NL-17-1916, Pressure and Temperature Limits Report
| ML17331B320 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/27/2017 |
| From: | Hutto J Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-1916 | |
| Download: ML17331B320 (43) | |
Text
t. Southern Nuclear J. J. Hutto Regulatory AHairs Director 40 lnvem~ss Center Paii,wJy Post Office Box 1295 Rim1ingham, AL 35242 205 992 5872 tel 205 992 760 I fax NOV 2 7 2017 jjlmuo@southemco.com Docket Nos.: 50-321 NL-17-1916 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Unit 1 Pressure and Temperature Limits Report for Hatch Ladies and Gentlemen:
In accordance with Technical Specification 5.6.7.c, Southern Nuclear Operating Company submits the enclosed Hatch Nuclear Plant Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 1. This revision adds references to three engineering evaluations. No changes were made to the pressure-temperature curves.
This letter contains no NRC commitments If you have any questions, please contact Ken McElroy at 205.992.7369.
Respectfully submitted, J. J. Hutto Regulatory Affairs Director JJH/PDB/CBG
Enclosure:
Unit 1 Pressure and Temperature Limits Report, Revision 1
U.S. Nuclear Regulatory Commission NL-17-1916 Page 2 Cc: Regional Administrator, Region II NRR Project Manager- Hatch Senior Resident Inspector- Hatch RTYPE: CHA02.004
Edwin I. Hatch Nuclear Plant Unit 1 and 2 Unit 1 Pressure and Temperature Limits Report for Hatch Enclosure Unit 1 Pressure and Temperature Limits Report, Revision 1
Southern Nuclear Operating Co.
Hatch Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR) for 38 and 49.3 Effective Full-Power Years (EFPY)
Revision 1
Hatch Unit 1 PTLR Revision 1 Page 2 of 40 Table of Contents Section Page 1.0 Purpose 4 2.0 Applicability 4 3.0 Methodology 5 4.0 Operating Limits 6 5.0 Discussion 7 6.0 References 12 Figure 1 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY 15 Figure 2 HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY 16 Figure 3 HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY 17 Figure 4 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY 18 Figure 5 HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY 19 Figure 6 HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY 20
Hatch Unit 1 PTLR Revision 1 Page 3 of 40 Section Page Table 1 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 21 EFPY Table 2 HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 23 EFPY Table 3 HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY 26 Table 4 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 29 EFPY Table 5 HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 31 EFPY Table 6 HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY 34 Table 7 Hatch Unit 1 ART Table for 38 EFPY 37 Table 8 Hatch Unit 1 ART Table for 49.3 EFPY 38 Table 9 Hatch Unit 1 Summary ofNozzle Stress Intensity Factors 39 Appendix A Hatch Unit 1 Reactor Vessel Materials Surveillance Program 40
Hatch Unit 1 PTLR Revision 1 Page 4 of 40 1.0 Purpose The purpose ofthe Hatch Nuclear Plant, Unit 1 (HNP-1) 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;
This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision 1 [1] and 0900876.401, Revision 0 [2].
2.0 Applicability This report is applicable to the HNP-1 RPV for up to 38 and 49.3 Effective Full-Power Years (EFPY) [3].
The following HNP-1 Technical Specification (TS) is affected by the information contained in this report:
- Limiting Condition for Operation and Surveillance Requirement 3.4.9 ("RCS Pressure and Temperature (PIT) Limits")
Hatch Unit 1 PTLR Revision 1 Page 5 of40 3.0 Methodology The limits in this report were derived as follows:
- 1. The methodology used is in accordance with Reference [1] and Reference [2], which have been approved by the NRC.
- 2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [4], using the RAMA computer code, as documented in Reference [5].
- 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) [6], as documented in Reference [7].
- 4. The pressure and temperature limits were calculated in accordance with Reference [1],
"Pressure- Temperature Limits Report Methodology for Boiling Water Reactors," June 2013, as documented in Reference [8].
- 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, 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 or other plant design assumption modifications in the UFSAR, cannot
Hatch Unit 1 PTLR Revision 1 Page 6 of40 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 38 EFPY and 49.3 EFPY for HNP-1 as documented in Reference [8]. The HNP-1 P-T curves for 38 EFPY are provided in Figures 1 through 3, and a tabulation of the overall composite curves (by region) is included in Tables 1 through 3. The HNP-1 P-T curves for 49.3 EFPY are provided in Figures 4 through 6, and a tabulation of the overall composite curves (by region) is included in Tables 4 through 6. The adjusted reference temperature (ART) values for the HNP-1 vessel beltline materials are shown in Table 7 for 38 EFPY and Table 8 for 49.3 EFPY taken from Reference [7]. The resulting P-T curves are based on the geometry, design and materials information for the HNP-1 vessel with the following conditions:
- Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figures 1 and 4:
Curve A): S 25"F/hour 1 [8].
- Normal Operating Heat-up/Cool-down rate limit (Figures 2 and 5: Curve B - non-nuclear heating, and Figures 3 and 6: Curve C- nuclear heating): S 100"F/hour2 [8].
1 Interpreted 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.
Hatch Unit 1 PTLR Revision 1 Page 7 of 40
- Minimum bolt-up temperature limit 2: 76"F [8].
To address the NRC condition regarding lowest service temperature in Reference [1], the minimum temperature is set to 76 °F, which is equal to the RTNDT,rnax + 60 °F, for all curves.
This value is consistent with the previous minimum temperature limits developed in [9], and is higher than previous minimum bolt-up specified in [10].
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. 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 RPV pressure is -14.7 psig.
5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1. 99 [6] 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 HNP-1 vessel plate, weld, and forging materials [7]. The Cu and Ni values were used with Table 1 ofRG 1.99 [6] to determine a chemistry factor (CF) per Paragraph 1.1 ofRG 1.99 for welds.
The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 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.
The peak RPV ID fluence value of2.43 x 10 18 n/cm 2 at 38 EFPY was developed in Reference [7]
based on linear interpolation between reported fluence values for 28.4 EFPY and 49.3 EFPY from Reference [5], which were calculated in accordance with RG 1.190 [4]. The peak RPV ID fluence value of3.08 x 10 18 n/cm2 at 49.3 EFPY was obtained from Reference [5] and was
Hatch Unit 1 PTLR Revision 1 Page 8 of 40 calculated in accordance with RG 1.190. These fluence values apply to the limiting beltline lower intermediate shell plate (Heat No. C4114-2). The fluence values for the limiting lower intermediate shell plate are based upon an attenuation factor of0.724 for a postulated 1/4T flaw.
As a result, the 114T fluence for 38 EFPY and 49.3 EFPY for the limiting lower intermediate shell plate are 1.76 x 10 18 n!cm2 and 2.23 x 10 18 n!cm2 , respectively, for HNP-1.
The water level instrument (WLI) nozzle is located in the lower intermediate shell beltline plates
[8]. The limiting fluence values are as described in the paragraph above. Based on the ART evaluation in Reference [7], the recirculation inlet and outlet nozzles do not exist in the beltline reg1on.
The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 114T (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 stresses at the 114T location. This approach is conservative because irradiation effects cause the allowable toughness at the 114T 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"Fihr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level AlB RPV thermal transients. For the hydrostatic pressure and leak test curve (Curve A), a coolant heat-up and cool-down temperature rate of~ 25"Fihr must be
Hatch Unit 1 PTLR Revision 1 Page 9 of 40 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 RTNoT, 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 forE> 1MeV) are shown in Table 7 for 38 EFPY and Table 8 for 49.3 EFPY [7]. The initial RTNDT values shown in Tables 7 and 8 have been previously approved for use by the NRC per Reference [19].
Per Reference [7] and in accordance with Appendix A of Reference [1], the HNP-1 representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [12]. For the plate material, Procedure 1 from Appendix A of [1] was used. The fitted CF for the limiting plate (Heat No. C4114-2), which is based on credible surveillance data, in the HNP-1 vessel bounds the RG 1.99 CF [12]. Therefore, the fitted CF is used for the limiting beltline plate. In addition, an archival plate heat (Heat No. C3985-2) from the HNP-1 vessel was included in the Supplemental Surveillance program (SSP) and irradiated data from SSP Capsules H and C are provided in Reference [12]. These data are also determined to be credible, and, consequently, a reduced margin term is used for this material as well. For the weld material, Procedure 2 from Appendix A of[1] was used. The HNP-1 representative weld material (20291) is contained in the Cooper and SSP Capsule C capsules [7, 12]. Reference [12] contains surveillance capsule test results for the HNP-1 representative weld material; however, since the material heats for the HNP-1 limiting weld material and representative surveillance capsule weld material do not match, the CF calculated using the RG 1.99 [6] tables is used.
Hatch Unit 1 PTLR Revision 1 Page 10 of 40 The ANSYS finite element computer program was used to develop the stress distributions through the feedwater (FW) nozzle [13]. These stress distributions were used in the determination of the stress intensity factors for the FW nozzles [14]. At the time the analyses were performed, the ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B
[15] Quality Assurance Program for nuclear quality-related work.
The plant-specific HNP-1 FW nozzle analysis was performed to determine stress intensity factors due to through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients [14]. Pressure and thermal stress distributions were taken from Reference [13]. Detailed information regarding the analysis can be found in References [13 , 14].
The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle [14]:
- With respect to thermal stresses, the thermal shock which represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions was analyzed [13]. The thermal stress distribution, corresponding to the limiting time point presented in [13], along a linear path through the nozzle comer is used [14]. Leakage is considered in the heat transfer calculations [13]. The thermal down shock of 450°F produces the highest tensile stresses at the 1/4T location. The BIEIIF methodology presented in the SI P-T Curve LTR [1] is used to calculate the thermal stress intensity, KIT, due to the thermal shock by fitting a third order polynomial equation to the path stress distribution for the thermal shock load case [14]. Because operation is along the saturation curve, the resulting KIT can be linearly scaled to determine the KIT to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant and could potentially result in a near zero stress distribution. Therefore, a minimum KIT is calculated based on the thermal ramp of 100°Fihr, which is associated with the shutdown transient [14]. The resulting combination of the thermal down shock and thermal ramp
Hatch Unit 1 PTLR Revision 1 Page 11 of 40 KIT values represent the bounding thermal stress intensity factors in the FW nozzle associated with the P-T curves for the non-beltline region.
- Boundary conditions and heat transfer coefficients used for the thermal stress analysis are as described Reference [13a]. Overall heat transfer coefficients representative of a triple sleeve sparger with Seal No. 1 failed were applied [13a].
- With respect to pressure stresses, a unit pressure of 1000 psig was applied to the internal surfaces of the finite element model (FEM) [13]. The pressure stress distribution was taken along the same path as the thermal stress distribution. Recognizing that the Reference [13] evaluation was performed using a 2-D axi-symmetric finite element model (FEM) and that it is known that the stress intensification caused by the nozzle geometry is under predicted in a 2-D axi-symmetric representation of the nozzle, a correction factor was applied to the stresses obtained from the 2-D axi-symmetric FEM as described in Reference [14]. The BIE/IF methodology presented in the SI P-T Curve LTR [1] is used to calculate the pressure stress intensity factor, KIP, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIP can be linearly scaled to determine the KIP for various RPV internal pressures.
- Material properties were taken from the HNP-1 code of construction [16]. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.
The following summarizes the development of the thermal and pressure stress intensity factors for the CDP nozzle [14]:
- The KIT term is calculated using the ASME XI, Non-mandatory Appendix G, Paragraph G-2214.3 [17] methodology for a heat-up/cool-down rate of 100 "F/hr as described in Reference [14].
- The KIP is calculated [14] using the WRC 175 methodology [18].
Hatch Unit 1 PTLR Revision 1 Page 12 of 40 6.0 References
- 1. BWROG-TP-11-022-A, Revision 1 (SIR-05-044, Revision 1-A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 2013.
- 2. BWROG-TP-11-023-A, Revision 0 (0900876.401, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations", dated May 2013.
- 3. Design Input Requests:
1001527.201.
1400365.200.
- 4. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001.
- 5. Transware Enterprises Inc. Report No. SNC-HA1-002-R-001 Revision 0, "Edwin I.
Hatch Unit 1 Fluence Evaluation at End of Cycle 25 and 49.3 EFPY.".
- 6. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel materials", May 1988.
- 7. Structural Integrity Associates Calculation No. 1001527.301, Revision 1, "Hatch Unit 1 RPV Material Summary and ART Calculation", July 2014.
- 8. Structural Integrity Associates Calculation No. 1001527.304, Revision 2, "Hatch Unit 1 P-T Curve Calculation for 38 and 49.3 EFPY", August 2014.
- 9. General Electric Document No. GE-NE-B1100827-00-01, "Plant Hatch Units 1 & 2 RPV Pressure Temperature Limits License Renewal Evaluation," March 1999.
Hatch Unit 1 PTLR Revision 1 Page 13 of 40
- 10. NRC Docket No. 50-321, "Edwin I. Hatch Nuclear Plant Unit No. 1, Amendment to Facility Operating License," Amendment No. 59, License No. DPR-57, August 1978, ADAMS Accession No. ML012950436.
- 12. BWRVIP-135, Revision 2: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2009.
1020231. EPRI PROPRIETARY INFORMATION. SI File No. BWRVIP-01-335P.
- 13. Hatch Unit 2 NUREG-0619 Evaluations:
- a. Liffengren, D. J., et al., "Edwin I. Hatch Nuclear Power Station, Unit 1 Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-0619,"
NEDE-30238, DRF-30238, August 1983, General Electric Company. SI File No.
1001527.210.
- b. Bothne, D., "Power Uprate Evaluation Report for Edwin I. Hatch Unit 1, Feedwater Nozzle NUREG-0619 Fracture Mechanics Analysis for Extended Power Uprate Conditions," GE-NE-B13-01869-065-01, July 1997, General Electric Company. SI File No. 1001527.210
- 14. Structural Integrity Associates Calculation No. 1001527.303, Revision 0, "Feedwater Water Level Instrument, and Core DP Nozzle Fracture Mechanics Evaluation for Hatch Unit 1 and Unit 2 Pressure-Temperature Limit Curve Development", December 2011
- 15. U. S. Code ofFederal Regulations, Title 10, Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".
- 16. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, 1965 Ed. Winter 1966 Addenda.
- 17. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Non-mandatory
Hatch Unit 1 PTLR Revision 1 Page 14 of 40 Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 2001 Ed.
through 2003 Addenda.
- 18. PVRC Recommendations on Toughness Requirements for Ferritic Materials. WRC Bulletin 175. August 1972.
- 19. NUREG-1803, "Safety Evaluation Report Related to the License Renewal ofthe Edwin I.
Hatch Nuclear Plant, Units 1 and 2," December 2001.
- 20. General Electric Report No. GE-NE-B1100691-01R1, "Plant Hatch Unit 1 RPV Surveillance Materials Testing and Analysis," March 1997. SI File No. 1001527.202.
- 21. Letter from L. N. Olshan (NRC) to H. L. Sumner (SNC), "Edwin I. Hatch Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC NOS. MB6106 and MB6107)", March 10, 2003.
- 22. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan EPRI Product 1025144, October 2012.
- 23. BWRVIP-308NP, BWR Vessel and Internals Project, Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule, EPRI Product 3001040553, July 2017.
- 24. Request for Engineering Review SNC884884, Hatch Unit 1 Fluence Evaluation for RPV, July 2017.
- 25. Structural Integrity Report 1001527.403 Revision 0, Notification of Pressure-Temperature (P-T) Curve Error Identified in SI Corrective Action Report (CAR) No.16-023, July 2017.
Hatch Unit 1 PTLR Revision 1 Page 15 of 40 Figure 1: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY Curve A - Pressure Test, Composite Curves
--Beltline ----Bottom Head - - Non-Beltline - overall iJ 1300 -
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Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 1 Page 16 of 40 Figure 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY Curve B - Core Not Critical, Composite Curves
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~ l lO liO 2/0 2 i0 3 ()
Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 1 Page 17 of40 Figure 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY Curve C- Core Critical, Composite Curves
-Beltline ----Bottom Head - - Non-Bel tline - overall 1300
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so 1l0 1i0 2~ 2 i0 310 I
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Hatch Unit 1 PTLR Revision 1 Page 18 of 40 Figure 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY Curve A- Pressure Test, Composite Curves
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50 1l0 qo 200 2~ 3 l0
_L r Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 1 Page 19 of 40 Figure 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY Curve B- Core Not Critical, Composite Curves
-Beltline ----Bottom Head - - Non-Beltline - overall 1300 . - - - - - - - - , -
-n.-----~ ---.---~-r------
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Temperature =76°F Minimum RPV Pressure = -14.7 psig 0 +--------1
'0 110 2.)0 Minimum Reactor Vessel Metal Temperature (°F)
Hatch Unit 1 PTLR Revision 1 Page 20 of 40 Figure 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY Curve C- Core Critical, Composite Curves
--Beltline ----Bottom Head - - Non-Beltline - overall 1300 I ,,
1200 ------t-----+ - - --1i--!!-----t----':.__-
1
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Minimum RPV Pressure =-14.7 psig o ~---~--1~~----+------~~====~======~
9 50 1:0 t lo 2 i0 3:0 I Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 1 Page 21 of 40 Table 1: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 365.2 98.8 415.1 114.4 465.0 126.3 514.9 135.9 564.8 144.0 614.7 150.9 664.6 157.0 714.5 162.4 764.5 167.3 814.4 171.8 864.3 175.8 914.2 179.6 964.1 183.1 1014.0 186.4 1063.9 189.5 1113.8 192.4 1163.7 195.2 1213.6 197.8 1263.5 200.2 1313.5 202.6 1363.4
Hatch Unit 1 PTLR Revision 1 Page 22 of 40 Table 1: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY (continued)
Bottom Head Region Curve A - Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 1226.1 78.7 1274.2 81.2 1322.4 83.6 1370.5 Non-Beltline Region Curve A - Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 312.6 106.0 312.6 106.0 934.2 109.5 982.6 112.7 1031.0 115.7 1079.3 118.6 1127.7 121.3 1176.1 123.9 1224.4 126.3 1272.8 128.6 1321.2
Hatch Unit 1 PTLR Revision 1 Page 23 of 40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY Beltline Region Curve B -Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 144.9 104.2 193.8 122.1 242.7 135.2 291.6 145.6 340.5 154.2 389.4 161.6 438.3 168.0 487.2 173.6 536.1 178.7 585.0 183.4 633.9 187.6 682.8 191.5 731.7 195.1 780.6 198.5 829.5 201.6 878.4 204.6 927.3 207.4 976.2 210.1 1025.1 212.6 1074.0 215.0 1122.9 217.3 1171.8 219.5 1220.7 221.6 1269.6 223.6 1318.5
Hatch Unit 1 PTLR Revision 1 Page 24 of40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY (continued)
Bottom Head Region Curve B - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 813.9 79.6 863.8 83.0 913 .8 86.2 963.7 89.2 1013.6 92.0 1063.6 94.7 1113.5 97.3 1163.5 99.7 1213.4 102.0 1263.3 104.2 1313.3
Hatch Unit 1 PTLR Revision 1 Page 25 of 40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY (continued)
Non-Beltline Region Curve B - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 198.2 84.6 236.4 91.8 274.5 97.9 312.6 136.0 312.6 136.0 724.2 139.0 773.5 141.8 822.9 144.4 872.2 146.8 921.5 149.2 970.9 151.4 1020.2 153.6 1069.6 155.6 1118.9 157.6 1168.3 159.4 1217.6 161.2 1266.9 163.0 1316.3
Hatch Unit 1 PTLR Revision 1 Page 26 of 40 Table 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 109.3 125.1 157.8 149.3 206.2 165.6 254.7 177.9 303.1 187.7 351.6 195.9 400.0 203.0 448.5 209.1 497.0 214.6 545.4 219.6 593.9 224.1 642.3 228.2 690.8 232.1 739.2 235.6 787.7 238.9 836.1 242.0 884.6 245.0 933.1 247.7 981.5 250.3 1030.0 252.8 1078.4 255.2 1126.9 257.5 1175.3 259.6 1223.8 261.7 1272.3 263.7 1320.7
Hatch Unit 1 PTLR Revision 1 Page 27 of 40 Table 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 450.5 83.6 498.8 90.1 547.2 95.9 595.6 101.1 643.9 105.8 692 .3 110.1 740.7 114.1 789.1 117.8 837.4 121.2 885.8 124.4 934.2 127.4 982.5 130.2 1030.9 132.9 1079.3 135.4 1127.7 137.9 1176.0 140.2 1224.4 142.4 1272.8 144.5 1321.1
Hatch Unit 1 PTLR Revision 1 Page 28 of 40 Table 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure OF psi 76.0 0.0 76.0 97.6 98.0 140.6 112.1 183.6 122.6 226.6 130.9 269.6 137.9 312.6 202.0 312.6 202.0 1563.0
Hatch Unit 1 PTLR Revision 1 Page 29 of40 Table 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY Beltline Region Curve A - Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 345.8 103.3 394.5 120.9 443.2 133.9 491.9 144.2 540.6 152.7 589.3 160.0 638.0 166.4 686.6 172.0 735.3 177.1 784.0 181.7 832.7 185.9 881.4 189.8 930.1 193.4 978.8 196.7 1027.4 199.9 1076.1 202.9 1124.8 205.7 1173.5 208.3 1222.2 210.8 1270.9 213.2 1319.6
Hatch Unit 1 PTLR Revision 1 Page 30 of40 Table 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY (continued)
Bottom Head Region Curve A - Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 1226.1 78.7 1274.2 81.2 1322.4 83.6 1370.5 Non-Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 312.6 106.0 312.6 106.0 934.2 109.5 982.6 112.7 1031.0 115.7 1079.3 118.6 1127.7 121.3 1176.1 123.9 1224.4 126.3 1272.8 128.6 1321.2
Hatch Unit 1 PTLR Revision 1 Page 31 of 40 Table 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY Beltline Region
, Curve B - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 130.4 110.3 179.8 130.4 229.2 144.7 278.6 155.9 328.0 164.9 377.4 172.6 426.8 179.3 476.2 185.2 525.6 190.4 575.0 195.2 624.4 199.5 673.8 203.5 723.2 207.2 772.6 210.7 822.0 213.9 871.4 216.9 920.8 219.8 970.2 222.5 1019.6 225.1 1069.0 227.5 1118.4 229.8 1167.8 232.0 1217.2 234.2 1266.6 236.2 1316.0
Hatch Unit 1 PTLR Revision 1 Page 32 of40 Table 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY (continued)
Bottom Head Region Curve B - Core Not Critical P-T Curve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 813.9 79.6 863.8 83.0 913.8 86.2 963.7 89.2 1013.6 92.0 1063.6 94.7 1113.5 97.3 1163.5 99.7 1213.4 102.0 1263.3 104.2 1313.3
Hatch Unit 1 PTLR Revision 1 Page 33 of 40 Table 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY (continued)
Non-Beltline Region Curve 8 - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 198.2 84.6 236.4 91.8 274.5 97.9 312.6 136.0 312.6 136.0 724.2 139.0 773.5 141.8 822.9 144.4 872.2 146.8 921.5 149.2 970.9 151.4 1020.2 153.6 1069.6 155.6 1118.9 157.6 1168.3 159.4 1217.6 161.2 1266.9 163.0 1316.3
Hatch Unit 1 PTLR Revision 1 Page 34 of 40 Table 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 102.8 133.5 151.5 159.6 200.1 176.6 248.8 189.3 297.5 199.4 346.1 207.8 394.8 215.0 443.5 221.3 492.2 226.9 540.8 231.9 589.5 236.4 638.2 240.6 686.9 244.5 735 .5 248.1 784.2 251.4 832.9 254.5 881.6 257.5 930.2 260.3 978.9 262.9 1027.6 265.4 1076.3 267.8 1124.9 270.1 1173.6 272.3 1222.3 274.3 1271.0 276.3 1319.6
Hatch Unit 1 PTLR Revision 1 Page 35 of 40 Table 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 76.0 0.0 76.0 450.5 83.6 498.8 90.1 547.2 95.9 595.6 101.1 643.9 105.8 692 .3 110.1 740.7 114.1 789.1 117.8 837.4 121.2 885.8 124.4 934.2 127.4 982.5 130.2 1030.9 132.9 1079.3 135.4 1127.7 137.9 1176.0 140.2 1224.4 142.4 1272.8 144.5 1321.1
Hatch Unit 1 PTLR Revision 1 Page 36 of40 Table 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure OF psi 76.0 0.0 76.0 97.6 98.0 140.6 112.1 183.6 122.6 226.6 130.9 269.6 137.9 312.6 217.0 312.6 217.0 1563.0
Hatch Unit 1 PTLR Revision 1 Page 37 of40 Table 7: Hatch Unit 1 ART Table for 38 EFPY Adjustments for 1/4t Heat No./ Chemistry Chemistry Description Code No. Flux Lot No. Initial RTHDrrf) t.RTNoT Margin Tenns , ARTNDT Flux Type FactorrF)
CU (wt%) NJ(wt%) rFl ar rFl all rFl rFl Lower Shell #1 G-4805-1 C4112-1 - 8 0.13 0.64 92 44.8 0.0 17.0 86.8 lower Shell #2 G-4805-2 C4112-2 - 10 0.13 0.64 92 44.8 0.0 17.0 88.8
~I Lower Shell#3 G-4805-3 C4149-1 -10 0.14 0.57 99 48.0 0.0 17.0 72.0 Lower-tnt. Shell #1 G-4803-7 C4337-1 - -20 0.17 0.62 128 68.7 0.0 17.0 82.7 Lower-tnt Shell #2 G-4804-1 C3985-2 - -20 0.11 0.60 65 34.9 0.0 8.5 31.9 Lower-tnt Shell #3 G-4804-2 C4114-2 - -20 0.12 0.70 221 119.3 0.0 8.5 116.3 Adjustments for 1/4t Heat No./ Chemistry Chemistry DeiCI'IpUon Code No. Flux Lot No. Initial RTNoTrFl t.RTNoT Margin Tenns ARTNDT Flux Type FactorrFl cu (wt%) Nl (wt%) rFI arrf) all rFl rFl Lower Long. Weld 1-307 13253/1092 3791 -50 0.221 0.732 189 91 .5 0.0 28.0 97.5
!I lower Int. long Weld #1 1-308 1P2809/1092 3854 -50 0.270 0.735 206 89.8 0.0 28.0 95.8 Lower Int. Long Weld #2 1-308 1P2815/1092 3854 -50 0.316 0.724 219 95.5 0.0 28.0 101 .5 Lower - lower Int. Girth Weld #1 1-313 90099/0091 3977 -10 0.197 0.060 91 45.7 0.0 22.9 81 .4 Lower - Lower Int. Girth Weld #2 11! 1-313 33A277/0091 3977 -50 0.258 0,165 126 63.2 0.0 28.0 69.2 Fluence Data Wall Thickness {ln.} Fluence at ID Attenuation, Fluence Factor, FF Flue nee @ 1/4t (n/cm 2 )
Location Full 1/4t (nlcm 2) =
1/4t e~* 24 "
f0*2a.o.101og I)
Lower Shell #1 6.375 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower Shell #2 6.375 1.594 2.05E+18 0.682 1.40E+18 0.487
~I Lower Shell#3 6.375 1.594 2.05E+18 0.682 1.40E+18 0.487 lower-tnt. Shell #1 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-tnt Shell #2 5.375 1.344 2.43E+1B 0.724 1.76E+18 0.539 Lower-tnt Shell #3 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower Long. Weld 6.375 1.594 2.02E+18 0.682 1.38E+18 0.484 il Lower Int. Long Weld #1 5.375 1.344 1.52E+18 0.724 1.10E+18 0.437 Lower Int. Long Weld #2 5.375 1.344 1.52E+18 0.724 1.10E+18 0.437 Lower - Lower Int. Girth Weld #1 5.375 1.344 2.05E+18 0.724 1.48E+18 0.500 Lower - lower Int. Girth Weld #2 5.375 1.344 2.05E+18 0.724 1.48E+18 0.500
- 1. If GE CF =236 ' F is used then this location becomes the limiting beltline location by 7.7 'F o~.er the current limiting location.
Hatch Unit 1 PTLR Revision 1 Page 38 of40 Table 8: Hatch Unit 1 ART Table for 49.3 EFPY Adjustments for 1/4t Heat No. / Chemistry Chemistry Deecrlptlon Code No.
Flux Type Flux Lot No. Initial RTNOT rF)
Factor rF) .6RTNoT Margin Tenns ARTNoT I Cu (wt%1 Nl(wt%1 rF) mffl aa ("F) rF) !
Lower Shell #1 G-4805-1 C4112-1 - 8 0.13 0.64 92 49.4 0.0 17.0 91 .4 Lower Shell #2 G-4805-2 C4112-2 - 10 0.13 0.64 92 49.4 0.0 17.0 93.4
~I Lower Shell#3 G-4805-3 C4149-1 -10 0.14 0.57 99 53.0 0.0 17.0 77.0 Lower-lnt. Shell #1 G-4803-7 C4337-1 - -20 0.17 0.62 128 76.0 0 .0 17.0 90.0 Lower-In! Shell #2 G-4804-1 C3985-2 - - -20 0.11 0.60 65 38.5 0.0 8.5 35.5 Lower-lnt Shell #3 G-4804-2 C4114-2 . -20 0.12 0.70 221 132.0 0.0 8.5 129.0 Adjustments for 1/4t Heat No./ Chemistry Chemhltry Deecrlptlon Coda No. Flux Lot No. Initial RTNor ("F) .6RTNoT Margin Tenns ARTNoT Flux Type Factor ("F)
Cu (wt%1 Nl (wt%1 ("F) CJJ ("F) aa ("F) ("F)
Lower Long. Weld 1-307 13253/1092 3791 -50 0 .221 0.732 189 101 .2 0.0 28.0 107.2 Lower Int. Long Weld #1 1-308 1P2809/1 092 3854 -50 0.270 0.735 206 100.6 0.0 28.0 106.6
~I~ Lower Int. Long Weld #2 Lower- Lower Int. Girth Weld #1 1-308 1-313 1P2815/1092 90099/0091 3854 3977
-50
-10 0.316 0.197 0.724 0.060 219 91 107.0 50.4 0.0 0.0 28.0 25.2 113.0 90.8 Lower- Lower Int. Girth Weld #2 11 1 1-313 33A277/ 0091 3977 -50 0.258 0.165 126 69.6 0.0 28.0 75 .6 Fluence Data Wall Thickness On.} Fluence at ID Attenuation, Fluance Factor, FF Fluence @ 1/4t (n/cm 2) ,co.a.o.101og f)
Full 1/4t (nlcm 2) 1/4t., e-II.:Mx Location Lower Shell #1 6.375 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower Shell #2 6.375 1.594 2.56E+18 0.682 1.75E+18 0.537
~I Lower Shell#3 6.375 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower-In!. Shell #1 5.375 1.344 3.08E+18 0.724 2.23E+18 0.596 Lower-In! Shell #2 5.375 1.344 3.08E+18 0.724 2.23E+18 0.596 Lower-In! Shell #3 5.375 1.344 3.08E+18 0.724 2.23E+18 0.596 Lower Long. Weld 6.375 1.594 2.54E+18 0.682 1.73E+18 0.536 II Lower Int. Long Weld #1 5,375 1.344 1.95E+18 0.724 1.41E+18 0.489 Lower Int. Long Weld #2 5.375 1.344 1.95E+18 0.724 1.41E+18 0.489 Lower- Lower Int. Girth Weld #1 5.375 1.344 2.56E+18 0.724 1.85E+18 0 .551 I
Lower- Lower Int. Girth Weld #2 5.375 1.344 2.56E+18 0.724 1.85E+18 0.551
=236 ' F is used then this location becomes the limiting beltline location by 7.2 "F 01.erthe current limiting location.
- 1. lfGE CF
Hatch Unit 1 PTLR Revision 1 Page 39 of40 Table 9: Hatch Unit 1 Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Thermal, Ktt Thermal, Ktt Krp-app (450"F shock) (100 "F/hr Plate)
Feedwater 76.6 65.3 11.5 WLI 71.6 N/A 17.4 Core DP 32.3 N/A 1.7 Notes:
5
- 1. K, in units of ksi-in°*
Hatch Unit 1 PTLR Revision 1 Page 40 of 40 Appendix A HATCH UNIT 1 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, two surveillance capsules have been removed from the Hatch Nuclear Plant Unit 1 (HNP-1) reactor vessel. The first capsule was removed in 1984 after 5.75 EFPY and the second was removed in 1996 after 14.3 EFPY [20]. The surveillance capsule 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 [20].
Southern Nuclear Operating Company committed to use the ISP in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated March 10, 2003 [21]. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC for the period of extended operation [22]. HNP-1 continues to be a host plant under the ISP [12]. Two more HNP-1 capsules are scheduled to be removed and tested under the ISP in approximately 2016 and 2029.