ML15253A311
| ML15253A311 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/30/2015 |
| From: | NextEra Energy Duane Arnold |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15253A328 | List: |
| References | |
| NG-15-0235, TSCR-144 | |
| Download: ML15253A311 (44) | |
Text
ATTACHMENT 6 TO NG-15-0235 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST (TSCR-1 44)
TO REVISE AND RELOCATE PRESSURE AND TEMPERATURE LIMITS CURVES TO A PRESSURE AND TEMPERATURE LIMITS REPORT PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
FOR 32 AND 54 EFFECTIVE FULL POWER YEARS (EFPY)
NON-PROPRIETARY VERSION 43 pages follow
FPL Energy Duane Arnold,.LLC Duane Arnold Energy Center Pressure and Temperature Limits Report (PTLR) for 32 and 54 Effective Full-Power Years (EFPY)
NON-PROPRIETARY VERSION-
ATTACHMENT 6 Page 2 of 43 Table of Contents Section Pane 1.0 Purpose 4
2.0 Applicability 4
- 3.0 Methodology 5
4.0 Operating Limits 6
5.0 Discussion 7
6.0 References 13 Figure 1 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 16 EFPY Figure 2 DAEC P-T Curve B (Normal Operation - Core Not Critical),
17 32 EFPY Figure 3 DAEC P-T Curve C (Normal Operation - Core Critical), 32 18 EFPY Figure 4 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 19 EFPY Figure 5 DAIEC P-T Curvae B (Normal Operation - Core Not Critical),
20 54 EFPY Figure 6 DAEC P-T Curve C (Normal Operation - Core Critical), 54 21 EFPY Table 1 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 22 EFPY Table 2 DAEC P-T Curve B (Normal Operation - Core Not Critical),
25 32 EFPY Table 3 DAEC P-T Curve C (Normal Operation - Core Critical), 32 28 EFPY Table 4 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 31 EFPY Table 5 DAEC P-T Curve B (Normal Operation - Core Not Critical),
34 2
- ATTACHMENT 6
- Page 3 of 43 54 EFPY Table 6 DAEC P-T Curve C (Normal Operation - Core Critical), 54 37 EFPY Table 7 DAEC ART Table for 32 BEPY 40 Table 8 DAEC ART Table for 54 EPPY 41 Table 9 DAEC Summaiy* of Nozzle Stress Intensity Factors 42 Appendix A DAIEC Reactor Vessel Material Surveillance Program 43 3
ATTACHMENT 6 Page 4 of 43 1.0 Purpose The purpose of the Duane Arnold Energy Center (DAEC) Pressur'e and Temperature Limits Report (PTLR) is to present operating limits relating to:
- 1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool-down and Hydrostatic/Class 1 Leak Testing.;
- 2. RCS Heat-up and Cool-down rates;
- 3. Reactor Pressure Vessel (RPV) to RCS coolant AT requirements during Reci-culation Pump startups;
- 4. RPV bottom head coolant temperature to RPV coolant temperature AT requirements during Recirculation Pump startups;
This report has been prepared in accordance with the requirements of Licensing Topical Report SIR-05-044, Revision 1-A [1].
2.0 Applicability This report is applicable to the DAEC RPV for up to 32 and 54 Effective Full-Power Years (EFPY).
The following DAEC Technical Specification (TS) is affected by the information contained in the present report:
TS 3.4.9 RCS Pressure and Temperature (PIT) Limits 4
ATTACHMENT 6 Page 5 of 43 3.0 Methodology The limits in this report were derived as follows:
- 1. The methodology used is in accordance with Reference [1], which has been approved by the NRC.
- 2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [2], using the RAMiA computer code, as documented in Reference [3].
- 3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99)
[4], as documented in Reference [5].
- 4. The pressure and temperature limits were calculated in accordance with Reference
[1 ], "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,"
as documented in Reference [6].
- 5. This revision of the pressure and temperature limits is to incorporate the following changes:
Initial issue of PTLR.
Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [7], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.
Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.
5
ATTACHMENT 6 Page 6 of 43 4.0 Operating Limits The pressure-temperature (P-T). curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incor~porate the appropriate non-beltline limits and irradiation embrittlement effects in the beitline region.
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) core not critical operation, referred to as Curve 13; and (c) core critical operation, referred to as Curve C.
Complete P-T curves were developed for 32 and 54 EPPY for DAEC, as documented in Reference [6]. The DAEC P-T curves for 32 EPPY are provided in Figure 1 through Figure 3, and a tabulation of the overall composite curves (by region) is included in Table 1 through Table
- 3. The DAEC P-T curves for 54 EFPY are provided in Figure 4 through Figure 6, and a tabulation of the curves is included in Table 4 thr'ough Table 6. The ART values for the DAEEC vessel beitline materials are shown in Table 7 for 32 EFPY and Table 8 for 54 IEFPY, taken from Reference [5].
The resulting P-T curves are based on the geometry, design, and materials informnation for the DAEC vessels with the following conditions:
- Heatrup/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figure 1 and Figure 4: Curve A): < 250F/hourl.
- Normal Operating Heat-up/Cool-down rate limit (Figure 2 and Figure 5: Curve B - non-nuclear heating, and Figure 3 and Figure 6: Curve C - nuclear heating): < 1000F/hour 2.
- RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 145°F.
SInterpreted as the temperature change in any 1-hour period is less than or equal to 25°F 2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.
6
ATTACHMENT 6 Page 7 of 43
- Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 50°F.
o RPV flange and adjacent shell temperature limit: _> 74°F.
To address the NRC condition regarding lowest service tem~perature (LST) in Reference [1], the minimum temperature is set to 74°F, which is equal to the RTNDT,rna. + 60°F, for all curves. This value is consistent with the minimum temperature limits and the minimum bolt-up temperature specified in Technical Specifications (Figure 3-.6-1 in Reference [9]). This value bounds the LSTs for the ferritic non-RPV components of the reactor coolant pressure boundary (RCPB).
Review of piping design specifications [10] the maximum impact test temperature is 0°F [page 2 in 10a, b, and cj, hence LSTs for non-RPV RCPB components are bounded by the boltup temperature of 74°F.
The composite P-T curves are extended below 0 psig to -14.7 psig based on the evaluation documented in Reference [11], which demonstrates that the P-T curves are applicable to negative gauge pressures. A pressure of -14.7 psig bounds the maximum expected vacuum pressure as well as externally applied pressures the RPV may experience. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at 0 psig. However, the minimum analyzed RPV pressure is -14.7 psig.
5.0 Discussion The adjusted reference temperature (ART) of the limiting beitline material is used to adjust the beltline P-T curves to account for irradiation effects. RG 1.99 [4] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.
The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the DAEC vessel plate, weld, and forging materials [8]. The Cu and Ni values were used with Table 7
ATTACHMENT 6 Page 8 of 43 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for plates and forgings. 1-owever, for materials where credible surveillance data exists, a fitted CF may be used if it bounds the RG 1.99 CF.
For DAEC, the peak RPV ID fluence values of 4.26 x 1018 n/cm2 at 32 EFPY and 7.49 x 1018 n/cm2 at 54 EFPY were obtained fr'om Reference [3] and was calculated in accordance with RG 1.190 [2]. These fluence values for the limiting lower-intermediate shell plate (Heat No. B0673-
- 1) are based upon an attenuation factor of 0.765 for a postulated 1/4t flaw. Consequently, the 1/4t fluence for 32 and 54 EFPY for the limiting lower-intermediate shell plate are 3.26 x 1018 n/cm2 and 5.73 x 1018 n/cm2, respectively, for DAEC.
The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from struactural discontinuities such as nozzles. Based on the ART evaluation in Reference [5], the instrumentation (N 16) and recirculation inlet (N2) nozzles are located in the beltline region. The Ni16 nozzle at DAEC is a ferritic forged nozzle design, which is welded to the RPV using a full penetration weld rather than the partial penetration nozzle design used in other plants. Although the ART value for the N 16 nozzle is higher than that of the N2 nozzle [5], the limiting beltline nozzle is determined by examining the thermal transients for each nozzle from References [12a, c]. The N1 6 nozzle does not have any significant cycling.
Therefore, the N2 nozzle is considered the limiting nozzle. The peak fluence values for the limiting beltline nozzle is sununarized in Table 7 for 32 EFPY and Table 8 for 54 EFPY. The feedwater (FW) nozzle is considered in the evaluation of the non-b eltilne (upper vessel) region P-T limits.
The P-T curves for the core not critical and core critical operating condition at a given EFPY apply for both the l/4t (inside surface flaw) and 3/4t (outside surface flaw) locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the l/4t 8
ATTACHMENT 6 Page 9of 43 and the 3/4t locations. Thifs is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4t location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stress at the 1/4t location. This approach is conservative because irradiation effects cause the allowable toughness at l/4t to be less than that at 3/4t for a given metal temperature. This approach causes no operational difficulties, since the BWR. is at steam saturation conditions during normal operation, which is well above the P-T curve limits.
For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cool-down temperature rate of<* 100°F/hour for which the curves are applicable. The core not critical' and the core critical curves were developed to bound all Service Level A/B RPV thermal transients. For the hydrostatic pressure and leak test cur~e (Curve A), a coolant heat-up and cool-down temperature rate of *: 25°F/hour must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if
- necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing.
if the pressure test heat-up/cool-down rate limits cannot be maintained.
The initial RTNDT, the chemistr (weight percent copper and nickel), and ART at the 1/4t location for all RPV beltline materials significantly affected by fluence (i.e. fluence >1017 n/cm2 for E > 1 MeV) are shown for DAEC in Table 7 for 32 EFPY and Table 8 for 54 IEFPY [5]. Use of initial RTNDT values in the determination of P-T curves for DAEC was approved by the NRC in Reference [22].
Per Reference [5] and in accordance with Appendix A of Reference [1], the DAIEC representative weld and plate surveillance material data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Progr'am (ISP) [13, 14].
9
ATTACHMENT 6 Page 10 of 43 Use of surveillance data from the BWRVIP ISP for DAEC was approved by the NRC in Reference [23].
For DAEC, the fitted CF for the target beitline plate (Heat No. B0673-1), which is based on credible surveillance data, bounds the RG 1.99 CF [13, 14]. Therefore, the fitted CF is used to calculate the ART for the target plate. References [13, 14] contain surveillance capsule test results for the DAEC representative weld material. The material heat for the DAEC representative surveillance capsule weld material (DA1 SMAW) does not match the target weld material heat (432Z0471). Consequently, the CF calculated using the RG 1.99 tables is used in the determination of ART for the target beltline weld.
The ANSYS Mechanical Release 8.1 (with Service Pack 1) [15] finite element computer program was used to develop the stress distributions through the FW nozzle, and these stress distributions were used in the determination of the stress intensity factors for the FW nozzles
[16]. At the time that the analyses above were perfor~med, the ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B [17] Quality Assurance Program for nuclear quality-related work.
The plant-specific DAEC FW nozzle finite element analysis [16] was performed to determine stress intensity factors due to thaough-wall pressure stress distributions and thermal stress distrbutions due to bounding thermal transients. The resulting stress intensity factors are summarized in Table 9. Detailed information regarding the stress analysis can be found in Reference [16], and the calculation of stress intensity factors is described in Reference [18, Section 3.2.2]. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle [16]:
,* With respect to operating conditions, stress distributions were developed for a thelrmal
- shock of 4000F in Reference [16]. The stresses were scaled to represent a thermal shock of 45 9°F, which represents the-maximum thermal shock for the feedwater 10
ATTACHMENT 6 Page 11 of 43 nozzle during normal operating conditions. The stress results for a 4590F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting feedwater nozzle, as a shock of 4590F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 1O00°F/I" startup transient [12].
Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 1 000F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region. The thermal stress distribution, corresponding to the limiting time presented in Reference [16], along a linear path through the nozzle corner is used. The BIB/IF methodology presented in Reference
[1] is used to calculate the thermal stress intensity factor, K1T due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.
- Heat transfer coefficients were determined using the methodology defined in Section 2.2.2 of Reference [1 6.b]. These values were determined for various regions of the finite element model and for 100, 40, 25, and 3 percent flow rates. The temperatures used are based upon a thennal shock from 500°F to 1 000F. Heat transfer coefficients are calculated for different flow rates, when appropriate.
- With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the finite element model. The pressure stress distribution was taken along the same path as the thermal stress distribution. The boundary integral equation/influence function (BIB/IF) methodology presented in Reference [1] was used to calculate the pressure stress intensity factor, K1p, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting I~p can be linearly scaled to determine the Kwp for various RPV internal pressures.
- An axisyrmmetric two-dimensional finite element model of the FW nozzle was const-ructed [16.a]. A factor of 3.2 was applied to the vessel radius to account for 3-D effects on the pressure stresses [16], Material properties were based on the 2001 11
ATTACHMENT 6 Page 12 of 43
- ASME Code,Section II, Part D, with 2003 Addenda [19]. The properties were taken at a temperature of 3 00°F, which is approximately the average temperature for the thermal shock transient from 500°F to 1 00°F. The use of temperature independent material properties is consistent with design basis documents. Use of temperature-dependent material properties is expected to have minimal impact on the results of the analysis.
For the N2 nozzle, the finite element analysis developed for th~e FW nozzle [16] was used to determine a th~rough-wall stress distribution for the similar N2 nozzle, with conservative scaling factors applied to account for differences in geometry and limiting thermal transient [18, Section 3.2.2]. The following inputs were used to scale the FW nozzle FEA stress distributions for the N2 nozzle;
- With respect to operating conditions, the bounding transient is the 1 00°F/hr startup and cooldown [12]. The results of the finite element analysis in [16.b] were scaled to that of a 1 00°F step change, which conservatively bounds the 1 00°F/lh* transient.
- To account for geometric differences between the FW and N2 nozzles, a thermal scaling factor was developed [18, Section 3.2.2]. The hoop stresses are extracted along a path through the blend radius per Reference [1 6.b]. The length of this path determines the through-wall temperature gradient, with a thicker section resulting in higher thermal stresses. Comparing the dimensions of the FW and N2 nozzles shows that the path length for the N2 nozzle is shorter that for that of the feedwater nozzle, which would result in a decrease in the stress coefficients. Therefore, the thermnal scaling factor to account for geometry differences was conservatively ignored.
- With respect to pressure stress, to account for the effect of geometric differences between the FW and N2 nozzles, a geometric scaling factor of 1.21 is applied to the FW nozzle pressure stress coefficients to obtain the corr'espon~ding stress coefficients for the N2 nozzle [18, Section 3.2.2].
12
ATTACHMENT 6 Page 13 of 43 6.0 References
- 1. Licensing Topical Report (LTR) BWROG-TP-1 1-022-A (SIR-05-044), Revision 1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," June 2013, ADAMS Accession No. ML13277A557.
- 2. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001.
- 3. TransWare Report No. DAE-FLM-001-R-004, Revision 0, "Duane Arnold Energy Center Fluence Assessment Report - End of Cycle 24," April 3, 2015. SI File No.
1500134.202.
- 4. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.
- 5. SI Calculation No. DAEC-20Q-324, Revision 2, "Duane Arnold RPV Beltline ART Evaluation," June 2015.
- 6. SI Calculation No. DAEC-20Q-326, Revision 1, "Duane Arnold Updated P-T Curve Calculation for 32 and 54 EFPY," June 2015.
- 7. Code of Federal Regulations, Title 10, Part 50, Section 59, "Changes, tests and experiments," Aug. 28. 2007.
- 8. General Electric Report No. GE-NE-A22-0O0OI0-08-01-R2, Revision 2, "Pressure-Temperature Curves for Duane Arnold Energy Center," August 2003. SI File No. DAEC-2OQ-2Ol.
- 9. Letter fr'om Mohan C. Thadani (N4RC) to Lee Liu (Iowa Electric Light and Power Company), "Duane Arnold Energy Center, License Amendment No. 121 to License No.
DPR-49," May 28, 1985, ADAMS Accession No. ML021890474, SI File No.
1500134.203.
13
ATTACHMENT 6 Page 14 of 43
- a. FPL Document No. APED-A61-019, GE Document No. 22A1295AD, Revision 3, "Pressure Integrity of Piping and Equipment Pressure Parts - Data Sheet."
- c. FPL Document No. BECH-M190<GLE>, Revision 1, "Piping Class GLE."
- d. FPL Document No. BECH-M190<EILE>, Revision 1, "Piping Class HLE."
- a. FPL Drawing No. APED-B11-003<2>, GE Drawing No. 135B9990, Sheet 2, Revision 1, "Nozzle Thermal Cycles (Recirculation Inlet)."
- b. FPL Drawing No. APED-B 11-003<4>, GE Drawing No. 135B9990, Sheet 4, Revision 2, "Nozzle Thermal Cycles (Feedwater)."
- c. FPL Drawing No. APED-B11-003<7>, GE Drawing No. 135B9990, Sheet 7, Revision 0, "Nozzle Thermal Cycles (Instrumentation & Core Diff. Press &
Liquid Control)."
- 13. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014. 3002003144. SI File No. BWRVIP-135P. EPRI PROPRIETARY INFORMATION.
- 14. BWRVIP-279NP, BWR Vessel and Internals Project, Testing and Evaluation of the Duane Arnold 108° Capsule, EPRI, Palo Alto, CA: 2014. 3002003134. SI File No.
1500134.202.
- 15. ANSYS Mechanical Release 8.1 (w/ Service Pack 1), Ansys, Inc., June 2004.
14
ATTACHMENT 6 Page 15 of 43
- a. SI Calculation No. DAEC-20Q-305, Revision 0, "Feedwater Nozzle Finite Element Model," July 2, 2007.
- b. SI Calculation No. DAEC-20Q-3 06, Revision 0, "Feedwater Nozzle Stress History Development for Green's Functions," July 2, 2007.
- 17. Code of Federal Regulations, Title 10, Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Aug. 28. 2007.
- 18. FPL Document No. APED-B 11-3320-326, Revision 0, SI Calculation No. DAEC-20Q-326, Revision 0, "Revised Pressure Temperature Curves," February 21, 2008.
- 19. ASME Section II, Part D, 2001 Edition through 2003 Addenda.
- 20. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," July 10, 2014..
- 21. BWRVIP-86, Revision 1 : BWIR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2008. 1016575.
- 22. Letter to Mr. Mark A. Peifer, Site.Vice President Duane Arnold Energy Center, from Darl S. Hood, Nuclear Regulatory Commission, dated August 25, 2003, subject: DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT REGARDING PRESSURE AND TEMPERATURE LIMIT CURVES (TAC NO. MB8750). ADAMS Accession No. ML032310536.
- 23. Letter to Mr. Gary Van Middlesworth, Vice President Duane Arnold Energy Center, from Richard B. Ennis, Nuclear Regulatory Commission, dated November 27, 2006, subject:
DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT REGARDING REVISION TO THE REACTOR PRESSURE vESSEL MATERIAL SURVEILLANCE PROGRAM (TAC NO. MC93 13). ADAMS Accession No. ML062830019.
15
ATTACHMENT 6 Page 16 of 43 Figure 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 EFPY Curve A - Pressure Test, Composite Curves
-Beltilne Bottom H-ead Non-Beltline
- Overall 1300 1200 1100 1000-900 800 6 00 200 100+-----
0 K 0
50 100 150 200 Minimum Reactor Vessel Metal Temperature (0F) 250 300 16
ATTACHMENT 6 Page 17 of 43 Figure 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 32 EFPY Curve B - Core Not Critical, Composite Curves
- Bitin m Bottom Head
-- Non-Beitline 4*-i=vOverall 1300 1200 1100 1000 g900 a.g 600 1400 0
-100 250 0
50 100 150 200 Minimum Reactor Vessel Metal Temperature (F) 300 17
ATTACHMENT 6 Page 18 of 43 Figure 3: DAEC P-T Curve C (Normal Operation - Core Critical), 32 EFPY Curve C - Core Critical, Composite Curves
--Beitline
Bottom Head
- Non-Beitline Overall 1200
! -I 11001
, i' 900........
,-- :--.- :I I
10001I I
g00
'0-
....g
- : :...!
- I........
5001 ='
600
-- [
3001....----.
I 1/i*7-**'MnmmRV I
-IPress raure = -474 psIg
-100 0
50 100 150 200 250 300 Minimum Reactor Vessel Metal Temperature (F)o 18
ATTACHMENT 6 Page 19 of 43 Figure 4: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY Curve A - Pressure Test, Composite Curves Be ltiltle
Bottom Head Non-Beitline
- OveraHl 1300
- -- f
'1 1 1200 I
I 110
.. i--I-I Ii
-0 0T--
-I.....
1
...- -- -1
- 1800 I
M nmu-B l
-Up S70
-6001
-J I
~
00-50 100 -5 0/20 Minimum Reactor Vessel Metal Temperature (*F) 19
ATTACHMENT 6 Page 20 of 43 Figure 5: DAEC iP-T Curve B (Normal Operation - Core Not Critical), 54 EFPY Curve B - Core Not Critical, Composite Curves
-Beil~ine Bottom Head
?
-= - -
-Non-Beitline Overall 1300 1200 1100 1000 900
.* 800
.E 500 1-400 300 200 100 I
0
-100 0
50 100 150 200 Minimum Reactor Vessel Metal Temperature (F) 250 300 20
ATTACHMENT 6 Page 21 of 43 Figure 6: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY 1300 Curve C - Core Critical, Composite Curves
-8eltline ram - Bottom Head m
Non-Beitline
- 'Overall
... *---"*"""* - r-
'........ T
-T -T I -* T T -..
- i i
im
!i 1200 1100 1000 900 800 4..
&- 4oo 300 200 100 0
-100 0
50 100 150 200 Minimum Reactor Vessel Metal Temperature (0F}
250 300 21
ATTACHMENT 6 Page 22 of 43 Table 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 EFPY Beitline Region
.4
- 9.
OF 74.0 74.0 96.4 111.8 123.6 133.1 141.1 148.0 154.0 159.4 164.3 168.8 172.8 176.6 180.1 183.4 186.5 189.4 192.1 194.7 197.2 199.5 201.8 psi 0.0 285.4 334.5 383.7 432.8 481.9 531.1 580.2 629.4 678.5 727.6 776.8 825.9 875.0 924.2 973.3 1022.5 1071.6 1120.7 1169.9 1219.0 1268.2 1317.3 22
ATJTACHMENT 6 Page 23 of 43 Table 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 EFPY (cont.)
Non-Beitline Regiion
°F psi 74.0 0.0 74.0 312.6 104.0 312.6 104.0 863.7 107.9 913.6 111.5 963.6 114.9 1013.5 118.0 1063.5 121.0 1113.4 123.8 1163.4 126.4 1213.3 129.0 1263.3 131.3 1313.2 23
ATTACHIMENT 6 Page 24 of 43 Table 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 32 EFPY (cont.)
Bottom Head Region
°F psi 74.0 0.0 74.0 795.7 79.2 843.7 83.9 891.6 88.2 939.6 92.1 987.5 95.8 1035.5 99.2 1083.4 102.4 1131.4 105.4 1179.3 108.2 1227.3 110.9 1275.3 113.5 1323.2
- 24
ATTACHMENT 6 Page 25 of 43 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 32 EFPY Beitline Region U~
U 74.0 74.0 102.1 120.0 133.2 143.6 152.2 159.5 165.9 171.6 176.7 181.3 185.6 189.5 193.1 196.4 199.6 202.6 205.4 208.0 210.6 213.0 215.3 217.5 219.6 221.6 psi 0.0 134.7 184.0 233.2 282.5 331.7 381.0 430.2 479.5 528.7 578.0 627.2 676.5 725.7 775.0 824.2 873.5 922.7 972.0 1021.2 1070.5 1119.7 1169.0 1218.2 1267.5 1316.7 25
ATTACHMENT 6 Page 26 of 43 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 32 EFPY (cont.)
Non-Beitline Reg~ion JTempeature' Pressure 0F psi 74.0 0.0 74.0 186.2 84.2 228.3 92.4 270.5 99.2 312.6 134.0 312.6 134.0 667.3 137.3 717.1 140.3 766.8 143.2 816.6 145.8 866.4 148.4 916.1 150.8 965.9 153.1 1015.6 155.2 1065.4 157.3 1115.2 159.3 1164.9 161.2 1214.7 163.0 1264.4 164.8 1314.2 26
ATITACHMENT 6 Page 27 of 43 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 32 EFPY (cont.)
Bottom Head Region Ua 9
9 P-T Curve Temperature 0F P~T Curve Pressure psi 74.0 74.0 80.9 86.9 92.3 97.2 101.6 105.7 109.5 113.0 116.3 119.3 122.2 125.0 127.6 130.0 132.4 134.6 0.0 542.4 591.0 639.6 688.2 736.8 785.4 834.0 882.6 931.2 979.8 1028.4 1077.0 1125.6 1174.2 1222.8 1271.4 1320.0 27
ATTACHMENT 6 Page 28 of 43 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 32 EFPY Beltline Reieion 0F psi 74.0 0.0 74.0 98.8 123.0 147.6 147.3 196.4 163.6 245.3 175.8 294.1 1.85.7 342.9 193.9 391..7 200.9 440.5 207.1 489.3 212.6 538.1 21.7.6 586.9 222.1 635.7 226.2 684.5 230.0 733.3 233.6 782.1 236.9 830.9 240.0 879.7 242.9 928.5 245.7 977.3 248.3 1026.1 250.8 1074.9 253.2 1123.8 255.4 1172.6 257.6 1221.4 259.7 1270.2 261.7 1319.0
- 28
ATTACHMENT 6 Page 29 of 43 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 32 EFPY (cont.)
Non-Beltline Reg~ion 9*
9 74.0 74.0 97.7 112.6 123.5 132.1 139.2 174.0 174.0 177.3 180.3 183.2 185.8 188.4 190.8 193.1 195.2 197.3 199.3 201.2 203.0 204.8 206.5 psi 0.0 93.5 137.3 181.2 225.0 268.8 312.6 312.6 667.3 717.1 766.8 816.6 866.4 916.1 965.9 1015.6 1065.4 1115.2 1164.9 1214.7 1264.4 1314.2 1364.0 29
ATT'ACHMENT 6 Page 30 of 43 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 32 EFPY (cont.)
Bottom Head Region
- F psi 74.0 0.0 74.0 361.2 88.1 409.3 99.0 457.4 108.0 505.4 115.7 553.5 122.3 601.6 128.1 649.6 133.3 697.7 138.1 745.8 142.4 793.9 146.4 841.9 150.0 890.0 153.5 938.1 156.7 986.1 159.7 1034.2 162.5 1082.3 165.2 1130.4 167.8 1178.4 170.2 1226.5 172.5 1274.6 174.8 1322.6 30
ATITACHMENT 6 Page 31 of 43 Table 4: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY Beltline Re~ion
°F psi 74.0 0.0 74.0 261.7 102.3 309.9 120.3 358.1 133.5 406.3 143.9 454.5 152.5 502.7 159.9 550.9 166.3 599.1 172.0 647.3 177.1 695.5 181.7 743.7 185.9 791.9 189.8 840.1 193.5 888.3 196.8 936.4 200.0 984.6 203.0 1032.8 205.8 1081.0 208.4 1129.2 2:11.0 1177.4 213.4 1225.6 215.7 1273.8 217.9 1322.0 31
ATTACHMENT 6 Page 32 of 43 Table 4: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY (cont.)
Non-Belfline Reg~ion OF psi 74.0 0.0 74.0 312.6 104.0 312.6 104.0 863.7 107.9 913.6 111.5 963.6 114.9 1013.5 118.0 1063.5 121.0 1113.4 123.8 1163.4 126.4 1213.3 1£29.0 1263.3 131.3 1313.2 32
ATTACHMENT 6 Page 33 of 43 Table 4: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY (cont.)
Bottom Head Region
.1 I
P-T Curve Temperature 0F P-TCurve Pressure psi 74.0 74.0 79.2 83.9 88.2 92.1 95.8 99.2 102.4 105.4 108.2 110.9 113.5 0.0 795.7 843.7 891.6 939.6 987.5 1035.5 1083.4 1131.4 1179.3 1227.3 1275.3 1323.2 33
ATTACHMENT 6 Page 34 of 43 Table 5: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY Beltline Reg~ion 9
74.0 74.0 109.9 130.6 145.2 156.5 165.7 173.4 180.2 186.1 191.4 196.2 200.5 204.6 208.3 211.7 215.0 218.0 220.9 223.6 226.2 228.6 230.9 233.2 235.3 237.3 psi 0.0 117.0 166.8 216.7 266.6 316.4 366.3 416.2 466.0 515.9 565.7 615.6 665.5 715.3 765.2 815.1 864.9 914.8 964.6 1014.5 1064.4 1114.2 1164.1 1214.0 1263.8 1313.7 34
ATTACHMENT 6 Page 35 of 43 Table 5: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY (cont.)
Non-Beltline Region 0F psi 74.0 0.0 74.0 186.2 84.2 228.3 92.4 270.5 99.2 312.6 134.0 312.6 134.0 667.3 137.3 717.1 140.3 766.8 143.2 816.6 145.8 866.4 148.4 916.1 150.8 965.9 153.1 1015.6 155.2 1065.4 157.3 1115.2 159.3 1164.9 161.2 1214.7 163.0 1264.4 164.8 1314.2 35
ATTACHMENT 6 Page 36 of 43 Table 5: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY (cont.)
Bottom Head Region
°F psi 74.0 0.0 74.0 542.4 80.9 591.0 86.9 639.6 92.3 688.2 97.2 736.8 101.6 785.4 105.7 834.0 109.5 882.6 113.0 931.2 116.3 979.8 119.3 1028.4 122.2 1077.0 125.0 1125.6 127.6 1174.2 130.0 1222.8 132.4 1271.4 134.6 1320.0 36
ATTACHMENT 6 Page 37 of 43 Table 6: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY Beitline Reg~ion e
74.0 74.0 133.7 160.2 177.4 190.2 200.3 208.8 216.0 222.3 227.9 233.0 237.5 241.7 245.6 249.2 252.5 255.7 258.6 261.4 264.0 266.6 268.9 271.2 273.4 275.5 277.5 psi 0.0 90.9 139.9 189.0 238.1 287.1 336.2 385.3 434.4 483.4 532.5 581.6 630.6 679.7 728.8 777.9 826.9 876.0 925.1 974.1 1023.2 1072.3 1121.4 1170.4 1219.5 1268.6 1317.6 37
ATTACHMENT 6 Page 38 of 43 Table 6: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY (cont.)
Non-Beltline Reg~ion e
"F 74.0 74.0 97.7 112.6 123.5 132.1 139.2 174.0 174.0 177.3 180.3 183.2 185.8 188.4 190.8 193.1 195.2 197.3 199.3 201.2 203.0 204.8 206.5 psi 0.0 93.5 137.3 181.2 225.0 268.8 312.6 312.6 667.3 717.1 766.8 816.6 866.4 916.1 965.9 1015.6 1065.4 1115.2 1164.9 1214.7 1264.4 1314.2 1364.0 38
ATITACHMENT 6 Page 39 of 43 Table 6: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY (cont.)
Bottom Head Region
- e.
e 74.0 74.0 88.1 99.0 108.0 115.7 122.3 128.1 133.3 138.1 142.4 146.4 150.0 153.5 156.7 159.7 162.5 165.2 167.8 170.2 172.5 174.8 psi 0.0 361.2 409.3 457.4 505.4 553.5 601.6 649.6 697.7 745.8 793.9 841.9 890.0 938.1 986.1 1034.2 1082.3 1130.4 1178.4 1226.5 1274.6 1322.6 39
ATTACHMENT 6 Page 40 of 43 Table 7: DAEC ART Table for 32 EIFPY Description ID No.
Heat No.
Lot No.
Initial RTNDT (0F)
Chemistry Cu Ni (wt %)
(wt %)
Chemistry Factor (0F)
Adjustments For 1/4t Margin ARTNDT Terms ART (0F)
(0F) (0F)
(0F)
Shell Ring #1 1-18 C6439-2
-40 0.09 0.51 58.0 36.5 17.0 0.0 110.5
~Shell Ring #1 1-19 B30402-1
-40 0.13 0.47 87.1 64.8 17.0 0.0 128.8
~Shell Ring #2 1-20 B0436-2
-10 0.15 0.64 111.0 76.8 17.0 0.0 120.8 Shell Ring #2 1-21 80673-1
-10 0.15 0.65
- ()
102.0 8.5 0.0 129.0 Lower D1,D2 432Z_4521 B020A27A
-50 0.01 0.98 20.0 11.3 5.6 0.0
-27.5 Lower D1,D2 432Z0471 B003A27A
-50 0.03 0.91 41.0 23.1 11.5 0.0
-3.8 wJ Lower-mnt EI,E2 432Z4521 B020A27A
-50 0.01 0.98 20.0 12.7 6.4 0.0
-24.6 Lower-]nt EI,E2 432Z0471 8003A27A
-50 0.03 0.91 41.0 26.1 13,0 0.0 2.2 Girth DE 09L853 L017A27A
-50 0.03 0.88 41.0 25.8 12.9 0.0 1.5 Girth DE 07L669 K004A27A
-50 0.03 1.02 41.0 25.8 12.9 0.0 1.5 Girth DE CTY538 A027A27A
-50 0.03 0.83 41.0 25.8 12.9 0.0 1.5 (n-Nozzle N16 Q2Q5VW
-40 0.18 0.85 141.8 63.7 17.0 0.0 137.7 0z Nozzle N2 Q2Q6VVV 40 0.18 0.84 141.6 29.2 14.6 0.0 98.5 412Z051 K910A27A
.l;W N2/N16 08R4818 K904A27A N _ Nozzle Welds 659T568 H721A27A
-50 0.03 1.00 41.0
.18.4 9.2 0.0
-13.2 S(bounding) 661Y494 F927A27A 661Y439 E916A27A Fluence Attenuation, Fluence Factor, Wall Thickness (in.)
at ID 1/4t Fluence (*114t FF Location Full 114t (n/crnz) e"°'Z4x (n/cm )
flo.z~u..viog ?)
- IShell Ring #1 1-18 4.469 1.117 3.33E+18 0.765 2.55E+18 0.629
~Shell Ring #1 1-19 "4.469 1.117 3.33E+18 0.765 2.55E=+18 0.629
- Shell ig#
-0 449 117 421+8 07532E1
.9 Ihl Ring #2 1-20 4.469 1.117 4.26E1=+8 0.765 3.26E+18 0.692 Sh well Rig 1-D21 4.469 1.117 4265E+18 0.765 1.95E+18 0.5692 Lower DI,D2 4.469 1.117 2.55E+18 0.765 1.95E+18 0.563
- , Lower-n EI,D2 4.469 1.117 3.43E+18 0.765 2.92E+18 0.5636 w Lower-lnt EI,E2 4.469
.1.117 3.43E=+18 0.765 2.62E+18 0.636 Gi owr-th DEI 4.469 1.117 3.43E+18 0.765 2.55E+18 0.636 Girth DE 4.469
- 1.117 3.33E+18 0.765 2.55E=+18 0.629 Girth DE "4.469 1.117 3.33E+18
- 0.765 2.55E=+18 0.629.
S Nozzle N16 4.469.
1.117 1.53E+18 0.765
- 1.17E+18
- 0.449 z
Nozzle N2 4.469 1.117 3.53E+17 0.765 2.70E+17 0.206 (b ozzending) 4.469 1.117 1.53E+18 0.765 1.17E+18 0.449 Note: The source of redacted proprietary information is Reference [13]
SREDACTED ({] } PROPRIETARY INFORMATION 40
ATTACHMENT 6 Page 41 of 43 Table 8: DAEC ART Table for 54 EFPY Description ID No.
Initial RTNDT (0F)
Chemistry Heat No. ILot No.
Adjustments For 1/4t Cu (wt Ni twt %A' Chemistry Factor
(°F)
ARTNDT Margin Terms ART
(°F'*
cJA j S
Shell Ring #1 1-18 C6439-2 40 0.09 0.51 58.0 45.1 17.0 0.0 119.1 Shell Ring #1 1-19 B0402-1
-40 0.13 0.47 87.1 67.8 17.0 0.0 141.8 Shell Ring #2 1-20 B0436-2 10 0.15 0.64 111.0 93.7 17.0 0.0 137.7
__ he___Rn__2__1-21_B0673-_
10 0.15 0.65
(( 124.5 8.5 0.0 151.5 Lower D1,D2 432Z4521 B020A27A -50 0.01 0.98 20.0 14.1 7.0 0.0 -21.9 Lower D1,D2 432Z0471 B003A27A -50 0.03 0.91 41.0 28.8 14.4 0.0 7.7 w1 Lower-lnt ElE2 432Z4521 B020A27A -50 0.01 0.98 20.0 15.6 7.8 0.0 -18.7 w Lower-lnt EI,E2 432Z0471 B003A27A -50 0.03 0.91 41.0 32.1 16.0 0.0 14.1 Girth DE 09L853 L017A27A -50 0.03 0.88 41.0 31.9 16.0 0.0 13.8 Girth DE 07L669 K004A27A -50 0.03 1.02 41.0 31.9 16.0 0.0 13.8 Girth DE CTY538 A027A27A -50 0.03 0,83 41.0 31.9 16,0 0.0 13,8 Nozzle N16 Q2Q5VW 40 0.18 0.85 141.8 85.0 17.0 0.0 159.0 0 Nozzle N2 Q2Q6VW -40 0.18 0.84 141.6 40.4 17.0 0.0 114.4 !l* 412Z051 K910A27A Ww N2/N16 08R4818 K904A27A Ii Nozzle Welds 659T568 H721A27A -50 0.03 1.00 41.0 24.6 12.3 0.0 -0.8 S(bounding) 661Y494 F927A27A 661Y439 E916A27A___ Fluence Attenuation, Fluence Factor, Wall Thickness (in.) at ID 114t Fluence @1/4t FF Location Full 114t (n/cmz) e"O'24X (nlcm ) f(U.2*-J.lUiog t) S Shell Ring #1 1-18 4.469 1.117 5.89E+18 0.765 4.50E+18 0.778 Shell Ring #1 1-19 4.469 1.117 5.89E+18 0.765 4.50E+18 0.778 Shell Ring #2 1-20 4.469 1.117 7.49E+18 0.765 5.73E:+18 0.844 SelRing #2 1-21 4.469 1.117 7.49E+18 0.765 5.73E=+18 0.844 Lower DI,D2 4.469 1.117 4.45E+I18 0.765 3.40E+18 0.703 Lower D1,D2 4.469 1.117 4.45E+18 0.765 3.40E+18 0.703 Lower-lnt E1,E2 4.469 1.117 5.98E+18 0.765 4.57E+18 0.782 S Lower-lnt E1,E2 4.469 1.117 5.98E+18 0.765 4.57E+18 0.782 Girth DE 4.469 1.117 5.89E+18 0.765 4.50E+18 0.778 Girth DE 4.469 1.117 5.89E+18 0.765 4.50E=+18 0.778 Girth DE 4.469 1.117 5.89E+18 0.765 4.50E+18 0.778 -* Nozzle N16 4.469 1.117 2.96E+18 0.765 2.26E+18 0.599 o Nozzle N2 4.469 1.117 6.25E+17 0.765 4.78E+17 0.286 (bNozzending) 4.469 1.117 2.96E=+18 0.765 2.26E+18 0.599 Note: The source of redacted proprietaiy information is Reference [13] SREDACTED f{ }j PROPRIETARY INFORMATIONI 41
ATTACHMENT 6 Page 42 of 43 Table 9: DAEC Summary of Nozzle Stress Intensity Factors Feedwater 80.5 64.5 N2 97.4 14.0 Note:. K1 in units of ksi-ino' 42
ATTACHMENT 6 Page 43 of 43 App en dix A DAEC REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Sur'veillance Program Requirements [20], the first surveillance capsule was removed from the DAEC reactor vessel in 1985 after 5.9 EFPY, and the second capsule was removed in 1997 after 14.36 EPPY [21]. The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. DAEC is currently committed to use the BWRVIP ISP and has made a licensing commitment to use the ISP for DAEC during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs and has been approved by the NRC. DAEC committed to use the ISP in place of its existing surveillance programs in the amendments issued by the NRC regarding the implementation of the BWRVIP ISP, dated November 27, 2006 [23]. Under the ISP, a capsule was removed in 2012 after 28 EFPY [14]. DAEC representative surveillance capsule materials were also removed and tested under the Supplemental Surveillance Program (SSP), and were contained in the SSP F capsules [21]. The surveillance capsules contained flux wires for neutron measurement, Charpy V-notch impact test specimens, and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core belttine region. DAEC continues to be a host plant under the ISP [13]. One more DAEC capsule is scheduled to be removed and tested under the ISP in approximately 2027 [21, Table 4-7]. 43}}