NL-16-0172, Response to Request for Information Regarding Changes to Technical Specification Section 1.0, 3.4.9, and 5.0
| ML16041A244 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/09/2016 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-16-0172 | |
| Download: ML16041A244 (90) | |
Text
Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35201 Tel 205.992.7872 SOUTHERN << \
Fax 205.992.7601 NUCLEAR A SOUTHERN COMPANY FE9 0 9 2'016 Docket Nos.: 50-321 NL-16-0172 50-366 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9. and 5.0 Ladies and Gentlemen:
By letter dated April 2, 2015, as supplemented by letter dated November 12, 2015 {Agencywide Documents Access and Management System {ADAMS)
Accession Nos. ML15092A856 and ML15322A089, respectively), Southern Nuclear Operating Company {SNC) requested a license amendment to modify Technical Specifications {TS) Section 1.0, "Definitions," Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9, "RCS Pressure and Temperature {PIT) Limits," and Section 5.0, "Administrative Controls," at the Edwin I. Hatch Nuclear Plant {HNP), Units 1 and 2. By letter dated January 11, 2016, the Nuclear Regulatory Commission {NRC) staff requested additional information to complete their review. Enclosure 1 provides the SNC response to the NRC's requests for additional information {RAI).
Enclo~ure 2 provides updated pressure temperature limits reports {PTLRs) for HNP Unit 1 and Unit 2.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at {205) 992-7369.
U.S. Nuclear Regulatory Commission NL-16-0172 Page 2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectful~bmitted, c.~-r~
C. R. Pierce Regulatory Affairs Director CRP/RMJ Sworn to and subscribed before me this j!__ day of~ '2016.
~it/~-tJt Notary ublic My commission expires: /b 'g *20f-t
Enclosures:
- 1. SNC Response to NRC RAis
- 2. Updated Unit 1 and Unit 2 Pressure and Temperature Limits Reports (PTLRs) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RType: CHA02.004 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch State of Georgia Mr. J. H. Turner, Director- Environmental Protection Division
Edwin I. Hatch Nuclear Plant - Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9, and 5.0 Enclosure 1 SNC Response to NRC RAis
Enclosure 1 to NL-16-0172 SNC Response to NRC RAis Reference
- 1. Sommerville, D.V., "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Sl R-05-044, Rev. 1-A, June 2013.
RAI8 The responses to RAI 2, RAI 4, and RAI 6 from the November 12, 2015, letter included information that could result in revised Pressure Temperature Limit Reports {PTLRs). If this information does revise the PLTRs, please submit a revision for both HNP, Units 1 and 2. If the information does not require revised PTLRs, please provide a justification.
SNC Response to RAI 8 The response to RAI 2 does not require a change to the PTLRs, since that response provides all information requested in RAI 2.
The response to RAI 4 does require a change to each PTLR to identify which procedure or procedures from Appendix A of the Licensing Topical Report [1] was applied for plate and weld material. The additional statements have been added to the PTLR for each Unit.
The response to RAI 6 requires a change to the PTLR for Hatch Unit 1, Table 3 (Non-beltline only) to indicate that the applicable temperature is 202 °F for pressures of 312.6 psig and greater.
Updated PTLRs for HNP Unit 1 and Unit 2 are provided as Enclosure 2.
RAI9 The NRC staff was able to successfully reproduce all of the pressure-temperature (P-T) curves for HNP, Unit 2, for 50.1 effective full power years (EFPY). The NRC staff was able to successfully reproduce all of the P-T curves for HNP, Unit 1, for 38 EFPY except for one portion of Curve C. Figures 1 and 2 provide the results of the NRC's independent evaluation for HNP, Unit 1, at 38 EFPY for Curves Band C, respectively. As noted by the two text box annotations in Figure 1 and the lower text box in Figure 2 that begins with, "Curve Cis controlled ... ",
the NRC's evaluation indicates that the licensee's Curve C may not meet ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G {2004 Edition, per the PTLR Topical Report BWROG-TP-11-022-A, Revision 1), minimum temperature limits and therefore may be non-conservative.
Page 15 of Calculation No. 1001527.304, Rev. 2, as provided in the November 12, 2015, supplement, states that the feedwater nozzle thermal stress intensity factor (Kir) was scaled based on the available temperature differential. The NRC staff approximated this scaling in their independent assessment based on the information available in the documents supplied by the licensee. The NRC staff recognizes that there may be some differences caused by this approximation.
However, based on the comparison of results in Figure 1, the staff's approximation yields very close agreement to the licensee's Curve B for E1-1 to NL-16-0172 SNC Response to NRC RAis temperatures greater than 76 degrees Fahrenheit (F), so any differences caused by the staff's approximation appear to be insignificant.
Please clarify whether the lower portions of Curve C for Hatch 1 for HNP, Unit 1, 38 and 49.3 EFPY adequately bound the non-beltline region.
E1-2 to NL-16-0172 SNC Response to NRC RAis Figure 1: Results of NRC's Independent Calculation of Curve B for Hatch Unit 1 at 38 EFPY I
1,300 I
? I h
tJ,.
I 1,200
?
J, I
~
I I
~
1,100 r I c.
~
- I 'I
_1,000 ~
1:11 I 1 ~
~
14 II Cll 900 i2i
- c a.
? I
~
~ ~ I 0
I BOO
....1 I p w
I I
en y
en 4. J w 700
> I a::
0 I t; 600
/i I ASMEApp. G calculations 9tend~ by 40'F to v I allow for a Curve C 500 I shift of 40'F .
,/
I
\
I I I 400 I l
'l -NRCBeltllne
/ T 300 -----V + PTLR Limling Beltline
/:.. -----NRC Bottom Head
\ } 0 PTLR Bottom Head v
200 I - - -
- - NRC WU Nozzle 0 -1 0 50 100 150 200 250 30 0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)
Non-Beltline (FW Nozzlt!) curve is more Jimitinc here than Unit 1 Heatup.Cooldown Curve B, 38 EFPY the limitin& beltline (WU Nozzle) curve.
E1-3 to NL-16-0172 SNC Response to NRC RAis Figure 2: Results of NRC's Independent Calculation of Curve C for Hatch Unit 1 at 38 EFPY I
1,300
- PTLR Non-Be!Ulne Curve C 1,100 - -- -- - Curve A Min. Temp. at Pressure
--NRC Composle Curve C ~
a;1,000 j, .
I Q
900 ii5
- z:::
- a. ...
0 I
_, BOO 1- 10CFR50App. G w
mini mum limits. Purple (I)
(I) point is atT = ill'F I
w 700 from revised Table 3 in
> Enclosure 7 to RAJ a: responses. This is 0
t; . j inconsistent with the II 600 ii5 a:
value of 202'F stated in the response to RAJ 6
- and plotted In Fiaure 3
~
If 1- 500 of the PTLR.
i w
C/1 a: 400
.)
- )
v (I)
(I) w a: 300
- a. * :
Curve C is c:ontrolled by the Non-
.........; /
200 r
Beltline (FW Noule) curve here when the Curve 8 plot is shifted by 40'F. Curve C in Flcure 3 of PTLR (shown by purple dotted 100 line) does not reflect this and appears non-ronservative.
0 I I 0 50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)
Unit 1 Curve C, 38 EFPY E1-4 to NL-16-0172 SNC Response to NRC RAis SNC Response to RAI 9 Structural Integrity (81) has re-verified that the values in the PTLR are correct.
Differences between the reported curves and the values calculated by the NRC during their review are likely due to differences in the iteration calculations used to determine saturation pressure-temperature pairs. No revision to the PTLR or underlying calculations is required.
As discussed on the teleconference with the NRC, the following values are provided for information:
T Curve B, T Curve C, K,c, KIT scaled, KIP_Allowable, P, OF OF ksi-in°" 5 ksi-in°" 5 ksi-in°" 5 psi 0 40 42.S 28.4 7.1 62 10 so 44.6 29.3 7.6 69 20 60 47.1 30.S 8.3 78 30 70 S0.2 31.9 9.1 89 so 90 S8.S 3S.2 11.7 122 70 110 71.0 39.4 1S.8 176 RAI10 In the response to RAI 6 of the November 12, 2015 supplement, the licensee states that the values in Table 3 of the HNP, Unit 1, PTLR should list a temperature of 202 degrees F for pressures greater than 312.6 pounds per square inch gauge (psig). A revised Table 3 was provided in Enclosure 7 to the licensee's RAI responses. However, revised Table 3 in that enclosure lists a revised temperature of 212 degrees F for pressures greater than 312.6 psig. This value does not match the licensee's response, nor does it match the value plotted in Figure 3 of the HNP, Unit 1, PTLR. Refer to the text box that begins with, "1 0 CFR 50 App. G minimum limits," in Figure 2 above. Please clarify this discrepancy.
SNC Response to RAI 10 The discrepancy identified in the RAI has been corrected by modifying Table 3 of the Hatch 1 PTLR as noted in the response to RAI 8. The corresponding figures in that PTLR do not need revision.
RAI11 In the response to RAI 7 in the November 12, 2015, supplement, the licensee
- states, The current methodology, used to develop the curves in the PTLR, is based on Reference [1] ...
However, Reference 1 is the NRC request for additional information, and not a methodology for computing P-T curves. Please provide the correct Reference 1.
E1-5 to NL-16-0172 SNC Response to NRC RAis SNC Response to RAI 11 The correct reference for methodology is Reference [2] of letter NL-15-2034, which corresponds to Reference [1] of the PTLR. That reference is also Reference [1] of the present letter.
E1-6
Edwin I. Hatch Nuclear Plant- Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9, and 5.0 Enclosure 2 Updated Unit 1 and Unit 2 Pressure and Temperature Limits Reports (PTLRs)
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 0
Hatch Unit 1 PTLR Revision 0 Page 2 of40 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 0 Page 3 of40 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 3 8 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 I 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 0 Page 4 of40 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 0 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 0 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 3 8 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 ofthe 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 0 Page 7 of40
- Minimum bolt-up temperature limit?: 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 RTNDr,max + 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 [1 0].
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 psi g. However, the minimum RPV pressure is -14.7 psi g.
5.0 Discussion The adjusted reference temperature (ART) ofthe 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 0 Page 8 of40 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 114T 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/cm 2 and 2.23 x 10 18 n/cm 2, 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 region.
The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T (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 114T 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::S 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::S 25 "F/hr must be
Hatch Unit 1 PTLR Revision 0 Page 9 of40 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 chemistry (weight-percent copper and nickel) and ART at the 114T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10 17 n/cm 2 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 ofReference [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 0 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 corner 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, Kn, 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 Kn can be linearly scaled to determine the K1T 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 Kn is calculated based on the thermal ramp of 100°F/hr, which is associated with the shutdown transient [14]. The resulting combination of the thermal down shock and thermal ramp
Hatch Unit 1 PTLR Revision 0 Page 11 of40 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 BIEIIF methodology presented in the SI P-T Curve LTR [1] is 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 K1p can be linearly scaled to determine the K1p 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 *pfhr as described in Reference [ 14].
- The K1p is calculated [14] using the WRC 175 methodology [18].
Hatch Unit 1 PTLR Revision 0 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 (0900876.40 1, 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-HAl-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 0 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 ofNuclear Power Plant Components, Non-mandatory
Hatch Unit 1 PTLR Revision 0 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 of the 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.
Hatch Unit 1 PTLR Revision 0 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 - Overa ll T7 1300 I I
+/
1200 '
J 1100 fl --
I I I ,p
_l_ii c
1000 I ,-
Jf I
900 Ar I I
I
'iO 800 - -I
- I IJ
~ I l Cll 700 I H I
w-'
> I
~
I h
"' L __ _ I J
Cll 600 r-- - I- -I a::
I I I ;//
e I I I if
.5I I I
- I 500 - I Cll I
I I,Jf' A'
I Ill I
.s: I Gl 400 I I I I 300 I ~- ~I: ..:1 - I *-
I I
200 I
Min imum Bolt-Up 100 I
II Temperature= 76°F M inimum RPV 1 II Pressure = -14.7 psig r T 0 I I I 0 100 200 250 3lx>
- I Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 0 Page 16 of 40 Figure 2: HNP-1 P-T Curve 8 (Normal Operation- Core Not Critical) for 38 EFPY Curve B- Core Not Critical, Composite Curves
--Beltline ---*Bottom Head - - Non-Beltline Overall I
1300 I
1 I J'
I I
1200 I I
I I I '
I I
~
I 1100 i,
tl~~
~~
I I
I I 1000 1-I I
-+-!-
' I
~~
I
'I j
I I
900 - - -1 l
I I I
,J f
tI QG 800 I
'ill I I
~ I I
~ 700 I
I r=~=r-Ql
> I
~ I II t:<Ill I
Ql a: 600 I
.5 I II
.E
. I .
I
- I G) 500 I
I
~
i Ql I
~ 400 I I 1 300 I
II I
I
- .* I 200 I M inimum Bolt-Up I~
~-
Temperature = 76°F 100 I Minimum RPV II lr Pressure= -14.7 psig 2k 0
i II 0 150 200 3 )()
Minimum Reactor Vessel Metal Temperature ('F) i T
Hatch Unit 1 PTLR Revision 0 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-Beltline Overa ll 1300 ~-------,-------,.-
I I
I I
1100 + - - - - - - - + - - - - - - ' - - - - -:
I I
1-I 1000 --4--------. I I
I I
I 900 L----1 I I
I I
I QO 800 ' --+---*- - - + -
- w;
-r/'----1------:-
I
~
] 700 '
~ t :
- 600 - - - - - + - - - - - - - - l
~~ I "i§ ,'
- 1 500 1--+-------+- - - - -- - - - + - - - - - 1 Ill I
~ I
~
a: 400 1- -
1 I
I
+--:-
I 300 r--
1 I
I I
200 ~-
I I
I I
Minimum RPV Pressure =-14.7 psig 0 -----~-~------+---~~============~
0 so 100 150 200 250 3 I
Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit l PTLR Revision 0 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
--Beltline ----Bottom Head - - Non-Beltline Overa ll 1300 -
- /i I
I I I J t
1200
! I I
I li 1100 - I I
~
I I
I 1000 i I I
I I I I I 900 -- I
-- ~ --
I I
I I
'bi 800 - 1- -
I VI I
~ I l I Cll 700 I 1- -
n._,,I I
15 I I
tl 111 I ac: 600 Cll E
500 *- - -
- I It
~
Cll ~~
~
Ill I I Cll I
.t 400 -: ~#
I I I
300 r-- -- J~l I
I 200 I Minimum Bolt-Up Temperature = 76°F 100 I I Minimum RPV Pressure =-14.7 psig 0 I' 0 so 100 150 200 250 31Jo j_
Minimum Reactor Vessel Metal Temperature (°F) l
Hatch Unit 1 PTLR Revision 0 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 - Overa ll 1300
.---------~--------,-:**------,-----1~ '-.-------.
1 1200 -i------t----:----+-.,, ------=--~-'~"r,_---+-------i uoo -
! I i J p
/;~----+- r------------.ifc----t--------;
/ I 1000 I - i - - - - - -1 I
I l,r---------.----,r2lf/i----t- -
900 ~---~!--~;,~*---~~~~~~----l
]:
.~ 800 I l1 I
,' \
I .,
~ ~-----~--- l -
r*l ,-
700 +-
I I J~
I
--!--- i-lrllc v'-~------+-
--t- : II
~ 500 l I 'f i
I I ! : ,
It 400 -t-,----i---: :
I
- --+-,-/li-'F,I------+--------f--------1 1 ,p 300 _ , - - ---;~'---------+--------+-------!
1
,./
200 ~------+---- , L .
_4'7 Minimum Bolt-Up
~ ~
Temperature = 76"F 100 ~---m.II___L 1~1 I _ _ _ j ______l_/========].J I
Minimum RPV Pressure =-14.7 psig I
0 l___.\------!f..-,------L___~=.j
~
I .L 0 so 100 1r 200 2~ 3 l0 Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 0 Page 20 of40 Figure 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY Curve C- Core Critical, Composite Curves
--Bettline ----Bottom Head Non-Beltline Overa ll 1300 I
I 1200
~------~-----------------i~--------~--~----~--~--~
1 1100 +----- --------~----1-:-+________l I I
l - - - - -;
I I
1000 - - -
I I
~ ,-
CiD 800 ,
] 700 --~--------~i--
i
~ 600 - ,~------~~ I
~ I e ,'
~ I
~ 500 ~~--------+----- 1
~ ,'
~ ,
~ 400 ----~-------+-
300 200 Minimum Bolt-Up Temperature = 76"F Minimum RPV
=
Pressure -14.7 psig 0 so 100 150 200 Minimum Reactor Vessel Metal Temperature (°F}
Hatch Unit 1 PTLR Revision 0 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 0 Page 22 of40 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-T Curve 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 0 Page 23 of40 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 0 Page 24 of40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY (continued)
Bottom Head Region Curve 8 - 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 0 Page 25 of40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 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 0 Page 26 of40 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 0 Page 27 of40 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 0 Page 28 of40 Table 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-T Curve P-TCurve Temperature Pressure "F 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 0 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 0 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 0 Page 31 of40 Table 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY Beltline Region Curve 8- 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 0 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 8 - 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 0 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 0 Page 34 of40 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 0 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 0 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-T Curve P-T Curve Temperature Pressure "F 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 0 Page 37 of 40 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 RTNOTffl ARTNoT Margin Terms ARTNoT Flux Type Factor (-F)
Cu (wt%1 Nl (wt%1 ("F) 111 ("F) a11 ("F) ("F)
Lower Shell #1 G-4805-1 C4112-1 . 8 0.13 0.64 92 44.8 0.0 17.0 86.8 II Lower Shell #2 G-4805-2 C4112-2 - 10 0.13 0.64 92 44.8 0.0 17.0 88.8 Lower Shell#3 G-4805-3 C4149-1 - -10 0.14 0.57 99 48.0 0.0 17.0 72.0 Lower-lot. Shell #1 G-4803-7 C4337-1 - -20 0.17 0.62 128 68.7 0.0 17.0 82.7 Lower-In! Shell #2 G-4804-1 C3985-2 . -20 0.11 0.60 65 34.9 0.0 8.5 31.9 Lower-lnt Shell #3 G-4804-2 C4114-2 - -20 0.12 0.70 221 119.3 0.0 8.6 116.3 Adjustments for 1/4t Heat No./ Chemistry Chemistry Description Code No. Flux Lot No. Initial RTNoTI"F) ARTNOT Margin Terms ARTNDT Flux Type Factor ("F)
Cu (wt%1 Nl (wt %1 rFl a1 ("F) a11 ("F) ("F)
Lower Long. Weld 1-307 1325311092 3791 -50 0 .221 0.732 189 91 .5 0.0 28.0 97.5 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 <*> 1-313 33A277/0091 3977 -50 0.258 0.165 126 63.2 0.0 28.0 69.2 Fluence Data Wall Thickness (ln. I Fluence at ID Au. nuatlon, Fluence Factor, FF Fluence @ 1/4t (nlcm 2)
Full 1/4t (nlcm 2) 1/4t = .....24. fl0.28-o.101og I)
Location Lower Shell #1 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 II Lower Shell #2 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower Shell#3 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower-lot. Shell #1 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-lnt Shell #2 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-In! Shell #3 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower Long. Weld 6.376 1.594 2.02E+18 0.682 1.38E+18 0.484 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
. .. .. location .
= 236 'F IS used then lh1s 1ocat1on becomes the hm1t1ng beltlme locat1on by 7. 7 'F owr the current hmtt1ng
Hatch Unit l PTLR Revision 0 Page 38 of 40 Table 8: Hatch Unit 1 ART Table for 49.3 EFPY Adjulllments for 114t Heat No. I Cllemlllry Chemlllry Description Code No. Flux Lot No. Initial RTNoTf'F) ARTNoT Margin Tonns ARTNDT Flux Type Factor ("F)
Cu (wt%) Nl (wt%1 ("F) O"t f'F) O"t. ("F) ("F)
Lower Shell #1 G-4805-1 C4112-1 - 8 0.13 0.64 92 49.4 0.0 17.0 91.4 II Lower Shell #2 G-4805-2 C4112-2 10 0.13 0.64 92 49.4 0.0 17.0 93.4 Lower Shell#3 G-4805-3 C4149-1 - - 10 0.1 4 0.57 99 53.0 0.0 17.0 77.0 Lower-In!. 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 Lowor-Jnt Sholl #3 G-4804-2 C4114-2 - -20 0.12 0.70 221 132.0 0.0 8.6 129.0 Adjustments for 114t Heat No.I Chemlmy Chemistry Do scription Code No. Flux Lot No. I nltla I RTNOT ("F) ARTNoT Margin Tonns ARTNoT Flux Type Factor("F)
Cu (wt%) Nl (wt%1 ("F) O"t ("F) ali ("F) ("F)
Lower Long. Weld 1-307 1325311092 3791 -50 0.221 0.732 189 101 .2 0.0 28.0 107.2 Lower Int. Long Weld #1 1-308 1P2809/1092 3854 -50 0.270 0.735 206 100.6 0.0 28.0 106.6 Lower Int. Long Weld #2 1-308 1P281511 092 3854 -50 0.316 0.724 219 107.0 0.0 28.0 113.0 Lower - Lower Int. Girth Weld #1 1-313 90099/0091 3977 -10 0.197 0.060 91 50.4 0.0 25.2 90.8 Lower - Lower Int. Girth Weld #2 (II 1-313 33A277/ 0091 3977 -50 0.258 0.165 126 69.6 0.0 28.0 75.6 Fluence Data Wall Thickness (!n.j Fluance at 10 Atlenuatlon, Fluence Factor, FF Fluence @ 114t (nlcm 2) fua.o.101oa 11 Full 1/4t (nicm 2) 114t = * .0.241<
Location Lower Shell #1 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 II Lower Shell #2 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower Shell#3 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower-lnt. 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 6.376 1.594 2.54E+18 0.682 1.73E+18 0.536 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 Lower - Lower Int. Girth Weld #2 5.375 1.344 2.56E+18 0.724 1.85E+18 0.551
- 1. If GE CF = 236 "F IS used then th1s locatiOn becomes the ilm1t1ng beltilne location by 7.2 "F owr the current ilm1tmg locat1on.
Hatch Unit I PTLR Revision 0 Page 39 of40 Table 9: Hatch Unit 1 Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Thermal, K,t Thermal, K.t K.p-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:
- 1. K1 in units of ksi-in°*5
Hatch Unit 1 PTLR Revision 0 Page 40 of40 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.
Southern Nuclear Operating Co.
Hatch Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR) for 37 and 50.1 Effective Full-Power Years (EFPY)
Revision 0
Hatch Unit 2 PTLR Revision 0 Page 2 of40 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 I HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) 15 for 37 EFPY Figure 2 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) 16 for 37 EFPY Figure 3 HNP-2 P-T Curve C (Normal Operation- Core Critical) 17 for 37 EFPY Figure 4 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak 18 Test) for 50.1 EFPY Figure 5 HNP-2 P-T Curve B (Normal Operation- Core Not Critical) 19 for 50.1 EFPY Figure 6 HNP-2 P-T Curve C (Normal Operation- Core Critical) 20 for 50.1 EFPY
Hatch Unit 2 PTLR Revision 0 Page 3 of40 Section Page Table I HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 3 7 21 EFPY Table 2 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 23 EFPY Table 3 HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY 26 Table 4 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 29 EFPY Table 5 HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 31 EFPY Table 6 HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY 34 Table 7 Hatch Unit 2 ART Table for 37 EFPY 37 Table 8 Hatch Unit 2 ART Table for 50.1 EFPY 38 Table 9 Hatch Unit 2 Summary ofNozzle Stress Intensity Factors 39 Appendix A Hatch Unit 2 Reactor Vessel Materials Surveillance Program 40
Hatch Unit 2 PTLR Revision 0 Page 4 of40 1.0 Purpose The purpose of the Hatch Nuclear Plant, Unit 2 (HNP-2) 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 Aoplicability This report is applicable to the HNP-2 RPV for up to 3 7 and 50.1 Effective Full-Power Years (EFPY) [3].
The following HNP-2 Technical Specifications (TS) are 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 2 PTLR Revision 0 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 I 0 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 2 PTLR Revision 0 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 37 EFPY and 50.1 EFPY for HNP-2 as documented in Reference [8]. The HNP-2 P-T curves for 37 EFPY are provided in Figures 1 through 3, and a tabulation ofthe overall composite curves (by region) is included in Tables 1 through 3. The HNP-2 P-T curves for 50.1 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-2 vessel beltline materials are shown in Table 7 for 37 EFPY and Table 8 for 50.1 EFPY, taken from Reference [7]. The resulting P-T curves are based on the geometry, design and materials information for the HNP-2 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 1OO"Fihour2 [8].
1 Interpreted as the temperature change in any !-hour period is less than or equal to 25°F.
2 Interpreted as the temperature change in any !-hour period is less than or equal to I 00°F.
Hatch Unit 2 PTLR Revision 0 Page 7 of40
- Minimum bolt-up temperature limit 2: 90"F [8].
To address the NRC condition regarding lowest service temperature in Reference [1], the minimum temperature is set to 90 °F, which is equal to the RTNDT,max + 60 °F, for all curves.
This value is consistent with the previous minimum temperature limits developed in [9] and the minimum bolt-up temperature specified in [1 0].
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 psi g. However, the minimum RPV pressure is -14.7 psi g.
5.0 Discussion The adjusted reference temperature (ART) ofthe 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-2 vessel plate, weld, and forging materials [7]. The Cu and Ni values were used with Table 1 of RG 1.99 [6] 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 CF per Paragraph 1.1 of RG 1.99 for plates and forgings. Since only one surveillance capsule containing the appropriate plate heat has been tested no fitted chemistry factor is available.
The RPV ID fluence value, associated with the limiting ART, of 1.95 x 10 18 n/cm 2 at 37 EFPY was developed in Reference [7] based on linear interpolation between reported fluence values for 26.6 EFPY and 50.1 EFPY from Reference [5], which were calculated in accordance with RG 1.190 [4]. The RPV ID fluence value, associated with the limiting ART, of2.60 x 10 18 n/cm2 at
Hatch Unit 2 PTLR Revision 0 Page 8 of40 50.1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1.190.
These fluence values apply to the limiting beltline lower shell plate (Heat No. C8553-1). The fluence values for the limiting lower shell plate are based upon an attenuation factor of 0.682 for a postulated l/4T flaw. As a result, the 114T fluence for 37 EFPY and 50.1 EFPY for the limiting lower shell plate are 1.33 x 10 18 n/cm 2 and 1. 77 x 10 18 n/cm 2 , respectively, for HNP-2.
The water level instrument (WLI) nozzle is located in the lower intennediate shell beltline plates
[8]. The RPV ID fluence value of2.45 x 10 18 n/cm 2 at 37 EFPY was developed in Reference [7]
based on linear interpolation between reported fluence values for 26.6 EFPY and 50.1 EFPY from Reference [5], which were calculated in accordance with RG 1.190 [4]. The peak RPV ID fluence value of and 3.28 x 10 18 n/cm 2 at 50.1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1.190 [4]. These fluence values apply to the limiting lower intennediate shell plate (Heat No. C8579-2). The fluence values for the WLI nozzle are based upon an attenuation factor of0.724 for a postulated 114T flaw. As a result, the 1/4T fluence for 3 7 EFPY and 50.1 EFPY for the limiting lower intermediate shell plate are 1. 77 x 10 18 n/cm 2 and 2.38 x 10 18 n/cm 2 , respectively, for HNP-2. The recirculation inlet (N2) and outlet (Nl) nozzles do not exist in the beltline region. However, the outer edge of the recirculation inlet nozzle forging is within '14 inch of the beltline [7]. Based on the ART evaluation in Reference [7], the N2 nozzle forging material is not limiting.
The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T (inside surface flaw) and 3/4T (outside surface flaw) locations. When combining pressure and thennal stresses, it is usually necessary to evaluate stresses at the 114T and the 3/4T locations. This is because the thennal 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 thennal gradient stresses at the 114T 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 l/4T location. This approach is conservative because irradiation effects cause the
Hatch Unit 2 PTLR Revision 0 Page 9 of40 allowable toughness at the I/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~ IOO"F/hr 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"F/hr 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 chemistry (weight-percent copper and nickel) and ART at the l/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > I 0 17 n/cm 2 forE> I MeV) are shown in Table 7 for 37 EFPY and Table 8 for 50.1 EFPY [7]. The initial RTNDT values shown in Tables 7 and 8 have been previously approved for use by the NRC per Reference [I 7].
Per Reference [7] and in accordance with Appendix A of Reference [I], the HNP-2 representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [12].
The HNP- 2 representative plate and weld materials C8554 and 51912, respectively, are contained in the HNP-2 surveillance capsules [7]. BWRVIP-135 [12] contains surveillance capsule test results for the Hatch Unit 2 representative plate and weld materials. The representative plate heat does not match the target plate heat; however, it does match the heat for
Hatch Unit 2 PTLR Revision 0 Page 10 of40 plate material used in other beltline plates. Since only one surveillance capsule containing this plate heat has been tested no fitted chemistry factor is available; therefore, the CF calculated using the RG1.99 [6] tables is used. The representative weld material heat does not match any weld material heats used in the Hatch Unit 2 beltline; therefore, the CF calculated using the RG 1.99 tables is used. Therefore, Procedure 2 from Appendix A of [ 1] was used for both plate and weld materials.
The ANSYS fmite 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 nozzle [14]. At the time that each of the analyses above was 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-2 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 L TR [ 1] is used to calculate the thermal stress intensity, Kn, due to the thermal shock by fitting a third order polynomial equation to the path
Hatch Unit 2 PTLR Revision 0 Page 11 of40 stress distribution for the thermal shock load case [14]. Because operation is along the saturation curve, the resulting Kn can be linearly scaled to determine the Kn 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 Kn is calculated based on the thermal ramp of 100°Fihr, which is associated with the shutdown transient [14]. The resulting combination ofthe thermal down shock and thermal ramp Kn 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 were developed based on testing as described in Appendix A of Reference [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 BIEIIF methodology presented in the SI P-T Curve L TR [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-2 code of construction [16]. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.
Hatch Unit 2 PTLR Revision 0 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 (0900876.40 1, 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-HA2-001-R-001 Revision 0, "Edwin I.
Hatch Unit 2 Fluence Evaluation at End of Cycle 22 and 50.1 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. 1001527.302, Revision 1, "RPV Material Summary and ART Calculation", July 2014.
- 8. Structural Integrity Associates Calculation No. 1001527.305, Revision 2, "Hatch Unit 2 P-T Curve Calculation for 37 and 50.1 EFPY", August 2014.
- 9. General Electric Document No. GE-NE-B1100827-00-0l, "Plant Hatch Units 1 & 2 RPV Pressure Temperature Limits License Renewal Evaluation," March 1999.
- 10. NRC Docket No. 50-321, "Issuance of Amendment No. 177 to Facility Operating License DPR-57 and Amendment No. 118 to Facility Operating License NPF-5 -Edwin
Hatch Unit 2 PTLR Revision 0 Page 13 of40 I. Hatch Nuclear Plant, Units 1 and 2," Amendment No. 177, License No. DPR-57, January 1992, ADAMS Accession No. ML012990100.
- 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 2 Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-0619," \
NEDC-30256, DRF-137-0010, August 1983, General Electric Company. SI File No. 1001527.210.
- b. Stevens, G. L., "Updated Feedwater Nozzle Fracture Mechanics Analysis for Edwin I. Hatch Nuclear Power Station Unit 2," GE-NE-523-95-0991, Rev. 0, DRF B13-01524, September 1991, General Electric Company. SI File No.
1001527.210.
- c. Bothne, D., "Power Uprate Evaluation Report for Edwin I. Hatch Unit 2, Feedwater Nozzle NUREG-0619 Fracture Mechanics Analysis for Extended Power Uprate Conditions," GE-NE-B13-01869-065-02, 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".
Hatch Unit 2 PTLR Revision 0 Page 14 of 40
- 16. American Society ofMechanical Engineers, Boiler and Pressure Vessel Code,Section III, 1968 Ed. through 1970 Addenda.
- 17. NUREG-1803, "Safety Evaluation Report Related to the License Renewal ofthe Edwin I.
Hatch Nuclear Plant, Units 1 and 2," December 2001.
- 18. General Electric Report No. SASR 90-104, "E. I. Hatch Nuclear Power Station, Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis," May 1991. SI File No. 1001527.205
- 19. 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.
- 20. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, EPRI Product 1025144, October 2012.
Hatch Unit 2 PTLR Revision 0 Page 15 of 40 Figure 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Curve A
- Pressure Test, Composite Curves
-Beltllne ---*BottomHead - - Non-a.ltllne Ovel'iilll J!
1300
,I
/I r
1200 I I I
/ I I
J 1100 ' - -
/
I I I ~ I
- - -1 Ill -
_j 1000 ~
11 I /f 900 il
~
t/
i 800 l 700 I I 600 U/~1!I i
I II
- I I I
I I I
500 I
I J:
I I I
400 i __ _
- t'-* I L
300 -
-r
~
200 Minimum Bolt-Up Temperature = 90°F 100 i Minimum RPV 0
I I I Pressure= -14.7 psig (l 5~ 100 150 200 2~
I
-r I Minimum RHctorV....IMml Tempermn.. ('F)
Hatch Unit 2 PTLR Revision 0 Page 16 of40 Figure 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY Curve 8- Core Not Critical, Composite Curves
--Beltline ----Bottom Head - - Non-Beltline - overa ll 1~0 -----------,----------.--------~~ ~ i------;,~,- ~ --~---------,
1200 1100
+-----------!-------/--t- - --illl----+--------1
, I
+--------+-------.~1----~----4---------~
l 1 '
1 t7 1---1---:1J1:-_ --+--------j
/ I 1000 -
+----
! 1f/
900 - -
I I
Qjj 800 - - 1-I
-m I I
~.
700 +--------~----------~1*---------FW-------------------~
ii 600
,/
- ,'/+-------;
A'.*
-#-+1"1~1--------------------j
- 1
~
~~
sao 400
- ------4-----~
- I
- - - - - -1'- --'l. l . - - - - - - - -
- t--1----*'-----~--------
300 l_,
~i~ ~----
p 200 r--------+---1~ Minimum Bolt-Up Temperature 90"F=
100 ----+-- 4W--l-----+---f=======j Minimum RPV Pressure =-14.7 psig O -------~----~-+----~--~====T 1 ========~
i 0 1 0 200 2 ;o Minimum Reactor Vessel Metal Temperature (°F)
Hatch Unit 2 PTLR Revision 0 Page 17 of40 Figure 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY Curve C- Core Critical, Composite Curves
- - Beltline ----Bottom Head - - Non-Belt line Overall 1300 , - - - - - - - - , . . . - - - - - - . - - - - - - - , - - - - - . , ., - --,,-----,....---~
I 1
f I 1200 - - - - - - - i - - - - - - t - - - - - - + - - - 1---1---f.ll--- - - 1 f I I I 1100 _,___ _ _ _ + - - - - - - f - - ~~-1---!ff-----l I
I I I
1000 ------+------+-----~--/--'-f--!JJ!------1 I I I
I 900 "iii 800
'Vi
.S:
!> 700 -
I
~ 600 -
- =*e
- 1 500 - - _ _ _.,.
i GJ
.t 400 200 Minimum Bolt-Up Temperature= 90°F 100 Minimum RPV Pressure= -14.7 psig 00~=:========T=====~~
II I_Ll~~
r~-----2+ ~------2~~
1~ I Minimum Reactor Vessel Metal Temperature ('F)
Hatch Unit 2 PTLR Revision 0 Page 18 of40 Figure 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY Curve A
- Pressure Test, Composite Curves
-Beltllne ---*Bottom Head - - Non-Beltllne Overall 2300 I
- I l j' I I /I J
1200 ! I I / I
/ I 1100 I
/ -
I ,' I
,,~:
I 4r I
1000 - 1- -
,.. I l
1_ :7 I I I
+
900 I
I I / /
l I r-800
_ I ;. ~
J_~* J l . A I I ;
700 : 1//
I I
I I I lI 11 i/ li 600 i 500 -
I I
r:
r-400 300 - ~~Ji 200 I Minimum Bolt-Up Temperature= 90°F 100 I Minimum RPV Pressure = -14.7 psig I
1r 0
0 100 200 2i0 T Minimum RftctarV....I Mlltlll Temperature rF)
Hatch Unit 2 PTLR Revision 0 Page 19 of 40 Figure 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY Curve B - Core Not Critical, Composite Curves r
- - Beltline ----Bottom Head - - Non-Belt llne Overall 1300 I -
I I I I
I I I I 1- 1 ~
I I
1200 I I
- I I
I 1100 _f I
I I II l I
I I
- - I I
I I I
1000 I I
r I
I II 900 I 0
I I I
, I I
1./
QO 800 I I I -
~-
- I
, I
~ I I /
1 700
':l' i Cll
> I
~ 1/
" I ,,
I J
(
. I I j I
Cll 600 '
a:
c I I
- e I
- 1 500 Cll
~
'"Cll I
~ 400 300 -~-L--~~ _____,
I#
II
~*
200 Minimum Bolt-Up 100 I II Temperature= 90°F Minimum RPV ~
0 I II 1_.
Pressure= -14.7 psig I I 0 so 100 150 200 250 Minimum Reactor Vessel Metal Temperature ('F)
Hatch Unit 2 PTLR Revision 0 Page 20 of40 Figure 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY Curve C- Core Critical, Composite Curves
- Bettline ----Bottom Head - - Non-Beltline Overall 1300 ...--
I 1 /
I 1200 I --t~-----r'-1-t--;f.----f I
I 1100 r- -
1000 ~
900 - --~
Dii 800
'ii'i L
.s:
l Ql 700
~ ,_
a: 600 -
Ql -- -
- e I '
- 1 500 I
I I
g Ql I
.s:"'
I Ql I 400 - i--
300 ,___
200 J Minimum Bolt-Up Temperature =90°F 100~==========~~1-+--------t--------+-------~
M inimum RPV Pressure= -14.7 psig 0 2=========~====~~_~-L----~------~---~
1 1 0 0 1 200 250 Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 2 PTLR Revision 0 Page 21 of 40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Beltline Region Curve A -Pressure Test P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 506.7 98.5 554.7 105.7 602.7 112.0 650.7 117.6 698.8 122.7 746.8 127.3 794.8 131.5 842.8 135.3 890.8 138.9 938.8 142.3 986.8 145.4 1034.9 148.4 1082.9 151.2 1130.9 153.8 1178.9 156.3 1226.9 158.7 1274.9 161.0 1322.9
Hatch Unit 2 PTLR Revision 0 Page 22 of40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY (continued)
Bottom Head Region Curve A- Pressure Test P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 112.5 1166.6 115.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A -Pressure Test P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 1310.2
Hatch Unit 2 PTLR Revision 0 Page 23 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY Beltline Region Curve 8 - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 248.1 101.1 296.8 110.2 345.5 117.9 394.2 124.6 442.9 130.5 491.6 135.7 540.3 140.5 589.0 144.9 637.7 148.9 686.4 152.6 735.1 156.0 783.8 159.2 832.5 162.3 881.2 165.1 929.9 167.8 978.6 170.4 1027.3 172.8 1076.0 175.1 1124.7 177.4 1173.4 179.5 1222.1 181.5 1270.8 183.5 1319.5
Hatch Unit 2 PTLR Revision 0 Page 24 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY (continued)
Bottom Head Region Curve 8 - Core Not Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 110.1 739.9 114.0 788.4 117.7 836.8 121.1 885.2 124.3 933.6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9
Hatch Unit 2 PTLR Revision 0 Page 25 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY (continued)
Non-Beltline Region Curve 8 - Core Not Critical P-T Curve P-T Curve Temperature Pressure
- F psi 90.0 0.0 90.0 312.6 150.0 312.6 150.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 26 of40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 140.6 112.2 189.6 127.5 238.7 139.2 287.7 148.7 336.8 156.7 385.8 163.6 434.9 169.6 483.9 175.0 533.0 179.9 582.0 184.3 631.1 188.4 680.1 192.1 729.2 195.6 778.2 198.9 827.3 202.0 876.3 204.9 925.4 207.6 974.4 210.2 1023.5 212.7 1072.5 215.0 1121.6 217.2 1170.6 219.4 1219.7 221.5 1268.7 223.4 1317.8
Hatch Unit 2 PTLR Revision 0 Page 27 of 40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478.1 127.5 527.5 133.7 576.8 139.3 626.1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 1316.4
Hatch Unit 2 PTLR Revision 0 Page 28 of40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.2 193.0 101.6 232.9 110.6 272.7 118.2 312.6 190.0 312.6 190.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 29 of40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 481.5 99.5 530.6 107.5 579.8 114.3 628.9 120.4 678.1 125.8 727.3 130.7 776.4 135.1 825.6 139.2 874.7 142.9 923.9 146.4 973.1 149.7 1022.2 152.8 1071.4 155.7 1120.6 158.4 1169.7 161.0 1218.9 163.5 1268.0 165.8 1317.2
Hatch Unit 2 PTLR Revision 0 Page 30 of40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY (continued)
Bottom Head Region Curve A -Pressure Test P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 112.5 1166.6 115.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 1310.2
Hatch Unit 2 PTLR Revision 0 Page 31 of40 Table 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY Beltline Region Curve 8 - Core Not Critical P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 229.2 102.4 278.6 112.2 328.0 120.5 377.4 127.6 426.8 133.8 476.2 139.3 525.6 144.3 575.0 148.8 624.4 153.0 673.8 156.8 723.2 160.3 772.6 163.7 822.0 166.8 871.4 169.7 920.8 172.5 970.2 175.1 1019.6 177.6 1069.0 180.0 1118.4 182.2 1167.8 184.4 1217.2 186.5 1266.6 188.5 1316.0
Hatch Unit 2 PTLR Revision 0 Page 32 of40 Table 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY (continued)
Bottom Head Region Curve B - Core Not Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 110.1 739.9 114.0 788.4 117.7 836.8 121.1 885.2 124.3 933 .6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9
Hatch Unit 2 PTLR Revision 0 Page 33 of 40 Table 5: HNP-2 P-T Curve B (Normal Operation -Core Not Critical) for 50.1 EFPY (continued)
Non-Beltline Region Curve 8 - Core Not Critical P*T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 150.0 312.6 150.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 34 of40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure
- F psi 90.0 0.0 90.0 132.1 114.2 181.4 130.5 230.8 142.7 280.1 152.5 329.5 160.7 378.8 167.8 428.1 173.9 477.5 179.4 526.8 184.4 576.2 188.9 625.5 193.0 674.8 196.9 724.2 200.4 773.5 203.7 822.9 206.8 872.2 209.7 921.6 212.5 970.9 215.1 1020.2 217.6 1069.6 220.0 1118.9 222.2 1168.3 224.4 1217.6 226.5 1266.9 228.5 1316.3
Hatch Unit 2 PTLR Revision 0 Page 35 of 40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478.1 127.5 527.5 133.7 576.8 139.3 626.1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 1316.4
Hatch Unit 2 PTLR Revision 0 Page 36 of40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.2 193.0 101.6 232.9 110.6 272.7 118.2 312.6 190.0 312.6 190.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 37 of40 Table 7: Hatch Unit 2 ART Table for 37 EFPY Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNOT("F) ARTNDT Margin Terms ARTNOT Factor ("F)
Cu(wt%1 Nl (wt %1 ("F) GJ("F) a~ rfl rFJ Lower Shell #1 G-6603-1 C8553-2 - -20 0.08 0.58 51 24.3 0.0 12.2 28.6 II Lower Shell #2 G-&603-2 C8553-1 24 0.08 0.58 51 24.3 0.0 12.2 72.6 Lower Shell#3 G-6603-3 C8571-1 - 0 0.08 0.53 51 24.3 0.0 12.2 48.6 Lower-In!. Shell #1 G-6602-2 C8554-1 . -20 0.08 0.57 51 27.6 0.0 13.8 35.2 Lower-In! Shell #2 G-6602-1 C8554-2 - -10 0.08 0.58 51 27.6 0.0 13.8 45.2 Lower-In! Shell #3 G-6601-4 C8579-2 - -4 0.11 0.48 73 39.5 0.0 17.0 69.5 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNOT("F) ARTNOT Margin Terms ARTNoT Factor ("F)
Cu(wt%1 Nl(wt%1 ("F) a! ("F) a~ (*F) rFl II Lower Long. Weld 101-842 10137 - -50 0.216 0.043 98 45.1 0.0 22.5 40.2 Lower Int. Long Weld 101-834 51874 - *50 0.147 0.037 68 30.0 0.0 15.0 10.0 Lower
- Lower Int. Girth Weld 301-871 4P6052 - -50 0.047 0.049 31 15.2 0.0 7.6 -19.6 Fluence Data Wall Thlckness(ln.j Flue nee at ID Attenuation, Fluence Factor, FF Flue nee @11/4t (n/cm 2 )
Full 1/4t (n/cm 2 ) 1/4t = e~.24x fC0.211.0.101og I)
Location Lower Shell #1 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Lower Shell #2 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Jl Lower Shell#3 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Lower-In!. Shell #1 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower-In! Shell #2 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower-lnt Shell #3 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower Long. Weld 6.375 1.594 1.81E+18 0.682 1.23E+18 0.460 Lower Int. Long Weld 5.375 1.344 1.55E+18 0.724 1.13E+18 0.441 Lower
- Lower Int. Girth Weld 5.375 1.344 1.95E+18 0.724 1.41E+18 0.490
Hatch Unit 2 PTLR Revision 0 Page 38 of 40 Table 8: Hatch Unit 2 ART Table for 50.1 EFPY Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNori"F) ARTNoT Margin Terms ARTNoT Factor rFl Cu (wt%1 Nl (wt%1 ("F) at ("F) a6 rFl rFl Lower Shell #1 G-6603-1 C8553-2 - -20 0.08 0.58 51 27.6 0.0 13.8 35.2
- 0.0 II Lower Shell #2 G-6603-2 C8553-1 24 0.08 0.58 51 27.6 13.8 79.2 Lower Shell#3 G-6603-3 C8571-1 - 0 0.08 0.53 51 27.6 0.0 13.8 55.2 Lower-In!. Shell #1 G-6602-2 C8554-1 - -20 0.08 0.57 51 31.2 0.0 15.6 42.4 Lower-In! Shell #2 G-6602-1 C8554-2 - -10 0.08 0.58 51 31.2 0.0 15.6 52.4 Lower-In! Shell #3 G-6601-4 C8579-2 - -4 0.11 0.48 73 44.6 0.0 17.0 74.6 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. lnltia I RTNor ("F) ARTNoT Margin Terms ARTNoT Factor ("F)
Cu (wt%1 Nl (wt%1 rFl at rFl a6 rFI rFl
!I Lower Long. Weld 101-842 10137 - -50 0.216 0.043 98 51.4 0.0 25.7 52.8 Lower Int. Long Weld 101-834 51874 - -50 0.147 0.037 68 34.4 0.0 17.2 18.8 Lower - Lower Int. Girth Weld 301-871 4?6052 - -50 0.047 0.049 31 17.2 0.0 8.6 -15.6 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNorrFl ARTNoT Margin Terms ARTNor Factor ("F)
Cu (wt %1 j; Nl (wt %11 ("F) at ("F) a6 ("F) rFl Recirculation Inlet Nozzle G-6607 Q2Q24W - 10 0.180 0.810 141 11 .9 0.0 5.9 33.7 Fluence Data Wall Thickness (ln.) Fluence at 10 Attenuation, Fluence Factor, FF Flue nee @11/4t (n/cm 2) ,co.a.o.101og II Full 1/4t (nlcm 2 ) 1/4t = e-0.24x Location Lower Shell #1 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 il Lower Shell #2 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 Lower Shell#3 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 Lower-In!. Shell #1 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611 ii:
Lower-In! Shell #2 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611 Lower-In! Shell #3 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611
!I Lower Long . Weld 6.375 1.594 2.42E+18 0.682 1.65E+18 0.525 Lower Int. Long Weld 5.375 1.344 2.10E+18 0.724 1.52E+18 0.506 Lower- Lower Int. Girth Weld 5.375 1.344 2.60E+18 0.724 1.88E+18 0.555 Recirculation Inlet Nozzle 6.375 1.594 1.00E+17 0.682 6.82E+16 0.084
Hatch Unit 2 PTLR Revision 0 Page 39 of40 Table 9: Hatch Unit 2 Summary of Nozzle Stress Intensity Factors Applied Pressure, Thermal, K,t Thermal, Ktt Nozzle Ktp-app (450"F shock) (100 *F/hr Plate)
Feed water 78.9 46.8 12.9 WLI 80.0 N/A 19.9 Notes:
- 1. K1 in units of ksi-in°*5
Hatch Unit 2 PTLR Revision 0 Page 40 of 40 Appendix A HATCH UNIT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, one surveillance capsule was removed from the Hatch Nuclear Plant Unit 2 (HNP-2) reactor vessel in 1989 following cycle 8 [18]. 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 [18].
Southern Nuclear Operating Company committed to use the ISP in place of its existing surveillance programs 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 [19]. 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 [20]. HNP-2 continues to be a host plant under the ISP
[12]. Two more HNP-2 capsules are scheduled to be removed and tested under the ISP in approximately 2017 and 2027.
Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35201 Tel 205.992.7872 SOUTHERN << \
Fax 205.992.7601 NUCLEAR A SOUTHERN COMPANY FE9 0 9 2'016 Docket Nos.: 50-321 NL-16-0172 50-366 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9. and 5.0 Ladies and Gentlemen:
By letter dated April 2, 2015, as supplemented by letter dated November 12, 2015 {Agencywide Documents Access and Management System {ADAMS)
Accession Nos. ML15092A856 and ML15322A089, respectively), Southern Nuclear Operating Company {SNC) requested a license amendment to modify Technical Specifications {TS) Section 1.0, "Definitions," Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9, "RCS Pressure and Temperature {PIT) Limits," and Section 5.0, "Administrative Controls," at the Edwin I. Hatch Nuclear Plant {HNP), Units 1 and 2. By letter dated January 11, 2016, the Nuclear Regulatory Commission {NRC) staff requested additional information to complete their review. Enclosure 1 provides the SNC response to the NRC's requests for additional information {RAI).
Enclo~ure 2 provides updated pressure temperature limits reports {PTLRs) for HNP Unit 1 and Unit 2.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at {205) 992-7369.
U.S. Nuclear Regulatory Commission NL-16-0172 Page 2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectful~bmitted, c.~-r~
C. R. Pierce Regulatory Affairs Director CRP/RMJ Sworn to and subscribed before me this j!__ day of~ '2016.
~it/~-tJt Notary ublic My commission expires: /b 'g *20f-t
Enclosures:
- 1. SNC Response to NRC RAis
- 2. Updated Unit 1 and Unit 2 Pressure and Temperature Limits Reports (PTLRs) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RType: CHA02.004 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch State of Georgia Mr. J. H. Turner, Director- Environmental Protection Division
Edwin I. Hatch Nuclear Plant - Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9, and 5.0 Enclosure 1 SNC Response to NRC RAis
Enclosure 1 to NL-16-0172 SNC Response to NRC RAis Reference
- 1. Sommerville, D.V., "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Sl R-05-044, Rev. 1-A, June 2013.
RAI8 The responses to RAI 2, RAI 4, and RAI 6 from the November 12, 2015, letter included information that could result in revised Pressure Temperature Limit Reports {PTLRs). If this information does revise the PLTRs, please submit a revision for both HNP, Units 1 and 2. If the information does not require revised PTLRs, please provide a justification.
SNC Response to RAI 8 The response to RAI 2 does not require a change to the PTLRs, since that response provides all information requested in RAI 2.
The response to RAI 4 does require a change to each PTLR to identify which procedure or procedures from Appendix A of the Licensing Topical Report [1] was applied for plate and weld material. The additional statements have been added to the PTLR for each Unit.
The response to RAI 6 requires a change to the PTLR for Hatch Unit 1, Table 3 (Non-beltline only) to indicate that the applicable temperature is 202 °F for pressures of 312.6 psig and greater.
Updated PTLRs for HNP Unit 1 and Unit 2 are provided as Enclosure 2.
RAI9 The NRC staff was able to successfully reproduce all of the pressure-temperature (P-T) curves for HNP, Unit 2, for 50.1 effective full power years (EFPY). The NRC staff was able to successfully reproduce all of the P-T curves for HNP, Unit 1, for 38 EFPY except for one portion of Curve C. Figures 1 and 2 provide the results of the NRC's independent evaluation for HNP, Unit 1, at 38 EFPY for Curves Band C, respectively. As noted by the two text box annotations in Figure 1 and the lower text box in Figure 2 that begins with, "Curve Cis controlled ... ",
the NRC's evaluation indicates that the licensee's Curve C may not meet ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G {2004 Edition, per the PTLR Topical Report BWROG-TP-11-022-A, Revision 1), minimum temperature limits and therefore may be non-conservative.
Page 15 of Calculation No. 1001527.304, Rev. 2, as provided in the November 12, 2015, supplement, states that the feedwater nozzle thermal stress intensity factor (Kir) was scaled based on the available temperature differential. The NRC staff approximated this scaling in their independent assessment based on the information available in the documents supplied by the licensee. The NRC staff recognizes that there may be some differences caused by this approximation.
However, based on the comparison of results in Figure 1, the staff's approximation yields very close agreement to the licensee's Curve B for E1-1 to NL-16-0172 SNC Response to NRC RAis temperatures greater than 76 degrees Fahrenheit (F), so any differences caused by the staff's approximation appear to be insignificant.
Please clarify whether the lower portions of Curve C for Hatch 1 for HNP, Unit 1, 38 and 49.3 EFPY adequately bound the non-beltline region.
E1-2 to NL-16-0172 SNC Response to NRC RAis Figure 1: Results of NRC's Independent Calculation of Curve B for Hatch Unit 1 at 38 EFPY I
1,300 I
? I h
tJ,.
I 1,200
?
J, I
~
I I
~
1,100 r I c.
~
- I 'I
_1,000 ~
1:11 I 1 ~
~
14 II Cll 900 i2i
- c a.
? I
~
~ ~ I 0
I BOO
....1 I p w
I I
en y
en 4. J w 700
> I a::
0 I t; 600
/i I ASMEApp. G calculations 9tend~ by 40'F to v I allow for a Curve C 500 I shift of 40'F .
,/
I
\
I I I 400 I l
'l -NRCBeltllne
/ T 300 -----V + PTLR Limling Beltline
/:.. -----NRC Bottom Head
\ } 0 PTLR Bottom Head v
200 I - - -
- - NRC WU Nozzle 0 -1 0 50 100 150 200 250 30 0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)
Non-Beltline (FW Nozzlt!) curve is more Jimitinc here than Unit 1 Heatup.Cooldown Curve B, 38 EFPY the limitin& beltline (WU Nozzle) curve.
E1-3 to NL-16-0172 SNC Response to NRC RAis Figure 2: Results of NRC's Independent Calculation of Curve C for Hatch Unit 1 at 38 EFPY I
1,300
- PTLR Non-Be!Ulne Curve C 1,100 - -- -- - Curve A Min. Temp. at Pressure
--NRC Composle Curve C ~
a;1,000 j, .
I Q
900 ii5
- z:::
- a. ...
0 I
_, BOO 1- 10CFR50App. G w
mini mum limits. Purple (I)
(I) point is atT = ill'F I
w 700 from revised Table 3 in
> Enclosure 7 to RAJ a: responses. This is 0
t; . j inconsistent with the II 600 ii5 a:
value of 202'F stated in the response to RAJ 6
- and plotted In Fiaure 3
~
If 1- 500 of the PTLR.
i w
C/1 a: 400
.)
- )
v (I)
(I) w a: 300
- a. * :
Curve C is c:ontrolled by the Non-
.........; /
200 r
Beltline (FW Noule) curve here when the Curve 8 plot is shifted by 40'F. Curve C in Flcure 3 of PTLR (shown by purple dotted 100 line) does not reflect this and appears non-ronservative.
0 I I 0 50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)
Unit 1 Curve C, 38 EFPY E1-4 to NL-16-0172 SNC Response to NRC RAis SNC Response to RAI 9 Structural Integrity (81) has re-verified that the values in the PTLR are correct.
Differences between the reported curves and the values calculated by the NRC during their review are likely due to differences in the iteration calculations used to determine saturation pressure-temperature pairs. No revision to the PTLR or underlying calculations is required.
As discussed on the teleconference with the NRC, the following values are provided for information:
T Curve B, T Curve C, K,c, KIT scaled, KIP_Allowable, P, OF OF ksi-in°" 5 ksi-in°" 5 ksi-in°" 5 psi 0 40 42.S 28.4 7.1 62 10 so 44.6 29.3 7.6 69 20 60 47.1 30.S 8.3 78 30 70 S0.2 31.9 9.1 89 so 90 S8.S 3S.2 11.7 122 70 110 71.0 39.4 1S.8 176 RAI10 In the response to RAI 6 of the November 12, 2015 supplement, the licensee states that the values in Table 3 of the HNP, Unit 1, PTLR should list a temperature of 202 degrees F for pressures greater than 312.6 pounds per square inch gauge (psig). A revised Table 3 was provided in Enclosure 7 to the licensee's RAI responses. However, revised Table 3 in that enclosure lists a revised temperature of 212 degrees F for pressures greater than 312.6 psig. This value does not match the licensee's response, nor does it match the value plotted in Figure 3 of the HNP, Unit 1, PTLR. Refer to the text box that begins with, "1 0 CFR 50 App. G minimum limits," in Figure 2 above. Please clarify this discrepancy.
SNC Response to RAI 10 The discrepancy identified in the RAI has been corrected by modifying Table 3 of the Hatch 1 PTLR as noted in the response to RAI 8. The corresponding figures in that PTLR do not need revision.
RAI11 In the response to RAI 7 in the November 12, 2015, supplement, the licensee
- states, The current methodology, used to develop the curves in the PTLR, is based on Reference [1] ...
However, Reference 1 is the NRC request for additional information, and not a methodology for computing P-T curves. Please provide the correct Reference 1.
E1-5 to NL-16-0172 SNC Response to NRC RAis SNC Response to RAI 11 The correct reference for methodology is Reference [2] of letter NL-15-2034, which corresponds to Reference [1] of the PTLR. That reference is also Reference [1] of the present letter.
E1-6
Edwin I. Hatch Nuclear Plant- Units 1 & 2 Response to Request for Information Regarding Changes to Technical Specification Sections 1.0, 3.4.9, and 5.0 Enclosure 2 Updated Unit 1 and Unit 2 Pressure and Temperature Limits Reports (PTLRs)
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 0
Hatch Unit 1 PTLR Revision 0 Page 2 of40 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 0 Page 3 of40 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 3 8 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 I 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 0 Page 4 of40 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 0 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 0 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 3 8 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 ofthe 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 0 Page 7 of40
- Minimum bolt-up temperature limit?: 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 RTNDr,max + 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 [1 0].
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 psi g. However, the minimum RPV pressure is -14.7 psi g.
5.0 Discussion The adjusted reference temperature (ART) ofthe 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 0 Page 8 of40 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 114T 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/cm 2 and 2.23 x 10 18 n/cm 2, 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 region.
The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T (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 114T 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::S 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::S 25 "F/hr must be
Hatch Unit 1 PTLR Revision 0 Page 9 of40 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 chemistry (weight-percent copper and nickel) and ART at the 114T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10 17 n/cm 2 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 ofReference [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 0 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 corner 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, Kn, 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 Kn can be linearly scaled to determine the K1T 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 Kn is calculated based on the thermal ramp of 100°F/hr, which is associated with the shutdown transient [14]. The resulting combination of the thermal down shock and thermal ramp
Hatch Unit 1 PTLR Revision 0 Page 11 of40 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 BIEIIF methodology presented in the SI P-T Curve LTR [1] is 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 K1p can be linearly scaled to determine the K1p 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 *pfhr as described in Reference [ 14].
- The K1p is calculated [14] using the WRC 175 methodology [18].
Hatch Unit 1 PTLR Revision 0 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 (0900876.40 1, 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-HAl-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 0 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 ofNuclear Power Plant Components, Non-mandatory
Hatch Unit 1 PTLR Revision 0 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 of the 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.
Hatch Unit 1 PTLR Revision 0 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 - Overa ll T7 1300 I I
+/
1200 '
J 1100 fl --
I I I ,p
_l_ii c
1000 I ,-
Jf I
900 Ar I I
I
'iO 800 - -I
- I IJ
~ I l Cll 700 I H I
w-'
> I
~
I h
"' L __ _ I J
Cll 600 r-- - I- -I a::
I I I ;//
e I I I if
.5I I I
- I 500 - I Cll I
I I,Jf' A'
I Ill I
.s: I Gl 400 I I I I 300 I ~- ~I: ..:1 - I *-
I I
200 I
Min imum Bolt-Up 100 I
II Temperature= 76°F M inimum RPV 1 II Pressure = -14.7 psig r T 0 I I I 0 100 200 250 3lx>
- I Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 0 Page 16 of 40 Figure 2: HNP-1 P-T Curve 8 (Normal Operation- Core Not Critical) for 38 EFPY Curve B- Core Not Critical, Composite Curves
--Beltline ---*Bottom Head - - Non-Beltline Overall I
1300 I
1 I J'
I I
1200 I I
I I I '
I I
~
I 1100 i,
tl~~
~~
I I
I I 1000 1-I I
-+-!-
' I
~~
I
'I j
I I
900 - - -1 l
I I I
,J f
tI QG 800 I
'ill I I
~ I I
~ 700 I
I r=~=r-Ql
> I
~ I II t:<Ill I
Ql a: 600 I
.5 I II
.E
. I .
I
- I G) 500 I
I
~
i Ql I
~ 400 I I 1 300 I
II I
I
- .* I 200 I M inimum Bolt-Up I~
~-
Temperature = 76°F 100 I Minimum RPV II lr Pressure= -14.7 psig 2k 0
i II 0 150 200 3 )()
Minimum Reactor Vessel Metal Temperature ('F) i T
Hatch Unit 1 PTLR Revision 0 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-Beltline Overa ll 1300 ~-------,-------,.-
I I
I I
1100 + - - - - - - - + - - - - - - ' - - - - -:
I I
1-I 1000 --4--------. I I
I I
I 900 L----1 I I
I I
I QO 800 ' --+---*- - - + -
- w;
-r/'----1------:-
I
~
] 700 '
~ t :
- 600 - - - - - + - - - - - - - - l
~~ I "i§ ,'
- 1 500 1--+-------+- - - - -- - - - + - - - - - 1 Ill I
~ I
~
a: 400 1- -
1 I
I
+--:-
I 300 r--
1 I
I I
200 ~-
I I
I I
Minimum RPV Pressure =-14.7 psig 0 -----~-~------+---~~============~
0 so 100 150 200 250 3 I
Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit l PTLR Revision 0 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
--Beltline ----Bottom Head - - Non-Beltline Overa ll 1300 -
- /i I
I I I J t
1200
! I I
I li 1100 - I I
~
I I
I 1000 i I I
I I I I I 900 -- I
-- ~ --
I I
I I
'bi 800 - 1- -
I VI I
~ I l I Cll 700 I 1- -
n._,,I I
15 I I
tl 111 I ac: 600 Cll E
500 *- - -
- I It
~
Cll ~~
~
Ill I I Cll I
.t 400 -: ~#
I I I
300 r-- -- J~l I
I 200 I Minimum Bolt-Up Temperature = 76°F 100 I I Minimum RPV Pressure =-14.7 psig 0 I' 0 so 100 150 200 250 31Jo j_
Minimum Reactor Vessel Metal Temperature (°F) l
Hatch Unit 1 PTLR Revision 0 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 - Overa ll 1300
.---------~--------,-:**------,-----1~ '-.-------.
1 1200 -i------t----:----+-.,, ------=--~-'~"r,_---+-------i uoo -
! I i J p
/;~----+- r------------.ifc----t--------;
/ I 1000 I - i - - - - - -1 I
I l,r---------.----,r2lf/i----t- -
900 ~---~!--~;,~*---~~~~~~----l
]:
.~ 800 I l1 I
,' \
I .,
~ ~-----~--- l -
r*l ,-
700 +-
I I J~
I
--!--- i-lrllc v'-~------+-
--t- : II
~ 500 l I 'f i
I I ! : ,
It 400 -t-,----i---: :
I
- --+-,-/li-'F,I------+--------f--------1 1 ,p 300 _ , - - ---;~'---------+--------+-------!
1
,./
200 ~------+---- , L .
_4'7 Minimum Bolt-Up
~ ~
Temperature = 76"F 100 ~---m.II___L 1~1 I _ _ _ j ______l_/========].J I
Minimum RPV Pressure =-14.7 psig I
0 l___.\------!f..-,------L___~=.j
~
I .L 0 so 100 1r 200 2~ 3 l0 Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 1 PTLR Revision 0 Page 20 of40 Figure 6: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 49.3 EFPY Curve C- Core Critical, Composite Curves
--Bettline ----Bottom Head Non-Beltline Overa ll 1300 I
I 1200
~------~-----------------i~--------~--~----~--~--~
1 1100 +----- --------~----1-:-+________l I I
l - - - - -;
I I
1000 - - -
I I
~ ,-
CiD 800 ,
] 700 --~--------~i--
i
~ 600 - ,~------~~ I
~ I e ,'
~ I
~ 500 ~~--------+----- 1
~ ,'
~ ,
~ 400 ----~-------+-
300 200 Minimum Bolt-Up Temperature = 76"F Minimum RPV
=
Pressure -14.7 psig 0 so 100 150 200 Minimum Reactor Vessel Metal Temperature (°F}
Hatch Unit 1 PTLR Revision 0 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 0 Page 22 of40 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-T Curve 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 0 Page 23 of40 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 0 Page 24 of40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 EFPY (continued)
Bottom Head Region Curve 8 - 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 0 Page 25 of40 Table 2: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 38 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 0 Page 26 of40 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 0 Page 27 of40 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 0 Page 28 of40 Table 3: HNP-1 P-T Curve C (Normal Operation- Core Critical) for 38 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-T Curve P-TCurve Temperature Pressure "F 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 0 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 0 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 0 Page 31 of40 Table 5: HNP-1 P-T Curve B (Normal Operation- Core Not Critical) for 49.3 EFPY Beltline Region Curve 8- 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 0 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 8 - 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 0 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 0 Page 34 of40 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 0 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 0 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-T Curve P-T Curve Temperature Pressure "F 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 0 Page 37 of 40 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 RTNOTffl ARTNoT Margin Terms ARTNoT Flux Type Factor (-F)
Cu (wt%1 Nl (wt%1 ("F) 111 ("F) a11 ("F) ("F)
Lower Shell #1 G-4805-1 C4112-1 . 8 0.13 0.64 92 44.8 0.0 17.0 86.8 II Lower Shell #2 G-4805-2 C4112-2 - 10 0.13 0.64 92 44.8 0.0 17.0 88.8 Lower Shell#3 G-4805-3 C4149-1 - -10 0.14 0.57 99 48.0 0.0 17.0 72.0 Lower-lot. Shell #1 G-4803-7 C4337-1 - -20 0.17 0.62 128 68.7 0.0 17.0 82.7 Lower-In! Shell #2 G-4804-1 C3985-2 . -20 0.11 0.60 65 34.9 0.0 8.5 31.9 Lower-lnt Shell #3 G-4804-2 C4114-2 - -20 0.12 0.70 221 119.3 0.0 8.6 116.3 Adjustments for 1/4t Heat No./ Chemistry Chemistry Description Code No. Flux Lot No. Initial RTNoTI"F) ARTNOT Margin Terms ARTNDT Flux Type Factor ("F)
Cu (wt%1 Nl (wt %1 rFl a1 ("F) a11 ("F) ("F)
Lower Long. Weld 1-307 1325311092 3791 -50 0 .221 0.732 189 91 .5 0.0 28.0 97.5 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 <*> 1-313 33A277/0091 3977 -50 0.258 0.165 126 63.2 0.0 28.0 69.2 Fluence Data Wall Thickness (ln. I Fluence at ID Au. nuatlon, Fluence Factor, FF Fluence @ 1/4t (nlcm 2)
Full 1/4t (nlcm 2) 1/4t = .....24. fl0.28-o.101og I)
Location Lower Shell #1 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 II Lower Shell #2 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower Shell#3 6.376 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower-lot. Shell #1 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-lnt Shell #2 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-In! Shell #3 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower Long. Weld 6.376 1.594 2.02E+18 0.682 1.38E+18 0.484 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
. .. .. location .
= 236 'F IS used then lh1s 1ocat1on becomes the hm1t1ng beltlme locat1on by 7. 7 'F owr the current hmtt1ng
Hatch Unit l PTLR Revision 0 Page 38 of 40 Table 8: Hatch Unit 1 ART Table for 49.3 EFPY Adjulllments for 114t Heat No. I Cllemlllry Chemlllry Description Code No. Flux Lot No. Initial RTNoTf'F) ARTNoT Margin Tonns ARTNDT Flux Type Factor ("F)
Cu (wt%) Nl (wt%1 ("F) O"t f'F) O"t. ("F) ("F)
Lower Shell #1 G-4805-1 C4112-1 - 8 0.13 0.64 92 49.4 0.0 17.0 91.4 II Lower Shell #2 G-4805-2 C4112-2 10 0.13 0.64 92 49.4 0.0 17.0 93.4 Lower Shell#3 G-4805-3 C4149-1 - - 10 0.1 4 0.57 99 53.0 0.0 17.0 77.0 Lower-In!. 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 Lowor-Jnt Sholl #3 G-4804-2 C4114-2 - -20 0.12 0.70 221 132.0 0.0 8.6 129.0 Adjustments for 114t Heat No.I Chemlmy Chemistry Do scription Code No. Flux Lot No. I nltla I RTNOT ("F) ARTNoT Margin Tonns ARTNoT Flux Type Factor("F)
Cu (wt%) Nl (wt%1 ("F) O"t ("F) ali ("F) ("F)
Lower Long. Weld 1-307 1325311092 3791 -50 0.221 0.732 189 101 .2 0.0 28.0 107.2 Lower Int. Long Weld #1 1-308 1P2809/1092 3854 -50 0.270 0.735 206 100.6 0.0 28.0 106.6 Lower Int. Long Weld #2 1-308 1P281511 092 3854 -50 0.316 0.724 219 107.0 0.0 28.0 113.0 Lower - Lower Int. Girth Weld #1 1-313 90099/0091 3977 -10 0.197 0.060 91 50.4 0.0 25.2 90.8 Lower - Lower Int. Girth Weld #2 (II 1-313 33A277/ 0091 3977 -50 0.258 0.165 126 69.6 0.0 28.0 75.6 Fluence Data Wall Thickness (!n.j Fluance at 10 Atlenuatlon, Fluence Factor, FF Fluence @ 114t (nlcm 2) fua.o.101oa 11 Full 1/4t (nicm 2) 114t = * .0.241<
Location Lower Shell #1 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 II Lower Shell #2 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower Shell#3 6.376 1.594 2.56E+18 0.682 1.75E+18 0.537 Lower-lnt. 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 6.376 1.594 2.54E+18 0.682 1.73E+18 0.536 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 Lower - Lower Int. Girth Weld #2 5.375 1.344 2.56E+18 0.724 1.85E+18 0.551
- 1. If GE CF = 236 "F IS used then th1s locatiOn becomes the ilm1t1ng beltilne location by 7.2 "F owr the current ilm1tmg locat1on.
Hatch Unit I PTLR Revision 0 Page 39 of40 Table 9: Hatch Unit 1 Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Thermal, K,t Thermal, K.t K.p-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:
- 1. K1 in units of ksi-in°*5
Hatch Unit 1 PTLR Revision 0 Page 40 of40 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.
Southern Nuclear Operating Co.
Hatch Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR) for 37 and 50.1 Effective Full-Power Years (EFPY)
Revision 0
Hatch Unit 2 PTLR Revision 0 Page 2 of40 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 I HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) 15 for 37 EFPY Figure 2 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) 16 for 37 EFPY Figure 3 HNP-2 P-T Curve C (Normal Operation- Core Critical) 17 for 37 EFPY Figure 4 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak 18 Test) for 50.1 EFPY Figure 5 HNP-2 P-T Curve B (Normal Operation- Core Not Critical) 19 for 50.1 EFPY Figure 6 HNP-2 P-T Curve C (Normal Operation- Core Critical) 20 for 50.1 EFPY
Hatch Unit 2 PTLR Revision 0 Page 3 of40 Section Page Table I HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 3 7 21 EFPY Table 2 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 23 EFPY Table 3 HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY 26 Table 4 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 29 EFPY Table 5 HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 31 EFPY Table 6 HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY 34 Table 7 Hatch Unit 2 ART Table for 37 EFPY 37 Table 8 Hatch Unit 2 ART Table for 50.1 EFPY 38 Table 9 Hatch Unit 2 Summary ofNozzle Stress Intensity Factors 39 Appendix A Hatch Unit 2 Reactor Vessel Materials Surveillance Program 40
Hatch Unit 2 PTLR Revision 0 Page 4 of40 1.0 Purpose The purpose of the Hatch Nuclear Plant, Unit 2 (HNP-2) 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 Aoplicability This report is applicable to the HNP-2 RPV for up to 3 7 and 50.1 Effective Full-Power Years (EFPY) [3].
The following HNP-2 Technical Specifications (TS) are 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 2 PTLR Revision 0 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 I 0 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 2 PTLR Revision 0 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 37 EFPY and 50.1 EFPY for HNP-2 as documented in Reference [8]. The HNP-2 P-T curves for 37 EFPY are provided in Figures 1 through 3, and a tabulation ofthe overall composite curves (by region) is included in Tables 1 through 3. The HNP-2 P-T curves for 50.1 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-2 vessel beltline materials are shown in Table 7 for 37 EFPY and Table 8 for 50.1 EFPY, taken from Reference [7]. The resulting P-T curves are based on the geometry, design and materials information for the HNP-2 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 1OO"Fihour2 [8].
1 Interpreted as the temperature change in any !-hour period is less than or equal to 25°F.
2 Interpreted as the temperature change in any !-hour period is less than or equal to I 00°F.
Hatch Unit 2 PTLR Revision 0 Page 7 of40
- Minimum bolt-up temperature limit 2: 90"F [8].
To address the NRC condition regarding lowest service temperature in Reference [1], the minimum temperature is set to 90 °F, which is equal to the RTNDT,max + 60 °F, for all curves.
This value is consistent with the previous minimum temperature limits developed in [9] and the minimum bolt-up temperature specified in [1 0].
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 psi g. However, the minimum RPV pressure is -14.7 psi g.
5.0 Discussion The adjusted reference temperature (ART) ofthe 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-2 vessel plate, weld, and forging materials [7]. The Cu and Ni values were used with Table 1 of RG 1.99 [6] 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 CF per Paragraph 1.1 of RG 1.99 for plates and forgings. Since only one surveillance capsule containing the appropriate plate heat has been tested no fitted chemistry factor is available.
The RPV ID fluence value, associated with the limiting ART, of 1.95 x 10 18 n/cm 2 at 37 EFPY was developed in Reference [7] based on linear interpolation between reported fluence values for 26.6 EFPY and 50.1 EFPY from Reference [5], which were calculated in accordance with RG 1.190 [4]. The RPV ID fluence value, associated with the limiting ART, of2.60 x 10 18 n/cm2 at
Hatch Unit 2 PTLR Revision 0 Page 8 of40 50.1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1.190.
These fluence values apply to the limiting beltline lower shell plate (Heat No. C8553-1). The fluence values for the limiting lower shell plate are based upon an attenuation factor of 0.682 for a postulated l/4T flaw. As a result, the 114T fluence for 37 EFPY and 50.1 EFPY for the limiting lower shell plate are 1.33 x 10 18 n/cm 2 and 1. 77 x 10 18 n/cm 2 , respectively, for HNP-2.
The water level instrument (WLI) nozzle is located in the lower intennediate shell beltline plates
[8]. The RPV ID fluence value of2.45 x 10 18 n/cm 2 at 37 EFPY was developed in Reference [7]
based on linear interpolation between reported fluence values for 26.6 EFPY and 50.1 EFPY from Reference [5], which were calculated in accordance with RG 1.190 [4]. The peak RPV ID fluence value of and 3.28 x 10 18 n/cm 2 at 50.1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1.190 [4]. These fluence values apply to the limiting lower intennediate shell plate (Heat No. C8579-2). The fluence values for the WLI nozzle are based upon an attenuation factor of0.724 for a postulated 114T flaw. As a result, the 1/4T fluence for 3 7 EFPY and 50.1 EFPY for the limiting lower intermediate shell plate are 1. 77 x 10 18 n/cm 2 and 2.38 x 10 18 n/cm 2 , respectively, for HNP-2. The recirculation inlet (N2) and outlet (Nl) nozzles do not exist in the beltline region. However, the outer edge of the recirculation inlet nozzle forging is within '14 inch of the beltline [7]. Based on the ART evaluation in Reference [7], the N2 nozzle forging material is not limiting.
The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T (inside surface flaw) and 3/4T (outside surface flaw) locations. When combining pressure and thennal stresses, it is usually necessary to evaluate stresses at the 114T and the 3/4T locations. This is because the thennal 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 thennal gradient stresses at the 114T 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 l/4T location. This approach is conservative because irradiation effects cause the
Hatch Unit 2 PTLR Revision 0 Page 9 of40 allowable toughness at the I/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~ IOO"F/hr 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"F/hr 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 chemistry (weight-percent copper and nickel) and ART at the l/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > I 0 17 n/cm 2 forE> I MeV) are shown in Table 7 for 37 EFPY and Table 8 for 50.1 EFPY [7]. The initial RTNDT values shown in Tables 7 and 8 have been previously approved for use by the NRC per Reference [I 7].
Per Reference [7] and in accordance with Appendix A of Reference [I], the HNP-2 representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [12].
The HNP- 2 representative plate and weld materials C8554 and 51912, respectively, are contained in the HNP-2 surveillance capsules [7]. BWRVIP-135 [12] contains surveillance capsule test results for the Hatch Unit 2 representative plate and weld materials. The representative plate heat does not match the target plate heat; however, it does match the heat for
Hatch Unit 2 PTLR Revision 0 Page 10 of40 plate material used in other beltline plates. Since only one surveillance capsule containing this plate heat has been tested no fitted chemistry factor is available; therefore, the CF calculated using the RG1.99 [6] tables is used. The representative weld material heat does not match any weld material heats used in the Hatch Unit 2 beltline; therefore, the CF calculated using the RG 1.99 tables is used. Therefore, Procedure 2 from Appendix A of [ 1] was used for both plate and weld materials.
The ANSYS fmite 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 nozzle [14]. At the time that each of the analyses above was 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-2 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 L TR [ 1] is used to calculate the thermal stress intensity, Kn, due to the thermal shock by fitting a third order polynomial equation to the path
Hatch Unit 2 PTLR Revision 0 Page 11 of40 stress distribution for the thermal shock load case [14]. Because operation is along the saturation curve, the resulting Kn can be linearly scaled to determine the Kn 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 Kn is calculated based on the thermal ramp of 100°Fihr, which is associated with the shutdown transient [14]. The resulting combination ofthe thermal down shock and thermal ramp Kn 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 were developed based on testing as described in Appendix A of Reference [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 BIEIIF methodology presented in the SI P-T Curve L TR [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-2 code of construction [16]. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.
Hatch Unit 2 PTLR Revision 0 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 (0900876.40 1, 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-HA2-001-R-001 Revision 0, "Edwin I.
Hatch Unit 2 Fluence Evaluation at End of Cycle 22 and 50.1 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. 1001527.302, Revision 1, "RPV Material Summary and ART Calculation", July 2014.
- 8. Structural Integrity Associates Calculation No. 1001527.305, Revision 2, "Hatch Unit 2 P-T Curve Calculation for 37 and 50.1 EFPY", August 2014.
- 9. General Electric Document No. GE-NE-B1100827-00-0l, "Plant Hatch Units 1 & 2 RPV Pressure Temperature Limits License Renewal Evaluation," March 1999.
- 10. NRC Docket No. 50-321, "Issuance of Amendment No. 177 to Facility Operating License DPR-57 and Amendment No. 118 to Facility Operating License NPF-5 -Edwin
Hatch Unit 2 PTLR Revision 0 Page 13 of40 I. Hatch Nuclear Plant, Units 1 and 2," Amendment No. 177, License No. DPR-57, January 1992, ADAMS Accession No. ML012990100.
- 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 2 Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-0619," \
NEDC-30256, DRF-137-0010, August 1983, General Electric Company. SI File No. 1001527.210.
- b. Stevens, G. L., "Updated Feedwater Nozzle Fracture Mechanics Analysis for Edwin I. Hatch Nuclear Power Station Unit 2," GE-NE-523-95-0991, Rev. 0, DRF B13-01524, September 1991, General Electric Company. SI File No.
1001527.210.
- c. Bothne, D., "Power Uprate Evaluation Report for Edwin I. Hatch Unit 2, Feedwater Nozzle NUREG-0619 Fracture Mechanics Analysis for Extended Power Uprate Conditions," GE-NE-B13-01869-065-02, 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".
Hatch Unit 2 PTLR Revision 0 Page 14 of 40
- 16. American Society ofMechanical Engineers, Boiler and Pressure Vessel Code,Section III, 1968 Ed. through 1970 Addenda.
- 17. NUREG-1803, "Safety Evaluation Report Related to the License Renewal ofthe Edwin I.
Hatch Nuclear Plant, Units 1 and 2," December 2001.
- 18. General Electric Report No. SASR 90-104, "E. I. Hatch Nuclear Power Station, Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis," May 1991. SI File No. 1001527.205
- 19. 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.
- 20. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, EPRI Product 1025144, October 2012.
Hatch Unit 2 PTLR Revision 0 Page 15 of 40 Figure 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Curve A
- Pressure Test, Composite Curves
-Beltllne ---*BottomHead - - Non-a.ltllne Ovel'iilll J!
1300
,I
/I r
1200 I I I
/ I I
J 1100 ' - -
/
I I I ~ I
- - -1 Ill -
_j 1000 ~
11 I /f 900 il
~
t/
i 800 l 700 I I 600 U/~1!I i
I II
- I I I
I I I
500 I
I J:
I I I
400 i __ _
- t'-* I L
300 -
-r
~
200 Minimum Bolt-Up Temperature = 90°F 100 i Minimum RPV 0
I I I Pressure= -14.7 psig (l 5~ 100 150 200 2~
I
-r I Minimum RHctorV....IMml Tempermn.. ('F)
Hatch Unit 2 PTLR Revision 0 Page 16 of40 Figure 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY Curve 8- Core Not Critical, Composite Curves
--Beltline ----Bottom Head - - Non-Beltline - overa ll 1~0 -----------,----------.--------~~ ~ i------;,~,- ~ --~---------,
1200 1100
+-----------!-------/--t- - --illl----+--------1
, I
+--------+-------.~1----~----4---------~
l 1 '
1 t7 1---1---:1J1:-_ --+--------j
/ I 1000 -
+----
! 1f/
900 - -
I I
Qjj 800 - - 1-I
-m I I
~.
700 +--------~----------~1*---------FW-------------------~
ii 600
,/
- ,'/+-------;
A'.*
-#-+1"1~1--------------------j
- 1
~
~~
sao 400
- ------4-----~
- I
- - - - - -1'- --'l. l . - - - - - - - -
- t--1----*'-----~--------
300 l_,
~i~ ~----
p 200 r--------+---1~ Minimum Bolt-Up Temperature 90"F=
100 ----+-- 4W--l-----+---f=======j Minimum RPV Pressure =-14.7 psig O -------~----~-+----~--~====T 1 ========~
i 0 1 0 200 2 ;o Minimum Reactor Vessel Metal Temperature (°F)
Hatch Unit 2 PTLR Revision 0 Page 17 of40 Figure 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY Curve C- Core Critical, Composite Curves
- - Beltline ----Bottom Head - - Non-Belt line Overall 1300 , - - - - - - - - , . . . - - - - - - . - - - - - - - , - - - - - . , ., - --,,-----,....---~
I 1
f I 1200 - - - - - - - i - - - - - - t - - - - - - + - - - 1---1---f.ll--- - - 1 f I I I 1100 _,___ _ _ _ + - - - - - - f - - ~~-1---!ff-----l I
I I I
1000 ------+------+-----~--/--'-f--!JJ!------1 I I I
I 900 "iii 800
'Vi
.S:
!> 700 -
I
~ 600 -
- =*e
- 1 500 - - _ _ _.,.
i GJ
.t 400 200 Minimum Bolt-Up Temperature= 90°F 100 Minimum RPV Pressure= -14.7 psig 00~=:========T=====~~
II I_Ll~~
r~-----2+ ~------2~~
1~ I Minimum Reactor Vessel Metal Temperature ('F)
Hatch Unit 2 PTLR Revision 0 Page 18 of40 Figure 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY Curve A
- Pressure Test, Composite Curves
-Beltllne ---*Bottom Head - - Non-Beltllne Overall 2300 I
- I l j' I I /I J
1200 ! I I / I
/ I 1100 I
/ -
I ,' I
,,~:
I 4r I
1000 - 1- -
,.. I l
1_ :7 I I I
+
900 I
I I / /
l I r-800
_ I ;. ~
J_~* J l . A I I ;
700 : 1//
I I
I I I lI 11 i/ li 600 i 500 -
I I
r:
r-400 300 - ~~Ji 200 I Minimum Bolt-Up Temperature= 90°F 100 I Minimum RPV Pressure = -14.7 psig I
1r 0
0 100 200 2i0 T Minimum RftctarV....I Mlltlll Temperature rF)
Hatch Unit 2 PTLR Revision 0 Page 19 of 40 Figure 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY Curve B - Core Not Critical, Composite Curves r
- - Beltline ----Bottom Head - - Non-Belt llne Overall 1300 I -
I I I I
I I I I 1- 1 ~
I I
1200 I I
- I I
I 1100 _f I
I I II l I
I I
- - I I
I I I
1000 I I
r I
I II 900 I 0
I I I
, I I
1./
QO 800 I I I -
~-
- I
, I
~ I I /
1 700
':l' i Cll
> I
~ 1/
" I ,,
I J
(
. I I j I
Cll 600 '
a:
c I I
- e I
- 1 500 Cll
~
'"Cll I
~ 400 300 -~-L--~~ _____,
I#
II
~*
200 Minimum Bolt-Up 100 I II Temperature= 90°F Minimum RPV ~
0 I II 1_.
Pressure= -14.7 psig I I 0 so 100 150 200 250 Minimum Reactor Vessel Metal Temperature ('F)
Hatch Unit 2 PTLR Revision 0 Page 20 of40 Figure 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY Curve C- Core Critical, Composite Curves
- Bettline ----Bottom Head - - Non-Beltline Overall 1300 ...--
I 1 /
I 1200 I --t~-----r'-1-t--;f.----f I
I 1100 r- -
1000 ~
900 - --~
Dii 800
'ii'i L
.s:
l Ql 700
~ ,_
a: 600 -
Ql -- -
- e I '
- 1 500 I
I I
g Ql I
.s:"'
I Ql I 400 - i--
300 ,___
200 J Minimum Bolt-Up Temperature =90°F 100~==========~~1-+--------t--------+-------~
M inimum RPV Pressure= -14.7 psig 0 2=========~====~~_~-L----~------~---~
1 1 0 0 1 200 250 Minimum Reactor Vessel Metal Temperature ("F)
Hatch Unit 2 PTLR Revision 0 Page 21 of 40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Beltline Region Curve A -Pressure Test P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 506.7 98.5 554.7 105.7 602.7 112.0 650.7 117.6 698.8 122.7 746.8 127.3 794.8 131.5 842.8 135.3 890.8 138.9 938.8 142.3 986.8 145.4 1034.9 148.4 1082.9 151.2 1130.9 153.8 1178.9 156.3 1226.9 158.7 1274.9 161.0 1322.9
Hatch Unit 2 PTLR Revision 0 Page 22 of40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY (continued)
Bottom Head Region Curve A- Pressure Test P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 112.5 1166.6 115.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A -Pressure Test P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 1310.2
Hatch Unit 2 PTLR Revision 0 Page 23 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY Beltline Region Curve 8 - Core Not Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 248.1 101.1 296.8 110.2 345.5 117.9 394.2 124.6 442.9 130.5 491.6 135.7 540.3 140.5 589.0 144.9 637.7 148.9 686.4 152.6 735.1 156.0 783.8 159.2 832.5 162.3 881.2 165.1 929.9 167.8 978.6 170.4 1027.3 172.8 1076.0 175.1 1124.7 177.4 1173.4 179.5 1222.1 181.5 1270.8 183.5 1319.5
Hatch Unit 2 PTLR Revision 0 Page 24 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY (continued)
Bottom Head Region Curve 8 - Core Not Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 110.1 739.9 114.0 788.4 117.7 836.8 121.1 885.2 124.3 933.6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9
Hatch Unit 2 PTLR Revision 0 Page 25 of40 Table 2: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 37 EFPY (continued)
Non-Beltline Region Curve 8 - Core Not Critical P-T Curve P-T Curve Temperature Pressure
- F psi 90.0 0.0 90.0 312.6 150.0 312.6 150.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 26 of40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 140.6 112.2 189.6 127.5 238.7 139.2 287.7 148.7 336.8 156.7 385.8 163.6 434.9 169.6 483.9 175.0 533.0 179.9 582.0 184.3 631.1 188.4 680.1 192.1 729.2 195.6 778.2 198.9 827.3 202.0 876.3 204.9 925.4 207.6 974.4 210.2 1023.5 212.7 1072.5 215.0 1121.6 217.2 1170.6 219.4 1219.7 221.5 1268.7 223.4 1317.8
Hatch Unit 2 PTLR Revision 0 Page 27 of 40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478.1 127.5 527.5 133.7 576.8 139.3 626.1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 1316.4
Hatch Unit 2 PTLR Revision 0 Page 28 of40 Table 3: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 37 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.2 193.0 101.6 232.9 110.6 272.7 118.2 312.6 190.0 312.6 190.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 29 of40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 481.5 99.5 530.6 107.5 579.8 114.3 628.9 120.4 678.1 125.8 727.3 130.7 776.4 135.1 825.6 139.2 874.7 142.9 923.9 146.4 973.1 149.7 1022.2 152.8 1071.4 155.7 1120.6 158.4 1169.7 161.0 1218.9 163.5 1268.0 165.8 1317.2
Hatch Unit 2 PTLR Revision 0 Page 30 of40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY (continued)
Bottom Head Region Curve A -Pressure Test P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 112.5 1166.6 115.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A -Pressure Test P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 1310.2
Hatch Unit 2 PTLR Revision 0 Page 31 of40 Table 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY Beltline Region Curve 8 - Core Not Critical P-TCurve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 229.2 102.4 278.6 112.2 328.0 120.5 377.4 127.6 426.8 133.8 476.2 139.3 525.6 144.3 575.0 148.8 624.4 153.0 673.8 156.8 723.2 160.3 772.6 163.7 822.0 166.8 871.4 169.7 920.8 172.5 970.2 175.1 1019.6 177.6 1069.0 180.0 1118.4 182.2 1167.8 184.4 1217.2 186.5 1266.6 188.5 1316.0
Hatch Unit 2 PTLR Revision 0 Page 32 of40 Table 5: HNP-2 P-T Curve B (Normal Operation- Core Not Critical) for 50.1 EFPY (continued)
Bottom Head Region Curve B - Core Not Critical P-T Curve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 110.1 739.9 114.0 788.4 117.7 836.8 121.1 885.2 124.3 933 .6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9
Hatch Unit 2 PTLR Revision 0 Page 33 of 40 Table 5: HNP-2 P-T Curve B (Normal Operation -Core Not Critical) for 50.1 EFPY (continued)
Non-Beltline Region Curve 8 - Core Not Critical P*T Curve P-T Curve Temperature Pressure "F psi 90.0 0.0 90.0 312.6 150.0 312.6 150.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 34 of40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY Beltline Region Curve C- Core Critical P-T Curve P-T Curve Temperature Pressure
- F psi 90.0 0.0 90.0 132.1 114.2 181.4 130.5 230.8 142.7 280.1 152.5 329.5 160.7 378.8 167.8 428.1 173.9 477.5 179.4 526.8 184.4 576.2 188.9 625.5 193.0 674.8 196.9 724.2 200.4 773.5 203.7 822.9 206.8 872.2 209.7 921.6 212.5 970.9 215.1 1020.2 217.6 1069.6 220.0 1118.9 222.2 1168.3 224.4 1217.6 226.5 1266.9 228.5 1316.3
Hatch Unit 2 PTLR Revision 0 Page 35 of 40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY (continued)
Bottom Head Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478.1 127.5 527.5 133.7 576.8 139.3 626.1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 1316.4
Hatch Unit 2 PTLR Revision 0 Page 36 of40 Table 6: HNP-2 P-T Curve C (Normal Operation- Core Critical) for 50.1 EFPY (continued)
Non-Beltline Region Curve C- Core Critical P-TCurve P-TCurve Temperature Pressure "F psi 90.0 0.0 90.2 193.0 101.6 232.9 110.6 272.7 118.2 312.6 190.0 312.6 190.0 1313.5
Hatch Unit 2 PTLR Revision 0 Page 37 of40 Table 7: Hatch Unit 2 ART Table for 37 EFPY Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNOT("F) ARTNDT Margin Terms ARTNOT Factor ("F)
Cu(wt%1 Nl (wt %1 ("F) GJ("F) a~ rfl rFJ Lower Shell #1 G-6603-1 C8553-2 - -20 0.08 0.58 51 24.3 0.0 12.2 28.6 II Lower Shell #2 G-&603-2 C8553-1 24 0.08 0.58 51 24.3 0.0 12.2 72.6 Lower Shell#3 G-6603-3 C8571-1 - 0 0.08 0.53 51 24.3 0.0 12.2 48.6 Lower-In!. Shell #1 G-6602-2 C8554-1 . -20 0.08 0.57 51 27.6 0.0 13.8 35.2 Lower-In! Shell #2 G-6602-1 C8554-2 - -10 0.08 0.58 51 27.6 0.0 13.8 45.2 Lower-In! Shell #3 G-6601-4 C8579-2 - -4 0.11 0.48 73 39.5 0.0 17.0 69.5 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNOT("F) ARTNOT Margin Terms ARTNoT Factor ("F)
Cu(wt%1 Nl(wt%1 ("F) a! ("F) a~ (*F) rFl II Lower Long. Weld 101-842 10137 - -50 0.216 0.043 98 45.1 0.0 22.5 40.2 Lower Int. Long Weld 101-834 51874 - *50 0.147 0.037 68 30.0 0.0 15.0 10.0 Lower
- Lower Int. Girth Weld 301-871 4P6052 - -50 0.047 0.049 31 15.2 0.0 7.6 -19.6 Fluence Data Wall Thlckness(ln.j Flue nee at ID Attenuation, Fluence Factor, FF Flue nee @11/4t (n/cm 2 )
Full 1/4t (n/cm 2 ) 1/4t = e~.24x fC0.211.0.101og I)
Location Lower Shell #1 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Lower Shell #2 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Jl Lower Shell#3 6.375 1.594 1.95E+18 0.682 1.33E+18 0.477 Lower-In!. Shell #1 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower-In! Shell #2 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower-lnt Shell #3 5.375 1.344 2.45E+18 0.724 1.77E+18 0.541 Lower Long. Weld 6.375 1.594 1.81E+18 0.682 1.23E+18 0.460 Lower Int. Long Weld 5.375 1.344 1.55E+18 0.724 1.13E+18 0.441 Lower
- Lower Int. Girth Weld 5.375 1.344 1.95E+18 0.724 1.41E+18 0.490
Hatch Unit 2 PTLR Revision 0 Page 38 of 40 Table 8: Hatch Unit 2 ART Table for 50.1 EFPY Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNori"F) ARTNoT Margin Terms ARTNoT Factor rFl Cu (wt%1 Nl (wt%1 ("F) at ("F) a6 rFl rFl Lower Shell #1 G-6603-1 C8553-2 - -20 0.08 0.58 51 27.6 0.0 13.8 35.2
- 0.0 II Lower Shell #2 G-6603-2 C8553-1 24 0.08 0.58 51 27.6 13.8 79.2 Lower Shell#3 G-6603-3 C8571-1 - 0 0.08 0.53 51 27.6 0.0 13.8 55.2 Lower-In!. Shell #1 G-6602-2 C8554-1 - -20 0.08 0.57 51 31.2 0.0 15.6 42.4 Lower-In! Shell #2 G-6602-1 C8554-2 - -10 0.08 0.58 51 31.2 0.0 15.6 52.4 Lower-In! Shell #3 G-6601-4 C8579-2 - -4 0.11 0.48 73 44.6 0.0 17.0 74.6 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. lnltia I RTNor ("F) ARTNoT Margin Terms ARTNoT Factor ("F)
Cu (wt%1 Nl (wt%1 rFl at rFl a6 rFI rFl
!I Lower Long. Weld 101-842 10137 - -50 0.216 0.043 98 51.4 0.0 25.7 52.8 Lower Int. Long Weld 101-834 51874 - -50 0.147 0.037 68 34.4 0.0 17.2 18.8 Lower - Lower Int. Girth Weld 301-871 4?6052 - -50 0.047 0.049 31 17.2 0.0 8.6 -15.6 Adjustments for 1/4t Chemistry Chemistry Description Code No. Heat No. Flux Type & Lot No. Initial RTNorrFl ARTNoT Margin Terms ARTNor Factor ("F)
Cu (wt %1 j; Nl (wt %11 ("F) at ("F) a6 ("F) rFl Recirculation Inlet Nozzle G-6607 Q2Q24W - 10 0.180 0.810 141 11 .9 0.0 5.9 33.7 Fluence Data Wall Thickness (ln.) Fluence at 10 Attenuation, Fluence Factor, FF Flue nee @11/4t (n/cm 2) ,co.a.o.101og II Full 1/4t (nlcm 2 ) 1/4t = e-0.24x Location Lower Shell #1 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 il Lower Shell #2 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 Lower Shell#3 6.375 1.594 2.60E+18 0.682 1.77E+18 0.541 Lower-In!. Shell #1 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611 ii:
Lower-In! Shell #2 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611 Lower-In! Shell #3 5.375 1.344 3.28E+18 0.724 2.38E+18 0.611
!I Lower Long . Weld 6.375 1.594 2.42E+18 0.682 1.65E+18 0.525 Lower Int. Long Weld 5.375 1.344 2.10E+18 0.724 1.52E+18 0.506 Lower- Lower Int. Girth Weld 5.375 1.344 2.60E+18 0.724 1.88E+18 0.555 Recirculation Inlet Nozzle 6.375 1.594 1.00E+17 0.682 6.82E+16 0.084
Hatch Unit 2 PTLR Revision 0 Page 39 of40 Table 9: Hatch Unit 2 Summary of Nozzle Stress Intensity Factors Applied Pressure, Thermal, K,t Thermal, Ktt Nozzle Ktp-app (450"F shock) (100 *F/hr Plate)
Feed water 78.9 46.8 12.9 WLI 80.0 N/A 19.9 Notes:
- 1. K1 in units of ksi-in°*5
Hatch Unit 2 PTLR Revision 0 Page 40 of 40 Appendix A HATCH UNIT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, one surveillance capsule was removed from the Hatch Nuclear Plant Unit 2 (HNP-2) reactor vessel in 1989 following cycle 8 [18]. 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 [18].
Southern Nuclear Operating Company committed to use the ISP in place of its existing surveillance programs 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 [19]. 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 [20]. HNP-2 continues to be a host plant under the ISP
[12]. Two more HNP-2 capsules are scheduled to be removed and tested under the ISP in approximately 2017 and 2027.