ML16271A374
| ML16271A374 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/23/2016 |
| From: | Boggs R NextEra Energy Duane Arnold |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16271A385 | List: |
| References | |
| NG-16-0177 | |
| Download: ML16271A374 (31) | |
Text
Enclosure 1 to NG-16-0177 Non-proprietary version of the NextEra Energy Duane Arnold PTLR Revision 30 pa~es follow
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) -Appendix A Revision 0 NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (EFPY)
Prepared by: Signatures on Proprietary Version Date: _ __
R. Scott Boggs Fleet Programs Engineering Reviewed by: Signatures on Proprietary Version Date-: _ __
Russ Severson Fleet Programs Engineering..
DAEC BWRVIP Engineer Approved by: Signatures on Proprietary Version Date: _ __
Steven P. Brown Director, Engineering Concurred by: Signatures on Proprietary Version Date: _ __
J. Michael Davis Manager, Licensing Page 1of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table of Contents Section Page 1.0 Purpose 3 2.0 Applicability 3 3.0 Methodology 4 4.0 Operating Limits 5 5.0 Discussion 6 6.0 References 12 Figure 1 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 16 EFPY Figure 2 DAEC P-T Curve B (Normal Operation - Core Not Critical), 17 54EFPY Figure 3 DAEC P-T Curve C (Normal Operation - Core Critical), 54 18 EFPY Table 1 DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 19 EFPY Table 2 DAEC P-T Curve B (Normal Operation - Core Not Critical}, 22 54EFPY Table 3 DAEC P-T Curve C (Normal Operation - Core Critical), 54 25 EFPY Table 4 DAEC ART Table for 54 EFPY 28 Table 5 DAEC Summary of Nozzle Stress Intensity Factors 29 Attachment A DAEC Reactor Vessel Material Surveillance Program 30 Page 2 of30
Non Proprietary Version - for NRC Submittal only
- Technical Requirements Manual (TRM) - Appendix A Revision 0 1.0 Purpose The purpose of the Duane Arnold Energy Center (DAEC) Pressure 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 LiT requirements during Recirculation Pump startups;
- 4. RPV bottom head coolant temperature to RPV coolant temperature LiT requirements during Recirculation Pump startups;
This repmi has been prepared in accordance with the requirements of Licensing Topical Report SIR-05-044, Revision 1-A [1].
2.0 Applicability This repmi is applicable to the DAEC RPV for up to 54 Effective Full-Power Years (EFPY).
Note that the 32 EFPY RCS Pressure and Temperature Limit Curves were eliminated from the PTLR as identified in the DAEC response to NRC SRXB-RAI-2 in DAEC Letter NG-16-0042, dated 2-19-2016 (ML16055A145).
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 Page 3of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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 bytheNRC.
- 2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [2], using the RAMA computer code, as documented in Reference [3].
- 3. The adjusted reference temperature (ART)values for the limiting bdtline 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 [Reference 24: NRC approval of Amendment 294, ADAMS Accession No.:ML16180A086].
Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data ofthe RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CPR 50.59 [7], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance. Changes to the above methodologies for preparation of the pressure-temperature (P-T) curves cannot be made without prior NRC approval.
Page 4 of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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.
4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline_ 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 B; and (c) core critical operation, referred to as Curve C.
Complete P-T curves were developed for 54 EFPY for DAEC, as documented in Reference [6].
The DAEC P-T curves for 54 EFPY are provided in Figure 1 through Figure 3, and a tabulation of the curves is included in Table 1 through Table 3. The ART values for the DAEC vessel beltline materials are shown in Table 4 for 54 EFPY, taken from Reference [5].
The resulting P-T curves are based on the geometry, design, and materials information for the DAEC vessels with the following conditions:
- Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figure 1: Curve A):~ 25°F/hour 1
- Normal Operating Heat-up/Cool-down rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating):~ 100°F/hour 2
- 1 Interpreted as the temperature change in any I-hour period is less than or equal to 25°F Page 5of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0
- RPV bottom head coolant temperature to RPV coolant temperature ~ T limit during Recirculation Pump startup:~ 145°F.
- Recirculation loop coolant temperature to RPV coolant temperature ~ T limit during Recirculation Pump startup:~ 50°F.
To address the NRC condition regarding lowest service temperature (LST) in Reference [1], the minimum temperature is set to 74°F, which is equal to the RTNDT,max + 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 LS Ts for the ferritic non-RPV components of the reactor coolant pressure boundary (RCPB).
Review of piping design specifications [1 OJ the maximum impact test temperature is 0°F [page 2 in lOa, b, and c], 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 beltline material is used to adjust the beltline P-T curves to account for irradiation effects. RG 1.99 [4] provides the methods for 2
Interpreted as the temperature change in any I-hour period is less than or equal to 100°F.
Page 6of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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 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 ofRG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 ofRG 1.99 for plates and forgings. However, 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 value of7.49 x 10 18 n/cm 2 at 54 EFPY was obtained from Reference [3] .and was calculated in accordance with RG 1.190 [2]. The fluence value for the limiting. lower-intermediate shell plate (Heat No. B0673-1) is based upon an attenuation factor of 0.765 for a postulated 1/4t flaw. Consequently, the 1/4t fluence for 54 EFPY for the limiting lower-intermediate shell plate is 5.73 x 10 18 n/cm2 for DAEC.
The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. Based on the ART evaluation in Reference [5], the instrumentation (N16) and recirculation inlet (N2) nozzles are located in the beltline region. The N16 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 N16 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 N16 nozzle does not have any significant cycling.
Therefore, the N2 nozzle is considered the limiting nozzle. The peak fluence value for the limiting beltline nozzle is summarized in Table 4 for 54 EFPY. The feedwater (FW) nozzle is considered in the .evaluation of the non-beltline (upper vessel) region P-T limits.
Page 7of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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 1/4t and the 3/4t locations. This 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 1/4t to be less than that at 3/4t for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, 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 curve (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 ifthe pressure test heat-up/cool-down rate limits cannot be maintained.
The initial RTNnT, the chemistry (weight percent copper and nickel), and ART at the 1/4t location for all RPV beltline materials significantly affected by fluence (i.e. fluence > 10 17 n/cm2 for E > 1 MeV) are shown for DAEC in Table 4 for 54 EFPY [5]. Use of initial RTNDT values in the determination of P-T curves for DAEC was approved by the NRG in Reference [22].
Page 8 of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Per Reference [5] and in accordance with Appendix A of Reference [I], the DAEC representative weld and plate surveillance material data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [13, 14].
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 beltline plate (Heat No. B0673-l), 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 (DAI 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 performed, 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 through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients. The resulting stress intensity factors are summarized in Table 5. Detailed infmmation regarding the stress analysis can be found in Reference [ 16], and the calculation of stress intensity factors is described in Reference [ 18, Page 9of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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 thermal shock of 400°F in Reference [ 16]. The stresses were scaled to represent a thermal shock of 459°F; which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions. The stress results for a 459°F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting feedwater nozzle, as a shock of 459°F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 100°F/hr startup transient [12].
Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 100°F/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 BIE/IF methodology presented in Reference
[1] is used* to calculate the thermal stress intensity factor, KIT, 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 [16.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 thermal shock from 500°F to 100°F. 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 (BIE/IF) methodology presented in Reference [1] was used to calculate the pressure stress intensity factor, KIP, by fitting a third order Page 10of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 polynomial equation to the path stress distribution for the pressure load case. The resulting K 1p can be linearly scaled to determine the K1p for various RPV internal pressures.
- An axisymmetric two-dimensional finite element model of the FW nozzle was constructed [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 ASME Code,Section II, Part D, with 2003 Addenda [19]. The properties were taken at a temperature of 300°F, which is approximately the average temperature for the thermal shock transient from 500°F to 100°F. The use of temperature independent material properties is consistent with design basis documents. Use oftemperature-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 the FW nozzle [16] was used to determine a through-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 100°F/hr startup and cooldown [12]. The results of the finite element analysis in [16.b] were scaled to that of a 100°F step change, which conservatively bounds the 100°F/hr 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 [16.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 feed water nozzle, Page 11 of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 which would result in a decrease in the stress coefficients. Therefore, the thermal 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 l .21 is applied to the FW nozzle pressure stress coefficients to obtain the corresponding stress coefficients for the N2 nozzle [18, Section 3.2.2].
6.0 References
- 1. Licensing Topical Report (LTR) BWROG-TP-11-022-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.
Page 12 of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0
- 8. General Electric Report No. GE-NE-A22-00100-08-0I-R2, Revision 2, "Pressure-Temperature Curves for Duane Arnold Energy Center," August 2003. SI File No. DAEC-20Q-201.
- 9. Letter from Mohan C. Thadani (NRC) to Lee Liu (Iowa Electric Light and Power Company),"Duane Arnold Energy Center, License Amendment No. I21 to License No.
DPR-49," May 28, I985, ADAMS Accession No. ML02I890474, SI File No.
I500134.203.
IO. DAEC RCPB Piping Design Specifications, SI File No. I500I34.202:
- a. FPL Document No. APED-A6I-019, GE Document No. 22AI295AD, Revision 3, "Pressure Integrity of Piping and Equipment Pressure Parts -Data Sheet."
- c. FPL Document No. BECH-MI90<GLE>, Revision I, "Piping Class GLE."
- d. FPL Document No. BECH-MI90<HLE>, Revision I, "Piping Class HLE."
. I I. SI Calculation No. I 500134.30 I, Revision 0, "DAEC RPV Vacuum Assessment," June 20I5.
I2. DAEC Nozzle Thermal Cycle Diagrams, SI File No. DAEC-20Q-205:
- a. FPL Drawing No. APED-BI 1-003<2>, GE Drawing No. 135B9990, Sheet 2, Revision I, "Nozzle Thermal Cycles (Recirculation Inlet)."
- b. FPL Drawing No. APED-BI 1-003<4>, GE Drawing No. 135B9990, Sheet 4, Revision 2, "Nozzle Thermal Cycles (Feedwater)."
- c. FPL Drawing No. APED-BI I-003<7>, GE Drawing No. 135B9990, Sheet 7, Revision 0, "Nozzle Thermal Cycles (Instrumentation & Core Diff. Press &
Liquid Control)."
Page 13of30
Non Proprietary Version -for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0
- 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.
- a. SI Calculation No. DAEC-20Q-305, Revision 0, "Feedwater Nozzle Finite Element Model, July 2, 2007.
- b. SI Calculation No. DAEC-20Q-306, 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-Bl l-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: BWR 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 Dari S. Hood, Nuclear Regulatory Commission, dated August 25, 2003, subject: DUANE Page 14of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 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. MC9313). ADAMS Accession No. ML062830019.
- 24. Letter to Mr. Thomas A. Vehec, Vice President, Duane Arnold Energy Center, from Mahesh L. Chawla, Nuclear Regulatory Commission, "DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT TO REVISE AND RELOCATE PRESSURE AND TEMPERATURE LIMIT CURVES TO A PRESSURE AND TEMPERATURE LIMITS REPORT (CAC NO. MF6617)," NRC approval of Amendment 294 and Safety Evaluation and dated July 25, 2016, (ADAMS Accession No.:ML16180A086)
Page 15of30
Non Proprietary Version - for NRC Submittal on ly Technical Requirements Manual (TRM) - Appendix A Revision 0 Figure 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY Curve A - Pressure Test, Composite Curves l
- - Beltline - - - Bottom Head - - Non-Beltline - overall 1300 I I I I
1200 -----
I I
I
,I I I 1100 + - -- - - + - - - ----1*- --fi---+---------+--lk-- - - + - - - - - - - I I I I
1000 ------+-- _L _ _-1-----
I I
I 900 ---rI I
I I
Qi BOO - -- -
'iii 1
~ I
~11 700 I
_L_
> I 15 I t: I I
~"' 600 - T - --+-----
l I
--,-I I
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1 I
I
- -[- -
200 --r-----------r---m-------i----,-------i----r================;-1 Minimum RPV 100 - - - - - + - - --to'--- - - - + - - -----i._---~--__J
=
Pressure -14.7 psig Minimum Bolt-Up o + - - - - - - - + - --fl+-- - - + - - - - + - - - - - - - + - -* Temperature = 7 4 °F
-100 -'-----~------+----~---~------+---~
0 50 100 150 200 250 300 M inimum Reactor Vessel Metal Temperature (°F)
Page 16of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Figure 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY Curve B - Core Not Critical, Composite Curves
- - Beltl ine - - - Bottom Head - - Non-Beltline - ove rall 1200 -----------
,f I
L ______ *ff+-- - - - + - - - - -
+-'----+-----__,_ _,_ ,~l--+-*r'- ----;---1----+------!
1 1100 I
I 1000 I
-,- - - ---+--
I 900 - - - - - +_..- - - - - - + - -IM*
I I
I QO - - - - - +-+If-
- ~~:==-
"ill
~ I
~QJ I
> I I
~
t:
ru QJ a: 600
- =
t----
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--/---+----*-
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I
- e
- J QJ
- I Ill QJ 500 r----
I
- I I
1
~ 400 - -i - -- + - - --.-- + --f.olr-- - t - - - - - + - - - - - 1 I
I I
300 - -- - +---/- - ~ - ----....,..._
- I I
100 i
_j__ ---
1 Minim um Bolt -Up o _,_I_ _ _ _ _ , __ _ _ __ - + - - - - - - + - - - - - + --t =
Tem perat ure 7 4 °F
-100 - ' - - - - - - - ' - - - - - - ' - - - - - - - ' - - - - - - ' - - - - - - - ' - - - - - - - '
0 50 100 150 200 250 300 Minimum Reactor Vessel Met al Te mperature (°F)
Page 17 of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 F igure 3: DAEC P-T C urve C (Normal Opera tion - Core C ritical), 54 EFPY Curve C - Core Critical, Composite Curves
- - Beltline - - - Bottom Head - - Non-Beltline - overall 1WO -.-----,------,-------,---,, --,,.-----.,.----,.~--,
I I
I 1200 *----*-*---
I I'
1100 +-----+-----+-----+~ ,*----<*+-----+--------<
I I I
1000 --+----- - - - - - + --!ft-I 900 --, I
- - - - -f#i-- - - - -
I "iiD
- w; soo -~--i-----* ->t--
I
.!:!: I I I
~GI - -.,- I I 700 - T-
!5 ti I I ' l t;."'
.E 600 -----+--
I
,,----+--- -
"i§ I
- J 500 *- - - + - - - -
I f
~
~ 400 +-----+---,
, - ~-+----+--*1--~'1'---*~------j I I I
300 - -I-I I
I 200 + - - - - - - + - - -1 I
I Minimum RPV I Pressu re = -14.7 psig 100 M inimum Bolt-Up o + - - - - - + --+Ii;--- - + - - - - - + - - - - - + -* Temperatu re= 74°F
-100 ...____ _ ___,__ _ _ __,___ _ ___,__ _ _ __,___ _ ___,__ _ ___,
0 50 100 150 200 250 300 Minimum Reactor Vessel Metal Temperature (°F)
Page 18 of3 0
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 1: DAEC P-T C urve A (Hydrosta tic Pressu re and Leak Test), 54 EFPY Beltline Region
.: Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF 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 1129.2 211.0 1177.4 213.4 1225.6 215 .7 1273.8 217.9 1322.0 Page 19of3 0
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY (cont.)
Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure 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 1213.3 129.0 1263.3 131.3 1313.2 Page 20of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 1: DAEC P-T Curve A (Hydrostatic Pressure and Leak Test), 54 EFPY (cont.)
Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF 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 Page 21of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure OF psi 74.0 0.0 74.0 117.0 109.9 166.8 130.6 216.7 145.2 266.6 156.5 316.4 165.7 366.3 173 .4 416.2 180.2 466.0 186.1 515.9 191.4 565.7 196.2 615.6 200.5 665.5 204.6 715 .3 208.3 765.2 211.7 815 .1 215.0 864.9 218.0 914.8 220.9 964.6 223.6 1014.5 226.2 1064.4 228.6 1114.2 230.9 1164.1 233 .2 1214.0 235.3 1263.8 237.3 1313.7 Page 22of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY (cont.)
Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure OF 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 Page 23of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 2: DAEC P-T Curve B (Normal Operation - Core Not Critical), 54 EFPY (cont.)
Bottom Head Reg ion
- Curve B - Core Not Critical P-T Cu rve P-T Curve Temperature Pressure OF 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 Page 24of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF psi 74.0 0.0 74.0 90.9 133 .7 139.9 160.2 189.0 177.4 238.1 190.2 287 .1 200.3 336.2 208.8 385 .3 216.0 434.4 222.3 483.4 227.9 532 .5 233.0 581.6 237.5 630.6 241.7 679 .7 245.6 728.8 249.2 777 .9 252 .5 826.9 255 .7 876.0 258.6 925 .1 261.4 974.1 264.0 1023.2 266.6 1072.3 268.9 1121.4 271.2 1170.4 273.4 1219.5 275 .5 1268.6 277 .5 1317.6 Page 25of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY (cont.)
Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF psi 74.0 0.0 74.0 93.5 97.7 137.3 112.6 181.2 123 .5 225 .0 132.1 268.8 139.2 312 .6 174.0 312.6 174.0 667.3 177.3 717 .1 180.3 766.8 183.2 816.6 185.8 866.4 188.4 916.1 190.8 965 .9 193 .1 1015.6 195.2 1065.4 197.3 1115.2 199.3 1164.9 201.2 1214.7 203.0 1264.4 204.8 1314.2 206.5 1364.0 Page 26of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 3: DAEC P-T Curve C (Normal Operation - Core Critical), 54 EFPY (cont.)
Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure OF 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 Page 27of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Ta ble 4: DAEC ART Table for 54 EFPY Adjustments For 1/4t Chemistry Margin Initial Chemistry Description ID .iRT NOT Terms ART Heat No. Lot No. RTNDT Factor No. Cu (oF) (oF)
(wt Ni O",i O";
%) (wt%) (oF) (oF) (oF) (oF)
Shell Ring #1 1-18 C6439-2 - 40 0.09 0.51 58 .0 45 .1 17.0 0.0 119.1
~I Shell Ring #1 1-19 80402-1 - 40 0.13 0.47 87.1 67.8 17.0 0.0 141 .8 Shell Ring #2 1-20 80436-2 - 10 0.15 0.64 111 .0 93 .7 17.0 0.0 137.7 Shell Rina #2 1-21 80673-1 0.15 0.65 a, o, c, a, e, r
- 10 (( 124.5 8.5 0.0 1151.5 Lower D1 ,D2 432Z4521 8020A27A -50 0.01 0.98 20.0 14.1 7.0 0.0 -21 .9 Lower D1 ,D2 432Z0471 8003A27A -50 0.03 0.91 41 .0 28 .8 14.4 0.0 7.7 Lower- Int E1 ,E2 432Z4521 8020A27A -50 0.01 0.98 20 .0 15.6 7.8 0.0 -18.7 ii Lower-Int Girth Girth E1 ,E2 DE DE 432Z0471 09L853 07L669 8 003A27A L017A27A K004A27A
-50
-50
-50 0.03 0.03 0.03 0.91 0.88 1.02 41 .0 41.0 41 .0 32 .1 31.9 31.9 16.0 16.0 16.0 0.0 00 0.0 14.1 13.8 13.8 Girth DE CTY538 A027A27A -50 0.03 0.83 41 .0 31.9 16.0 0.0 13.8 VJ
.!!! Nozzle N16 - Q2Q5VW - 40 0.18 0.85 141.8 85 .0 17.0 0.0 159.0 N
N 0 Nozzle N2 - Q2Q6VW - 40 0.18 0.84 141 .6 40.4 17.0 114.4 z 0.0 412Z051 K910A27A
~I "OVJ "N
N-N2/N16 Nozzle Welds - 08R4818 659T568 K904A27A H721A27A -50 0.03 1.00 41 .0 24 .6 12.3 0.0 -0.8 0 ~ z ::: (bounding) 661Y494 F927A27A 661Y439 E916A27A Fluence Attenuatio n, Fluence Factor, Wall Thickness lin.\ at ID 1/4t Fluence@ 1/4t FF 2 e*0.24x 2 f<0 .28-0.10109 f) Location Full 1/4t (n/cm ) (n/cm ) Shell Ring #1 1-18 4.469 1.117 5.89E+18 0.765 4.50E+18 0.778
~I 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 Shell Rina #2 1-21 4.469 1.1 17 7.49E+ 18 0.765 5.73E+18 0.844 Lower D1 ,D2 4.469 1.117 4.45E+18 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-Int E1 ,E2 4.469 1.117 5.98E+18 0.765 4.57E+18 0.782 ii Lower-Int Girth Girth E1 ,E2 DE DE 4.469 4.469 4.469 1.117 1.117 1.117 5.98E+18 5.89E+18 5.89E+18 0.765 0.765 0.765 4.57E+18 4.50 E+1 8 4.50E+1 8 0.782 0.778 0.778 Girth DE 4.469 1.117 5.89E+1 8 0.765 4.50E+ 18 0.778 VJ ~ "N Nozzle N16 4.469 1.117 2.96E+18 0.765 2.26E+18 0.599 N
0 Nozzle N2 4.469 1.117 6.25E+17 0.765 4.78E+17 0.286 z
~,VJ "N "O Nozzle N16 N-0 ~ - 4.469 1.117 2.96E+18 0.765 2.26E+18 0.599 (bounding) z :::
Note: The source of proprietary-marked information is Reference [13] EPRI ((x.xx}} PROPRIETARY INFORMATION Page 28of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Table 5: DAEC Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Kip-app Thermal, K11 Feed water 80.5 64.5 N2 97.4 14.0 5 Note : K1 in units of ksi-in° Page 29of30
Non Proprietary Version - for NRC Submittal only Technical Requirements Manual (TRM) - Appendix A Revision 0 Attachment A DAEC REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with l 0 CFR 50, Appendix H, Reactor Vessel Material Survei llance Program Requirements [20] , the first surveillance capsule was removed from the DAEC reactor vessel in 1985 after 5.9 EFPY, and the second capsu le was removed in J 997 after 14.36 EFPY [21]. The surveil lance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniax ial tensile test specimens fabricated using materials from the vessel materials with in 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 l 0 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 Survei llance 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 beltline region. DAEC continues to be a host plant under the ISP [13]. One more DAEC capsule is schedu led to be removed and tested under the ISP in approximately 2027 [21, Table 4-7]. Page 30 of 30}}