107.8 146.9 231.1 830 85.5 110.1 177.9 108.7 147.2 231.7 840 86.5 110.8 178.6 109.6 147.6 232:2 850 87.6 111.5 179.3 110.4 147.9 232.8 860 88.6 112.2 180.1 111.3 148.3 233.3 870 89.6 112.9 180.8 112.1 148.6 233.8 880 90.5 113.6 181.5 113.0 149.0 234.3 890 91.5 114.3 182.2 113.8 149.3 234.9 900 92.4 114.9 182.8 114.6 149.7 235.4 910 93.4 115.6 183.5 115.4 150.0 235.9 920 94.3 116.2 184.2 116.1 150.4 236.4 930 95.1 116.9 184.8 116.9 150.7 236.9 940 96.0 117.5 185.5 117.7 151.0 237.3 950 96.9 118.1 186.1 118.4 151.4 237.8 960 97.7 118.7 186.7 119.1 151.7 238.3 970 98.6 119.3 187.3 119.9 152.0 238.8 980 99.4 119.9 188.0 120.6 152.4 239.2 990 100.2 120.5 188.6 121.3 152.7 239.7 B-18

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Nor-Proprietary Version TABLE B-3. Duane Arnold P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 20 Fihr for Curve A For Fiures 5-1, 5-2, 5-3, 5-5, 5-, & 5-7 1000 101.0 121.1 189.2 . 122.0 153.0 240.2 1010 101.7 121.7 189.7 122.6 153.3 240.6 1020 102.5 122.2 190.3 123.3 153.6 241.1 1030 103.3 122.8 190.9 124.0 134.0 241.5 1040 104.0 123.4 191.5 124.6 154.3 241.9 1050 104.7 123.9 192.0 125.3 154.6 242.4 1060 105.4 124.5 192.6 125.9 154.9 242.8 1070 106.2 125.0 193.1 126.5 155.2 243.2 1080 106.9 125.5 193.7 127.2 155.5 243.7 1090 107.6 126.1 194.2 127.8 155.8 244.1 1100 108.2 126.6 194.8 128.4 156.1 244.5 1110 108.9 127.1 195.3 129.0 156.4 244.9 1120 109.6 127.6 195.8 129.6 156.7 245.3 1130 110.2 128.1 196.3 130.2 157.0 245.7 1140 110.9 128.6 196.8 130.7 157.3 246.1 1150 111.5 129.1 197.3 131.3 157.6 246.5 1160 112.1 129.6 197.9 131.9 157.9 246.9 1170 112.8 130.1 198.3 132.4 158.2 247.3 1180 113.4 130.6 198.8 133.0 158.5 247.7 1190 114.0 131.1 199.3 133.5 158.7 248.0 1200 114.6 131.5 199.8 134.1 159.0 248.3 1210 115.2 132.0 200.3 134.6 159.3 248.5 1220 115.8 132.5 200.8 135.2 159.6 248.8 1230 116.3 132.9 201.2 135.7 159.9 249.1 1240 116.9 133.4 201.7 136.2 160.2 249.3 1250 117.5 133.8 202.2 136.7 160.4 249.6 B-19

GE Nuclear Energy GS-NE-A22-00100-08-01a-R2 Non-Proprietary Version TABLE B-3. Duane Arnold P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 Fhr for Curves B & Cand 20 F/hr for Curve A For Figures 5-1. 5-2, 6-3, 5-5.5-6, & 5-7 b y~HA TUN HA VSEBELTNE 1260 118.0 134.3 202.6 137.2 160.7 249.8

.1270 118.6 134.7 203.1 137.7 161.0 230.1 1280 119.1 135.2 203.5 138.2 161.2 250.3 1290 119.7 135.6 204.0 138.7 161.5 250.6 1300 120.2 136.0 204.4 139.2 161.8 250.8 1310 120.7 136.5 204.8 139.7 162.1 251.1 1320 121.3 136.9 205.3 140.2 162.3 251.3 1330 121.8 137.3 205.7 140.6 162.6 251.6 1340 122.3 137.7 206.1 141.1 162.8 251.8 1350 122.8 138.1 206.5 141.6 163.1 252.0 1360 123.3 138.6 207.0 142.0 163.4 252.3 1370 123.8 139.0 207.4 142.5 163.6 252.5 1380 124.3 139.4 207.8 142.9 163.9 252.8 1390 124.8 139.8 208.2 143.4 164.1 253.0 1400 125.3 140.2 208.6 143.8 164.4 253.2 B-20

GE Nudiear Einergy GE-NE-AC22-00100-0e-Ola-R2 Non-Proprietary Version TABLE 5-4. Duane Amold Power Uprate Composite P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 Flhrfor Curves B &C and 20 F/hrfor Curve A For Figure 5 0 68.0 74.0 68.0 74.0 74.0 10 68.0 74.0 6X.0 74.0 74.0 20 68.0 74.0 68.0 ?4.0 74.0 30 68.0 74.0 68.0 74.0 100.3 40 69.0 74.0 68.0 82.7 122.7 50 69.0 74.0 68.0 97.2 137.2

-60 68.0 74.0 68.0 109.0 148.0 70 68.0 74.0 68.0 116.6 136.6 so 69.0 74.0 68.0 123.6 163.6 90 68.0 74.0 68.0 129.5 169.5 loo 68.0 74.0 68.0 134.7 174.7 110 68.0 74.0 68.0 139.3 179.3 120 68.0 74.0 68.0 143.6 183.6 130 68.0 74.0 68.0 147.5 187.5 140 68.0 74.0 68.0 131.l 191.1 150 68.0 74.0 68.0 1S4.2 194.2 160 68.0 - 74.0- 63.0 157.2 197.2 170 68.0 74.0 68.0 160.1 200Q1 ISO 68.0 74.0 68.0 162.7 202.7 190 68.0 74.0 68.0 165.2 205.2 200 68.0 74.0 68.0 '167.5 207.S 210 69.0 74.0 68.0 169.7 209.7 6-21

GE Nuclear Energy Gt-NE-A22-00100-08 Ola-R2 Non-Proprietary Version TABLE 6-4. Duane Amold PowerUprate ComposKe P-T Curve Values for25 EFPY Required Coolant Temperatures at 100 Foirfor Curves B & C and 20 'FnIr for Curve A For Fgure 59 220 68.0 74.0 68.0 171.9 211.9 230 68.0 74.0 68.0 173.9 213.9 240 68.0 74.0 68.0 175.8 215.8 250 68.0 78.5 68.0 177.7 217.7 260 68.0 - 83.6 68.0 179.4 219.4 270 68.0 88.3 68.0 181.l 221.1 280 68.0 92.6 68.0 182.8 222.8 290 68.0 96.5 68.0 184.4 224.

300 68.0 1002 68.0 185.9 225.9 310 68.0 103.6 68.0 187.3 227.3 312.5 68.0 104.4 68.0 177.7 227.7 312. 68.0 104.4 68.0 187.7 227.7 320 68.0 106.8 68.0 188.8 228.8 330 68.0 109. 68.0 190.2 2302 340 68.0 112.6 68.0 194.5 231.5 350 68.0 100.2 68.0 192.8 232.8 360 68.0 117.8 68.0 194.1 234.1 370 68.0 1202 68.0 195.3 235.3 30 68.0 1223 68.0 196.7 236.7 390 68.0 124.7 68.0 197.6 237.6 400 68.0 126.8 68.0 199.8 232.8 410 68.0 12.8 68.0 199.9 239.9 420 68.0 130.8 68.0 201.0 241.0 430 68.0 132.7 68.0 202.0 242.0 B-22

GE Nuclear Energy GS-NS-A22-00100-08-01a-R2 Non-Proprietary Version TABLE B-4. Duane Anold Power Uprate ComposKe P-T Curve Values for25 EFPY Required Coolant Temperatures at 100 *Fihr for Curves B & C and 20 F/hr for Curve A For Figure 5-9 440 68.0 134.5 68.0 203.1 243.1 450 68.0 136.2 68.0 204.1 244.1 460 68.0 137.9 68.0 205.0 245.0 470 68.0 139.5 68.0 206.0 246.0 480 68.0 141.1 68.0 207.0 247.0 490 68.0 142.6 68.0 207.9 247.9 500 68.0 144.1 68.0 208.8 248.8 510 68.0 145.5 68.0 209.7 249.7 520 68.0 147.0 68.2 210.6 250.6 530 68.0 148.3 70.2 211.4 251.4 540 68.0 149.6 72.1 212.3 252.3 550 68.0 150.9 73.9 213.1 253.1 560 68.0 152.2 75.1 213.9 253.9 570 68.0 153.4 77A 214.7 254.7 580 68.0 154.6 79.0 215.5 255.5 590 68.0 155.8 80.6 216.2 256.2 600 68.0 156.9 82.2 217.0 257.0 610 68.0 158.1 83.7 217.7 257.7 620 68.0 159.1 85.1 218.5 258.3 630 68.0 160.2 86.5 219.2 2592 640 68.0 161.3 87.9 219.9 259.9 650 68.0 162.3 89.2 220.6 260.6 660 68.0 163.3 90.5 221.3 261.3 670 68.0 164.3 91.8 222.0 262.0 8-23

GE Nuclear Energy GPE-NE-A22-0010008-01a-112 Non-Proprietary Version TABLE B-4. Duane Amold Power Uprate ComposKe P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 FIhr for Curves B & C and 20 Fihr for Curve A For Figure 5-9 690 68.0 166.2 94.3 223.3 263.3 700 69.2 167.1 95A4 224.0 264.0 710 70.7 168.0 96.6 224.6 264.6 720 72.1 168.9 97,7 225.2 2652 730 73.5 169.8 98.8 22S.8 265.8 740 74.8 170.7 99.9 226.5 266.3 750 76.1 171.5 101.0 227.1 267.1 760 77.4 172.4 102.0 227?7 267.7 770 78.6 173.2 103.0 228.3 268.3 780 79.8 174.0 104.0 228.9 268.9 790 81.0 174.8 105.0 229.4 269.4 goo 82.2 175.6 105.9 230.0 270.0 810 83.3 176.4 106.9 230.6 270.6 820 84.4 177.1 107.8 231.1 271.1 830 85.5 177.9 108.7 231.7 271.7 840 86.5 178.6 109.6 232.2 272.2 850 87.6 179.3 110.4 232.8 272.8 860 88.6 180.1 111.3 233.3 273.3 870 89.6 180.8 112.1 233.8 273.8 880 90.5 1RI.5 113.0 234.3 -274.3 90 91.0 162.2 113. 234.9 274.9 900 92.4 182.9 114.6 235.4 273.4 910 93.4 193.5 915.4 235.9 275.9 B-24

GE Nuclear Energy GE-NE-A22 00100 08-01a-R2 Non-Proprietary Version TABLE -4. DuaneAmold PowerUprate ComposKe P-TCurve Values for25 EFPY Required Coolant Temperatures at 100 Fhr for Curves B & C and 20 F/hr for Curve A For Figure 5-9 920 94.3 184.2 1161 236.4 276.4 930 95.1 184.8 116.9 236.9 276.9 940 96.0 185.5 117.7 237.3 277.3 950 96.9 186.1 118.4 237.8 277.8 960 97.7 186.7 119.1 238.3 278.3 970 98.6 187.3 119.9 238.8 278.8 980 99.4 188.0 120.6 239.2 279.2 990 100.2 188.6 121.3 239.7 279.7 1000 101.0 189.2 122.0 240.2 280.2 1010 101.7 189.7 122.6 240.6 280.6 1020 102.5 190.3 123.3 241.1 281.1 1030 103.3 190.9 124.0 241.5 281.5 1040 104.0 1915 124.6 241.9 281.9 1050 104.7 192.0 125.3 242.4 282.4 1060 105.4 192.6 125.9 242.8 282.8 1070 106.2 193.1 126.5 243.2 283.2 1080 106.9 193.7 127.2 243.7 283.7 1090 107.6 194.2 127.8 244.1 284.1 1100 108.2 194.8 128.4 244.5 284.5 1110 108.9 195.3 129.0 244.9 284.9 1120 109.6 195.8 129.6 245.3 285.3 1130 110.2 196.3 130.2 245.7 285.7 1140 110.9 196.8 130.7 246.1 286.1 1150 111.5 197.3 131.3 246.5 286.5 B-25

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version TABLE B-4. Duane Amold Power Uprate Composite P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 Fir for Curves B & C and 20 F/hr for Curve A For Figure 5-9 1160 112.1 197.9 131.9 246.9 286.9 1170 112.8 198.3 132.4 247.3 287.3 1180 113.4 198.8 133.0 247.7 287.7 1190 114.0 199.3 133.5 248.0 288.0 1200 114.6 199.8 134.1 248.3 288.3 1210 115.2 200.3 134.6 248.5 28.5 1220 115.8 200.8 13S.2 248.8 288.8 1230 116.3 201.2 135.7 249.1 289.1 1240 116.9 201.7 136.2 249.3 289.3 12S0 117.5 202.2 136.7 249.6 289.6 1260 118.0 202.6 137.2 249.8 289.8 1270 118.6 203.1 137.7 250.1 290.1 1280 119.1 203.5 138.2 250.3 290.3 1290 119.7 204.0 138.7 250.6 290.6 1300 120.2 204.4 139.2 250.8 290.8 1310 120.7 204.8 139.7 251.1 291.1 1320 121.3 205.3 140.2 251.3 291.3 1330 121.8 205.7 140.6 251.6 291.6 1340 122.3 206.1 141.1 2S1.8 29t.8 1350 122.8 206.5 141.6 252.0 292.0 1360 123.3 207.0 142.0 252.3 292.3 1370 123.8 207.4 142.5 252.5 292.5 1380 124.3 207.8 142.9 252.8 292.8 1390 124.8 20O.2 143.4 2S3.0 293.0 B-26

GE Nuclear Energy G;E-N4E-A2200100-08-01a-R2 Non-Proprietary Version TABLE B-4. Duane Arnold Power Uprate Composite P-T Curve Values for 25 EFPY Required Coolant Temperatures at 100 *Fhr for Curves B & C and 20 F/hr for Curve A For Figure 50 BCU E-OOM UPER T UP

t-'. .~"'

<- EAD_ fEi.TlJ;N AT HEIAE) SIfLTUNEA e0-BlTJb, NElr 1400 12S.3 208.6 143.8 253.2 293 2 B-27

GE Nudear Energy GE-NE-A220010 Ola-R2 Non-Proprietary Version APPENDIX C Operating And Temperature Monitoring Requirements C-1

GE Nuciear Energy GE--NE-A22-00100-08-0la-R2 Non-Proprietary Version C.1 NON-BELTIUNE 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 at showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water In the recirculation loops provide good estimates of the beltilne temperature during pressure testing. Thermocouples on the RPV flange to shell Junction outside surface should be used to monitor compliance with upper vessel curve. Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D.

First however, It should be determined whether there are significant temperature differences between the beltline region and the bottom head region.

C.2 DETERMINING WHICH CURVE TO FOLLOW The following subsections outline the criteria needed for determining which curve is governing during different situations. The application of the P-T curves and some of the assumptions inherent In the curves to plant operation is dependent on the proper monitoring of vessel temperatures.

C-2

GE Nuclear Energy G3E-NE-A22-00100 0-0la-R2 Non-Proprietary Version C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at limes when the coolant temperature is changing by 5200F 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 200 F per hour during a hydrotest and when the core is not critical.

C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core Is critical.. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.

C.3 REACTOR OPERATION VERSUS OPERATING UMITS For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the ight of the P-T curves). The operations where P-T curve compliance Is typically monitored closely are planned events, such as vessel boltup, leakage testing and startup/shutdown operations, where operator actions can directly Influence vessel pressures and temperatures.

The most severe unplanned transients relative to the P-T curves are those that result from SCRAMs, wich sometimes Include recirculation pump trips. Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high.

Experience with such events has shown that operator action is necessary to avoid P-T curve exceedance, but there is adequate time for operators to respond.

C-3

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

Head flange boltup

Leakage test (Curve A compliance)

Startup (coolant temperature change of less than or equal to 100F in one hour period heatup)

Shutdown (coolant temperature change of less than or equal to 100OF in one hour period cooldown)

Recirculation pump trip, bottom head stratification (Curve B compliance)

C-4

GE Nuclear Energy GE-NEA22-00100 01a-R2 Non-Proprietary Version APPENDIX D GE SIL 430 D-I

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version

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.

TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)

Measurement Use Limitations Steam dome saturation Primary measurement Must convert saturated temperature as determined above 2120F for Tech steam pressure to from main steam instrument Spec 100-F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 2120F for Tech Must comply with SIL 251 Spec IOOF/hr beatup to avoid vessel stratification.

and cooldown rate.

Alternate measurement When above 2120F need to above 2120 F. allow for temperature variations (up to 10-15°F lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy GE-NE-A2200100 01a-R2 Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Umitations 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).

RHIR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec IOOPF/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta Us will be indicated coolant temperature. Delta T limit is l oF for BWR/6s and 1450 F for earlier BWRs.

Primary measurement to Must have drain line comply with Tech Spec flow. Use to veify brittle fracture compliance with Tech limits during cooldown. Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).

Alternate information Must compensate for outside only measurement for metal temperature lag bottom head inside/ during heatup/cooldown.

outside metal surface Should have drain line flow.

temperatures.

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use limitations Closure head flanges Primary measurement for Use for metal (not coolant) outside surface TICs BWR/6s to comply with temperature. Install Tech Spec brittle fracture temporary T/Cs for metal temperature limit alternate measurement, if for head boltup. required.

One of two primary measure-ments for BWRI6s for hydro test.

RPV flange-to-shell Primary measurement for Use for metal (not coolant) junction outside BWRs earlier than 6s to temperature. Response surface T/Cs comply with Tech Spec faster than closure head brittle fracture metal flange T/Cs.

temperature limit for head boltup.

One of two primary Use RPV closure head flange measurements for BWRs outside surface as alternate earlier than 6s for measurement.

hydro test. Preferred in lieu of closure head flange TICs if available.

RPV shell outside Information only. Slow to respond to RPV surface T/Cs coolant changes. Not available on BWR/6s.

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail-able on BWR/6s.

D-4

GIE Nuclear Energy GE-NE-A22-00100-0-Ola-R2 Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside I of 2 primary measurements Should verify that vessel surface T/Cs to comply with stratification is not Tech Spec brittle fracture present for vessel hydro.

metal temperature (see SIL No. 251).

limit for hydro test.

Primary measurement to Use during heatup to verify comply with Tech Spec compliance with Tech Spec brittle fracture metal metal temperature/reactor temperature limits pressure curves.

during heatup.

Note: RPV vendor specified metal T limits for vessel heatup and cooldown should be checked during initial plant startup tests when initial RPV vessel heatup and cooldown tests are run.

D-5

GE Nuclear Energy GE--NE-A22-00100-08-01a-R2 Non-Proprietary Version Product

Reference:

821 Nuclear Boiler Prepared By: A.C. Tsang Approved for Issue: Issued By:

B.H. Eldridge, Mgr. D.L Allred, Manager Service Infornation Customer Service Infonation and Analysis Notice:

SILs pertain only to GE BWRs. GE prepares SlLs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained inSILs to any plant or facility other than GE BWRs as designed and furnished by GE. Deternination of applicability of information contained In any SIL to a specific GE 8WR and implementation of recommended action are responsibilities of the owner of that GE BWR.SlLs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of Ks 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 SLs. GE assumes no responsibility for liability or damage, which may result from the use of information contained InSILs.

D-6

GE Nudear Energy GE-NS-A22 001000-a-R2 Non-Proprietary Version APPENDIX E Determination of Beltilne Region and Impact on Fracture Toughness E-1

GE Nueclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version IOCFR50, Appendix G defines the beildine region of the reactor vessel as follows:

"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' 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 nm 2 . Therefore, if nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 acme, then fracture toughness evaluation must be conducted and demonstrated on those nozzles which experience a fluence at or exceeding 1.0e17 n/cOm 2 The following dimensions are obtained from the referenced drawings:

Shell # 2 - Top of Active Fuel (TAF): 351' (from vessel 0)

Shell # 1 - Bottom of Active Fuel (BAF): 201" (from vessel 0)

Top of Active Fuel - Extended Beltline: 370 (from vessel 0)

Bottom of Active Fuel - Extended Beeline: 183 (from vessel 0)

Elevation of Top of N2 Nozzle in Shell # 1: 195.8125" (from vessel 0)

Elevation of Centerline of N2 Nozzle In Shell # 1: 182' (from vessel 0)

Elevation of Centerline of Nozzle NIS in Shell #2: 348" (from vessel 0)

Elevation of Centerline of Nozzle N1 In Shell #1: 131" (from vessel 0)

Elevation of Top of N1 Nozzle in Shell #1: 156.5" (from vessel 0)

Based on the axial flux profile, the RPV flux level at -18" below the BAF dropped to less than 0.024 of the peak flux level at the same radius. Ukewise, Me RPV flux level at -19' above the TAF dropped to less than 0.024 of the peak flux at the same radius.

Therefore, since the best estimate peak RPV fluence is 4.17E18 n/a 2 as defined in Section 4.2.12 and Appendix G, fluence at -18" below BAF and -19" above TAF is E-2

GE Nuclear Energy GE-NE-A22-00100-08-0la-R2 Non-Proprietary Version expected to be less than 1.0E17 n/cm2 at 32 EFPY. The beitline region considered in the development of the P-T curves is adjusted to include the additional area above and below the active fuel region. The adjusted beltilne region extends from -183' to -370' above reactor vessel "Or.

As shown above, the N2 Recirculation Inlet and N16 Instrumentation nozzles are within the core beltline region, and are bounded by the beltline curve as stated in Appendix A.

The recirculation inlet nozzle Is closest to the bettline region (the top of the Recirculation Inlet nozzle is -I3 above BAF - Extended Beftline), and the instrumentation nozzle is within the BAF-TAF region of the reactor vessel. Therefore, since it can be shown that no other nozzles are located where the peak neutron fuence is expected to exceed or equal .0E17 n/cm 2 , then It can be concluded that all remaining reactor vessel nozzles are outside the beltline region of the reactor vessel.

Based on the above, it Is concluded that none of the Duane Arnold reactor vessel nozzles, other than the Recirculation Inlet Nozzle and the Instrumentation Nozzle, which are considered in the P-T curve evaluation, are In the betline region.

E-3

GE Nuclear Energy GE-NE-A220100 Ola-R2 Non-Proprietary Version Appendix E

References:

None E-4

GE Nuclear Energy GE-NE-A2200100-08-O1a-R2 Non-Proprietary Version APPENDIX F Evaluation For Upper Shelf Energy Equivalent Margin Analysis (EMA)

F-1

GE Nuclear Energy GE-NE-A22-00100--1a-R2 Non-Proprietary Version Paragraph V.B of IOCFR50 Appendix G [1 sets limits on the upper shelf energy of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Regulatory Guide 199, Rev. 2 [23 methods and WROG Equivalent Margin Analyses (EMA) 3, 4 methods, are summarized InTables F-1 through F-4.

Unirradiated upper shelf data was not available for all of the material heats In the Duane Arnold belfline region. Due to this lack of specific pre-operational USE data, Duane Arnold is evaluated to verify that the BWROG EMA is applicable. The USE decrease prediction values from Reg.

Guide 1.99, Rev. 2 were used for the beltline components as shown in Tables F-1 through F-4.

These calculations are based upon the 32 EFPY peak 1/4T fluence for each component as provided in Table 4-4.

Based on the results presented In Tables F-I through F-4, the USE EMA values for the reactor vessel beeline materials remain within the limits of Reg. Guide 1.99, Rev. 2 and IOCFR50 Appendix G for 32 EFPY of operation.

F-2

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Table F-1 Plate Equivalent Margin Analysis PLANT APPLICABILITY VERIFICATION FORM FOR Duane Arnold - BWR 4/M I - Including Unrated Power Condition BWR/3-6 PLATE Surveillance Plate (Heat B0673-1) USE:

%MCu= 015 Ist Capsule Fluence = 4.9x101 7 n/crm2 2nd Capsule Fluence = 1.1xl0 18 n/cm 2 Unirradiated to st Capsule Measured % Decrease = 25 (Charpy Curves)

Unirradiated to 2nd Capsule Measured % Decrease f1 (Charpy Curves) 1st Capsule Rev 2 Predicted % Decrease 12(Rev 2, Figure 2) 2nd Capsule Rev 2 Predicted % Decrease = 14 (Rev 2, Figure 2)

Limiting Beltline Plate (Heat B0673-1) USE-

%Cu= 0.15 32 EFPY 1/4 T Fluence = 3.19x101 8 n/cm 2 Rev 2 Predicted. % Decrease = 19(Rev 2, Figure 2)

Adjusted % Decrease = N/A (Rev 2, Position 2.2)

I 19 % c 21%, so vessel plates are bounded by equivalent margin analysis I F-3

GE Nuclear Energy GE-NE-A22010001-Oa-R2 Non-Proprietary Version Table F-2 Weld Equivalent Margin Analysis PLANT APPLICABILITY VERIFICATION FORM FOR D-ane Arnold - BWR 4/MK I - Includin! Unrated Power Condition Iw/R2l-6 WELT)

Surveillance Weld (Heat Unknown) USE:

%Cu= 0.02 1st Capsule Fluences 4.9x10 17 n/nc2 2nd Capsule Fluence= 1.1x10 1 8 n/cm2 Unirradiated to 1st Capsule Measured % Decrease O.(Charpy Curves) a Unirradiated to 2nd Capsule Measured % Decrease = (Chaipy Curves) 1st Capsule Rev 2 Predicted % Decrease = 7 (Rev 2, Figue 2) 2nd Capsule Rev 2 Predicted % Decrease - _ (Rev 2, Figure 2)

Uimitin Betline Weld (Heat 432Z0471 Lot B003A27A) USE:

%Cu = 0.03 32 EFPY 1/4 T Fluence = 3.19x1018 n/Jm2 Rev 2 Predicted % Decrease 12.5 (Rev 2, Figure 2)

Adjusted % Decrease = N/A (Rev 2, Position 2.2) 12.5% 534%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Non-Proprietary Version Table F-3 Nozzle Equivalent Margin Analysis PLANT APPLICABILTY VERIFICATION FORM FOR Duane Arnold - BWR 4MK I - Including Unrated Power Condition BMWRS-6 PLATE Surveillance Plate (Heat B0673-1 USE:

%Ocu = 0.1 5 1st Capsule Fluence = 4.9x1017 n/cM2 2nd Capsule Fluence l.1x10 8 n/cm2 Unirradiated to 1st Capsule Measured % Decrease = 2.5 (Charpy Curves)

Unirradiated to 2nd Capsule Measured % Decrease = (Charpy Curves) 1st Capsule Rev 2 Predicted % Decrease = 12 (Rev 2, Figure 2) 2nd Capsule Rev 2 Predicted % Decrease = 14 (Rev 2, Figure 2)

Beltline Nozzle (N16 - Heat 0205VW) USE:

%Cu = 1I 32 EFPY 1/4 T Fluence 8.63x10 17 n/cm 2 Rev 2 Predicted % Decrease = 15.5 (Rev 2, Figure 2)

Adjusted % Decrease NL (Rev 2, Position 2.2) 1.5 % c 21%, so vessel nozzles are bounded by equivalent margin analysis F-5

Kit toiIamr cnar~

Mu r.l=-tJr--A90-nn4nrj-M-t'ilgt-R9 Non-Proprietary Version Table F-4 Nozzle Equivalent Margin Analysis PLANT APPLICABILITY VERIFICATION FORM FOR uane Arnold - BWR 4/MK I - Includina Unrated Power Condition BWR3-6 PLATE Surveillance Plate (Heat B0673-I1) USE:

%Cu-0.15 IstCapsuleFluence= 4.9x1l 17 nfcm 2 2nd Capsule Fluence= I.lxlO1 8 n/cm 2 Unirradiated to Ist Capsule Measured % Decrease - 25 (Charpy Curves)

Unirradiated to 2nd Capsule Measured % Decrease - 13 (Charpy Curves) 1st Capsule Rev 2 Predicted % Decrease 2 (Rev 2, Figure 2) 2nd Capsule Rev 2 Predicted % Decrease = 14 (Rev 2, Figure 2)

LimitinQ Betline Nozzle (N2 - Heat 0206VW) USE:

%Cu - 0.18 32 EFPY 1/4 T Fluence = 5.85x101 7 n/cm2 Rev 2 Predicted % Decreasc = 14.5 (Rev 2, Figure 2)

Adjusted % Decrease = IL& (Rev 2, Position 2.2) 14.5 % c 21%, so vessel nozzles are bounded by equivalent margin analysis F-6

GE Nuclear Energy GE-NE-A22-0100-08-Ola-R2 Non-Proprietary Version Appendix F

References:

1. Fracture Toughness Requirements, Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

2. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

3. H.S. Mehta, TA. Caine, and S.E. Plaxton, 10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels", GENE, San Jose, CA, February 1994 (NEDO-32205-A, Revision 1).

4. LA. England (BWR Owners' Group) to Daniel G. McDonald (US NRC), BWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin Analysis - Approved Version', BWROG-94037, March 21, 1994.

F-7

GE Nudear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version APPENDIX G Fluence Evaluation G-1

GE Nuclear Energy GE-NE-A22-00100-08-01a-R2 Nor-Proprietary Version G.I Overvew and Objective Neutron rradiation of reactor pressure vessel (RPV) causes reduction in material ductility and creates structural embrittlement at higher operating temperatures. The effect is particularly significant when impurities such as nickel, copper, or phosphorus are imbedded in noticeable levels, as commonly true for the RPV steel. Therefore determination of neutron fluence level Is one of the first steps toward RPV fracture toughness evaluations. Neutron fluence is accumulated neutron flux during irradiation time.

G.2.1 Scope Fast neutron flux densities in the beltline region extending from the core through the RPV are calculated in this task. The operating condition assumed for this analysis corresponds to 120% of the ORTP.

The methodology used for the neutron flux calculation is documented in a Licensing Topical Report (LTR) NEDC-32983P-A [11, which was approved by the NRC for licensing applications in the Safety Evaluation Report (SER) [2]. In general, GE's methodology described In the LTR follows the Intent of Regulatory Guide 1.190 [4] for neutron flux evaluation. This methodology is briefly discussed below.

G.2.2 Method of Evaluton The three-dimensional spatial distribution of flux density In the vicinity of the reactor vessel is simulated by combining the results of two separate thwodimensional neutron transport analyses. The first of these is performed in (rie) geometry and provides the radial and azimuthal variation of flux Inthe core to the vessel at an elevation near the core midplane. The second analysis, which is performed in (r,z) geometry, establishes the relative variation of flux density with elevation. The (rxz) analysis is the basis for determining an axial "adjustment" factor to be applied to the core midplane (r, 0) analysis results.

The neutron flux calculations are performed with the two-dimensional discrete ordinates neutral particle transport code DORT [3]. DORT Is distributed as part of the TORT code package by the Radiation Shielding Information Center at Oak Ridge National Laboratory and is te updated version of the DOT series of codes.

G-2

GE Nuclear Energy GE-NE-A2200100-08-Ola-R2 Non-Proprietary Version The DAEC core configuration Is Illustrated In Figure G-2-1. 1[

The cross-section data used Inthe DORT calculation are processed with the nuclear cross-section processing package Inthe GENE Engineering Computation Program (ECP) library. [

ii G.2.2.1 r, 0) Model a

GE Nuclear Energy GE-NE-A22-O OO-08-Ola-R2 Non-Proprietary Version G2.2.2 (rz) Model H~

1 G-4

GE Nuclear Energy GE-NE-A22-0010040-a-R2 Non-Proprietary Version G23 Inputs end Assumptions The AEP rated power was assumed to be 1912 MW#,. [

E The reactor vessel and shroud dimensions, together with the configuration and placement of surveillance capsule holders, are provided In [6].

G-S

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Figure G-2-1: A Quadrant of DAEC Core aI E

G-6

GE Nuclear Energy GE-NE-A2201 0O0-0-Ola-R2 Non-Proprietary Version Figure G-2-2: Schematic View of (r, 0) Model n

D G-7

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version Figure G-2-3: Schematic View of (rz) Model 11 G-8

GE Nuciear Energy GE-NE-A22-00100 s-R2 Non-Proprietary Version G.3 Evaluaton Results 0.3.1 AEP Flux and Fluence at RPV ID Prior to implementation of extended power uprate, the calculated peak flux density (E>1 MeV) at the RPV ID is (( ], corresponding to current rated power of 1658 MW,. The EPU peak full power (1912 MW,) flux density is Reference 6 indicated that EPU Is expected to occur at 18.18 EFPY. or 5.74e8 EFPS. Assuming an 80% operation capacity factor during a 40 year plant life, or an equivalent 1.01e9 seconds of operation at full licensed power, then the 40 year plant life fluence is calculated as [ .]D This value has been reported in Revision 0 of this report.

The flux densities reported above do not include any bias adjustment, since they were calculated prior to NRC approval of NEDC-32983P, which recommends a ff E] bias adjustment If a (( D bias is applied, then the 40 year plant life fluence with 80%

capacity factor becomes D = 4.17e18 n/Cn2. This fluence value is used for P-T curve evaluation in this report.

G.3.2 AEP Flux and Fluence at Surveillance Capsule Location The calculated EPU flux density (E>1 MeV) at the 36" surveillance capsule location Is a1 ! without bias adjustment, or E with bias adjustment The lead factor is therefore (( = 0.77.

G-9

GE Nuclear Energy GE-NE-A22-00100-08-Ola-R2 Non-Proprietary Version G.4 REFERENCES

1. NEDC-32983P-A, Revision 1, Ucensing Topical Report General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluationse, December 2001.

Z Letter, S.A. Richards, USNRC to J.F. Kiapproth, GE-NE, Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (TAC NO. MA9891), MFN 01 -050, September 14, 2001.

3. CCC-543, TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14, Radiation Shielding information Center (RSIC), January 1994.

4. Regulatory Guide 1.190, uCalculatonal and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence', USNRC March 2001.

5. Letter, S. A. Richards (USNRC) to G. A. Wafford, *Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR I - Implementing Improved GE Steady-State Methods (TAC No. MA6481)', November 10, 1999.

6. Alliant Energy Transmittal NG-00-1026, R. McGee (DAEC) to W. Farrell (GE)

Transmittal of DIR T0313 AEP Flux Analysis', 6-8-2000.

G-10

no . !

Attachment 3 to NG-03-0608 August 18, 2003 General Electric Co. Affidavit of Proprietary Information and Request for Withholding from Public Disclosure

General Electric Company AFFIDAVIT I, George B. Stramback, state as follows:

(I) I am Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report GE-NE-A22-00100-08-01-R2, Pressure-TemperatureCurves for Duane Arnold Energy Center, Revision 2, Class I (GE Proprietary Information), dated August 2003. The proprietary information is delineated by a double underline inside double square brackets. Figures and large equation objects are identified with double square brackets before and after the object. In each case, the superscript notation 3 ) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.790(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission.

975F2d87I (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, resulting in potential products to General Electric; GBSM03-8-Af Duane Arnold P-T curves Rev 2.doc Affidavit Page

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a., and (4)b, above.

(5) To address 10 CFR 2.790 (b) (4),- the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and beliet consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed methods and processes, which GE has developed and applied to pressure-temperature curves for the BWR over a number of years. The development of the BWR pressure-temperature curves was achieved at a significant cost, on the order of % million dollars, to GE.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends OBS-03-8-Af Duane Arnold P-T curves Rev 2.doc Affidavit Page 2

beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this - -Iday of Oa1 u Z 2003.

dGorge B.Stramback General Electric Company

GBS-03-8-Af Duane Arnold P-T curves Rev 2.do Affidavit Page 3/

(IPS-03-8-Af Duane Arnold P-T curves Rev 2.doc Affidavit Page 3 .

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v t e Letter Sequence Other
TAC:MB8750, (Open)
Initiation Acceptance... Administration Questions 15 August 2003 22 August 2003 22 August 2003 Answers Results Approval... Other: ML032270197, ML032310536, ML032340204, NG-03-0608, Revised Non-Proprietary GE Report GE-NE-A22-00100-08-01a-R2, Pressure-Temperature Curves for Duane Arnold Energy Center
MONTHYEAR2003-08-22 [Table View]

NG-03-0608, Revised Non-Proprietary GE Report GE-NE-A22-00100-08-01a-R2, Pressure-Temperature Curves for Duane Arnold Energy Center

Text

SUMMARY

1.0 INTRODUCTION

6.0 CONCLUSION

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

REFERENCES:

Reference:

References:

References:

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