RA-19-0359, License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System
| ML20134H888 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/12/2020 |
| From: | Maza K Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-19-0359 | |
| Download: ML20134H888 (33) | |
Text
Kim Maza Site Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562 10 CFR 50.90 May , 2020 RA-19-0359 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63
Subject:
License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) hereby requests an amendment to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Renewed Facility Operating License. The proposed change revises HNP Technical Specification (TS) 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System, to reflect an update to the pressure and temperature limit curves in Figures 3.4-2 (Reactor Coolant System Cooldown Limitations) and 3.4-3 (Reactor Coolant System Heatup Limitations). The proposed change also reflects that the revised HNP pressure and temperature limit curves in TS Figures 3.4-2 and 3.4-3 will be applicable until 55 effective full power years (EFPY). Lastly, the proposed change revises TS Figure 3.4-4 (Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System) to reflect that the setpoint values are based on 55 EFPY reactor vessel data.
The proposed change is necessary because Duke Energy identified after the removal and examination of reactor pressure vessel surveillance capsule Z at the end of cycle 21 that the existing HNP TS Figures 3.4-2 and 3.4-3 are non-conservative. Current plant operations are administratively controlled consistent with Nuclear Regulatory Commission (NRC)
Administrative Letter (AL) 98-10, Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety.
The Enclosure provides a description and assessment of the proposed change. Attachment 1 provides the existing HNP TS pages marked to show the proposed change. Attachment 2 provides existing TS Bases pages marked to show the proposed change for information only.
The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The basis for this determination is included in the Enclosure.
Duke Energy requests approval of the proposed amendment within one year of the date this submittal is accepted by the NRC staff for review. Once approved, Duke Energy will implement the license amendment within 120 days.
There are no regulatory commitments contained in this submittal.
U.S. Nuclear Regulatory Commission RA-19-0359 Page2 In accordance with 10CFR 50.91, Duke Energy is notifying the State of North Carolina of this license amendment request by transmitting a copy of this letter and enclosure to the designated State Official.
If there are any questions or if additional information is needed, please contact Mr. Art Zaremba, Manager - Nuclear Fleet Licensing at 980-373-2062 or Arthur.Zaremba@duke-energy.com.
I declare under penalty of perjury that the foregoing is true and correct. Executed on May 12, 2020.
Sincerely, Kim Maza Site Vice President Harris Nuclear Plant
Enclosure:
Description and Assessment of the Proposed Change Attachments:
- 1. Technical Specifications Markup
- 2. Technical Specifications Bases Markup cc (with Enclosure/Attachments):
L. Dudes, NRC Regional Administrator, Region II J. Zeiler, NRC Senior Resident Inspector, HNP T. Hood, NRC Project Manager, HNP W. L. Cox, Ill, Section Chief N.C. DHSR
U.S. Nuclear Regulatory Commission RA-19-0359 Page 2 In accordance with 10 CFR 50.91, Duke Energy is notifying the State of North Carolina of this license amendment request by transmitting a copy of this letter and enclosure to the designated State Official.
If there are any questions or if additional information is needed, please contact Mr. Art Zaremba, Manager - Nuclear Fleet Licensing at 980-373-2062 or Arthur.Zaremba@duke-energy.com.
I declare under penalty of perjury that the foregoing is true and correct. Executed on May XX, 2020.
Sincerely, Kim Maza Site Vice President Harris Nuclear Plant
Enclosure:
Description and Assessment of the Proposed Change Attachments:
- 1. Technical Specifications Markup
- 2. Technical Specifications Bases Markup cc (with Enclosure/Attachments):
L. Dudes, NRC Regional Administrator, Region II J. Zeiler, NRC Senior Resident Inspector, HNP T. Hood, NRC Project Manager, HNP W. L. Cox, III, Section Chief N.C. DHSR
U.S. Nuclear Regulatory Commission RA-19-0359 Page 1 ENCLOSURE Description and Assessment of the Proposed Change
Subject:
License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System
- 1.
SUMMARY
DESCRIPTION
- 2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change
- 3. TECHNICAL EVALUATION
- 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions
- 5. ENVIRONMENTAL CONSIDERATION
- 6. REFERENCES ATTACHMENTS:
- 1. Technical Specifications Markup
- 2. Technical Specifications Bases Markup
U.S. Nuclear Regulatory Commission RA-19-0359 Page 2
- 1.
SUMMARY
DESCRIPTION Duke Energy Progress, LLC (Duke Energy) hereby requests an amendment to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Renewed Facility Operating License. The proposed change revises HNP Technical Specification (TS) 3/4.4.9, Pressure/Temperature Limits -
Reactor Coolant System to reflect an update to the pressure and temperature (P/T) limit curves in Figures 3.4-2 (Reactor Coolant System Cooldown Limitations) and 3.4-3 (Reactor Coolant System Heatup Limitations). The proposed change also reflects that the revised HNP P/T limit curves in TS Figures 3.4-2 and 3.4-3, as well as the existing power-operated relief valve (PORV) setpoints in Figure 3.4-4 (Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System), will be applicable until 55 effective full power years (EFPY).
There are no changes being proposed to the PORV setpoints in Figure 3.4-4.
- 2. DETAILED DESCRIPTION 2.1 System Design and Operation All components of the HNP Reactor Coolant System (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients and reactor trips. HNP is required to limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation.
The HNP TS contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The typical use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
Operating limits are established that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).
Both the HNP Updated Final Safety Analysis Report Section 5.3.2 and the TS 3/4.4.9 Bases provide additional details regarding the methodology that was used to develop the existing P/T limit curves that are contained in the HNP TS.
2.2 Current Technical Specifications Requirements HNP Limiting Condition for Operation (LCO) 3.4.9.1 (Applicability: Modes 1, 2 and 3) requires that the reactor coolant temperature and pressure (except for the pressurizer) be maintained in accordance with Figures 3.4-2 and 3.4-3 (P/T curves). LCO 3.4.9.2 (Applicability: Modes 4, 5 and 6 with reactor vessel head on) also requires that the reactor coolant temperature and pressure (except for the pressurizer) be maintained in accordance with Figures 3.4-2 and 3.4-3.
The maximum heatup and cooldown rates are specified in TS Table 4.4-6 for LCO 3.4.9.2.
LCO 3.4.9.1 and 3.4.9.2 limits apply to all components of the RCS, except the pressurizer, and define allowable operating regions and permit many operating cycles while providing a wide margin to non-ductile failure. Violating either LCO 3.4.9.1 or 3.4.9.2 limits would result in
U.S. Nuclear Regulatory Commission RA-19-0359 Page 3 placing the reactor vessel outside of the bounds of the stress analyses and could increase stresses in other RCPB components.
The existing HNP TS P/T limits curves in Figures 3.4-2 and 3.4-3 were approved by the issuance of Amendment No. 100 on July 28, 2000 (Reference 1).
TS Figure 3.4-4 provides the PORV setpoints for low temperature overpressure protection (LTOP) and the existing values are based on 36 EFPY reactor vessel data.
2.3 Reason for the Proposed Change By letter dated October 23, 2019 (Reference 2), Duke Energy submitted the summary technical report for the reactor pressure vessel (RPV) surveillance program (Framatome ANP-3798NP Revision 0) in accordance with 10 CFR 50, Appendix H requirements and communicated that a change to the HNP TS P/T limit curves would be required. The Framatome analysis of the surveillance capsule revealed that existing HNP TS Figures 3.4-2 and 3.4-3 are non-conservative. Specifically, the analysis indicates a shift in the Adjusted Reference Temperature (ART) of 14oF for the 1/4T location and 11oF for the 3/4T location, which has the effect of shifting the TS heatup and cooldown curves to the right.
Although the operational curves used by the main control room to operate the plant are conservative with respect to the P/T limit curves developed as part of the capsule Z testing project and thus remain valid, the TS P/T limit curves (i.e., Figures 3.4-2 and 3.4-3) are non-conservative and require a revision to satisfy regulatory requirements.
2.4 Description of the Proposed Change TS Figure 3.4-2, Reactor Coolant System Cooldown Limitations - Applicable to Up to 36 EFPY, is revised as follows:
The existing RCS cooldown limitations curves are superseded entirely by new curves applicable up to 55 EFPY.
The cooldown limitations curve for a rate of 100 /HR is removed.
The RTNDT [reference nil-ductility temperature] at 1/4 T value of 191 212 in the Material Property Bases.
The RTNDT at 3/4 T value of 179 198 in the Material Property Bases.
The title of Figure 3.4-2 is revised to state Reactor Coolant System Cooldown Limitations - Applicable Up to 55 EFPY.
U.S. Nuclear Regulatory Commission RA-19-0359 Page 4 TS Figure 3.4-3, Reactor Coolant System Heatup Limitations - Applicable Up to 36 EFPY, is revised as follows:
The existing RCS heatup limitations curves are superseded entirely by new curves applicable up to 55 EFPY.
The RTNDT at 1/4 T value of 191 212 in the Material Property Bases.
The RTNDT at 3/4 T value of 179 198 in the Material Property Bases.
The title of Figure 3.4-3 is revised to state Reactor Coolant System Heatup Limitations -
Applicable Up to 55 EFPY.
TS Figure 3.4-4, Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System, is revised as follows:
The note *VALUES BASED ON 36 EFPY REACTOR VESSEL DATA is revised to state *VALUES BASED ON 55 EFPY REACTOR VESSEL DATA.
The TS markup provided in Attachment 1 reflects the proposed changes described above.
The proposed changes are supported by changes to the TS Bases. In addition to reflecting the proposed changes to the TS, the TS 3/4.4.9 Bases are revised for clarity and consistency. The regulation at 10 CFR 50.36 states A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications. Changes to the TS Bases will be made in accordance with the Technical Specifications Bases Control Program following approval of the requested amendment. The proposed TS Bases changes in Attachment 2 are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commissions Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 (58 FR 39132).
Therefore, the HNP TS Bases changes are provided for information and approval of the TS Bases is not requested.
- 3. TECHNICAL EVALUATION Pressure/Temperature Limit Curves Development The proposed bounding TS cooldown (Figure 3.4-2) and heatup (Figure 3.4-3) limitation curves presented in Attachment 1 are derived from uncorrected P/T limits calculated for the HNP reactor vessel through 60 years of operation (55 EFPY). For various heatup and cooldown rates ( F/hr), the P/T limits were calculated for the reactor vessel beltline shell, nozzle and closure head locations for normal heatup and normal cooldown conditions as well as ISLH test conditions. P/T limits were calculated in accordance with the requirements of the 2007 Edition with 2008 Addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Appendix G, which permits the use of the K Ic fracture
U.S. Nuclear Regulatory Commission RA-19-0359 Page 5 curve. The requirements of 10 CFR 50, Appendix G (Fracture Toughness Requirements) were also complied with in calculating the P/T limit curves through 55 EFPY.
Heatup and cooldown limit curves were calculated using the most limiting value of RT NDT corresponding to the limiting material in the beltline region of the RPV. Limiting RT NDT is also referred to as limiting ART. Per HNP TS Figures 3.4-2 and 3.4-3, the controlling material (i.e.,
the limiting material) is Plate B4197-2, which is the Intermediate Shell Plate in the beltline region of the RPV.
The materials outside of the traditional beltline region which are expected to receive fluence values greater than 1x1017 n/cm2 were evaluated. The evaluation found 12 reactor vessel locations outside the traditional beltline region with 55 EFPY fluence values greater than 1x10 17 n/cm2. The locations were:
- 2. Upper Shell,
- 3. Inlet Nozzle Weld 15-A, 15-B, 15-C,
- 4. Inlet Nozzle,
- 5. Outlet Nozzle Weld 16-A, 16-B, 16-C,
- 6. Outlet Nozzle,
- 7. Upper Shell Longitudinal Welds BE/BF,
- 8. Lower to Bottom Head Circumferential Weld,
- 9. Torus Shell,
- 12. Bottom Head Dome.
The materials at these twelve locations were analyzed for ART and were found not to be limiting. The controlling material for this analysis remains the intermediate shell plate, heat number B4197-2, which is a location inside the traditional beltline region. Thus, limiting ART values for the 1/4T and 3/4T locations of the reactor vessel beltline wall were obtained for Intermediate Shell Plate B4197-2 with fluence projections up to 55 EFPY and are reported in Table 1 below.
The Regulatory Guide (RG) 1.99, Revision 2 methodology was used along with fast neutron fluence estimates on the reactor vessel to calculate ART values in Table 1. A NRC-approved Framatome calculation-based fluence methodology (BAW-2241P-A) that is in accordance with the requirements of RG 1.190 was used to predict the fluence in the HNP reactor vessel. The cumulative fast (E > 1.0 MeV) fluence for several locations of interest at 55 EFPY at the wetted surface (i.e., inner clad surface) are provided in Table 2. The cumulative fast (E > 1.0 MeV) fast fluence for several locations of interest at 55 EFPY at the inner RPV surface (i.e., clad/vessel interface) are provided in Table 3.
U.S. Nuclear Regulatory Commission RA-19-0359 Page 6 Table 1: Limiting ART Values for HNP Reactor Vessel Components at 55 EFPY Vessel Component Material Identification Wall Location Limiting RTNDT 1/4T 211.1°F (rounded to 212°F)
Beltline Region (7.75") Intermediate Shell Base Metal/Axial Flaw Plate Heat B4197-2 3/4T 196.9°F (rounded to 198°F) 134.9°F (bounded by base Intermediate Shell to 1/4T Beltline Region (7.75") metal axial flaw)
Lower Shell Weld AB Circumferential Weld 118.3°F (bounded by base (Wire Heat 5P6771) 3/4T metal axial flaw)
Nozzle Shell Region 1/4T 120.0°F (rounded to 121°F)
Upper Shell Plate (9.25")
Heat C0123-1 3/4T 93.5°F (rounded to 94°F)
Base Metal/Axial Flaw Table 2: Fast Neutron Fluence (E > 1 MeV) for the Reactor Vessel Wetted Inside Surface 55 EFPY Fluence Description (n/cm2)
Forgings / Plates (original 40-year beltline)
Intermediate Shell Plates (ISP) 6.97E+19 Lower Shell Plates (LSP) 6.79E+19 Wetted Surface Max 6.97E+19 Welds (original 40-year beltline)
Upper Shell Plates (USP) to ISP Circular Welds (AC) 3.49E+18 ISP to LSP Circular Weld (AB) 6.77E+19 ISP Longitudinal Weld (BC/BD) 2.60E+19 LSP Longitudinal Weld (BA/BB) 2.53E+19 Extended Beltline USP 3.04E+18 USP Longitudinal Weld (BE/BF) 3.49E+18 Inlet Nozzle Lower Weld (15-A, 15-B, 15-C) 3.87E+17 Outlet Nozzle Lower Weld 1.86E+17 LSP to Bottom Head Circular Weld 1.65E+18
U.S. Nuclear Regulatory Commission RA-19-0359 Page 7 Table 3: Fast Neutron Fluence (E > 1 MeV) for the Reactor Vessel Clad-Base Metal Interface Description 55 EFPY Fluence (n/cm2)
Forgings / Plates (original 40-year beltline)
Intermediate Shell Plates (ISP) 6.87E+19 Lower Shell Plates (LSP) 6.70E+19 Inner RPV Surface Max 6.87E+19 Welds (original 40-year beltline)
Upper Shell Plates (USP) to ISP Circular Welds (AC) 3.46E+18 ISP to LSP Circular Weld (AB) 6.68E+19 ISP Longitudinal Weld (BC/BD) 2.57E+19 LSP Longitudinal Weld (BA/BB) 2.51E+19 Extended Beltline USP 3.00E+18 USP Longitudinal Weld (BE/BF) 3.46E+18 Inlet Nozzle Lower Weld (15-A, 15-B, 15-C) 3.83E+17 Outlet Nozzle Lower Weld 1.84E+17 LSP to Bottom Head Circular Weld 1.65E+18 1/4T at Maximum Peak Location 4.04E+19 3/4T at Maximum Peak Location 9.99E+18 The proposed TS P/T limit curves for HNP provided in Attachment 1 (i.e., the proposed change) were drawn using data points from a Framatome technical report prepared for HNP. The data points for the 55 EFPY P/T limit curves are provided in Tables 4, 5 and 6 below. The 55 EFPY P/T limit curves are based on the limiting beltline material ART values, which are affected by both the fluence and the initial material properties of that material. The P/T limits are not adjusted for instrument error.
Table 4: Uncorrected Heatup P/T Limits 55 EFPY (pressure in psig)
Heatup Rates, °F/hr Temp,
°F 0 5 10 15 20 30 40 50 100 75 621 621 621 621 621 621 621 621 554 80 621 621 621 621 621 621 621 621 554 85 621 621 621 621 621 621 621 621 554 90 621 621 621 621 621 621 621 621 554 95 621 621 621 621 621 621 621 621 554 100 621 621 621 621 621 621 621 621 554 105 621 621 621 621 621 621 621 621 554 110 621 621 621 621 621 621 621 621 554 115 621 621 621 621 621 621 621 621 554 120 621 621 621 621 621 621 621 621 554 125 704 702 700 699 693 666 643 621 554 130 711 709 707 706 701 673 648 621 554 135 719 717 715 713 710 680 654 631 554 140 728 726 723 721 719 689 661 636 554
U.S. Nuclear Regulatory Commission RA-19-0359 Page 8 Table 4: Uncorrected Heatup P/T Limits 55 EFPY (pressure in psig)
Heatup Rates, °F/hr Temp,
°F 0 5 10 15 20 30 40 50 100 145 738 735 733 730 728 699 669 643 554 150 749 746 743 740 738 709 678 650 554 155 761 758 755 752 749 721 689 659 555 160 774 771 767 764 761 734 700 669 557 165 789 785 781 778 774 748 712 681 560 170 805 801 797 793 789 765 727 692 565 175 823 818 814 809 805 783 743 706 570 180 843 838 833 828 823 803 761 722 577 185 865 859 853 848 843 825 781 740 583 190 889 883 877 870 864 849 803 760 593 195 916 909 902 895 889 876 827 781 604 200 946 938 930 923 915 901 854 805 617 205 979 970 961 953 945 929 883 832 632 210 1015 1005 996 986 977 960 916 861 648 215 1055 1044 1033 1023 1013 994 953 894 667 220 1099 1087 1075 1064 1053 1032 993 931 688 225 1148 1135 1122 1109 1097 1074 1037 970 711 230 1202 1187 1173 1159 1146 1120 1085 1014 736 235 1262 1245 1229 1214 1199 1171 1139 1063 765 240 1328 1310 1292 1275 1258 1227 1197 1116 798 245 1401 1380 1361 1342 1324 1289 1256 1175 833 250 1481 1459 1437 1416 1396 1357 1321 1240 871 255 1570 1545 1521 1498 1476 1433 1392 1312 914 260 1669 1641 1614 1589 1564 1516 1472 1391 961 265 1777 1747 1717 1689 1661 1608 1559 1478 1013 270 1898 1864 1831 1799 1769 1710 1655 1575 1071 275 2031 1993 1957 1922 1888 1823 1762 1681 1134 280 2177 2136 2096 2057 2019 1947 1880 1798 1204 285 2340 2294 2249 2206 2164 2084 2010 1926 1281 290 2519 2468 2419 2371 2324 2236 2153 2069 1367 295 2717 2661 2606 2553 2502 2404 2312 2226 1461 300 2858 2858 2813 2754 2697 2589 2487 2392 1564 305 2858 2858 2858 2858 2858 2793 2680 2575 1677 310 2858 2858 2858 2858 2858 2858 2858 2777 1802 315 2858 2858 2858 2858 2858 2858 2858 2858 1940 320 2858 2858 2858 2858 2858 2858 2858 2858 2092 325 2858 2858 2858 2858 2858 2858 2858 2858 2260 330 2858 2858 2858 2858 2858 2858 2858 2858 2445 335 2858 2858 2858 2858 2858 2858 2858 2858 2649 340 2858 2858 2858 2858 2858 2858 2858 2858 2858 345 2858 2858 2858 2858 2858 2858 2858 2858 2858 350 2858 2858 2858 2858 2858 2858 2858 2858 2858
U.S. Nuclear Regulatory Commission RA-19-0359 Page 9 Table 5: Uncorrected Cooldown P/T Limits 55 EFPY (pressure in psig)
Cooldown Rates, °F/hr Temp,
°F 0 5 10 15 20 30 40 50 100 75 621 621 621 621 609 584 558 532 399 80 621 621 621 621 612 587 561 535 402 85 621 621 621 621 615 590 564 538 406 90 621 621 621 621 619 593 568 542 411 95 621 621 621 621 621 597 572 546 416 100 621 621 621 621 621 602 576 551 422 105 621 621 621 621 621 606 581 556 428 110 621 621 621 621 621 612 587 562 436 115 621 621 621 621 621 618 593 569 440 120 697 685 673 661 649 625 601 576 450 125 704 692 680 668 656 632 608 584 460 130 711 699 688 676 664 641 617 593 472 135 719 708 696 684 673 650 627 602 485 140 728 717 705 694 683 659 636 613 500 145 738 727 716 704 693 670 648 625 517 150 749 738 727 716 705 683 661 639 535 155 761 750 739 729 718 697 676 655 555 160 774 764 753 743 733 712 692 672 577 165 789 779 769 759 749 729 710 691 605 170 805 795 786 776 767 748 730 712 632 175 823 814 805 796 787 769 752 735 662 180 843 834 826 817 809 792 776 761 696 185 865 857 849 841 833 818 805 791 733 190 889 882 874 867 860 847 834 822 775 195 916 909 903 897 890 878 867 857 821 200 946 940 934 928 923 913 904 895 872 205 979 973 969 964 959 951 944 938 928 210 1015 1011 1007 1003 1000 994 989 985 992 215 1055 1052 1049 1046 1044 1040 1038 1037 1055 220 1099 1097 1095 1094 1093 1092 1093 1095 1099 225 1148 1147 1147 1147 1147 1148 1148 1148 1148 230 1202 1202 1202 1202 1202 1202 1202 1202 1202 235 1262 1262 1262 1262 1262 1262 1262 1262 1262 240 1328 1328 1328 1328 1328 1328 1328 1328 1328 245 1401 1401 1401 1401 1401 1401 1401 1401 1401 250 1481 1481 1481 1481 1481 1481 1481 1481 1481 255 1570 1570 1570 1570 1570 1570 1570 1570 1570 260 1669 1669 1669 1669 1669 1669 1669 1669 1669 265 1777 1777 1777 1777 1777 1777 1777 1777 1777 270 1898 1898 1898 1898 1898 1898 1898 1898 1898 275 2031 2031 2031 2031 2031 2031 2031 2031 2031 280 2177 2177 2177 2177 2177 2177 2177 2177 2177 285 2340 2340 2340 2340 2340 2340 2340 2340 2340
U.S. Nuclear Regulatory Commission RA-19-0359 Page 10 Table 5: Uncorrected Cooldown P/T Limits 55 EFPY (pressure in psig)
Cooldown Rates, °F/hr Temp,
°F 0 5 10 15 20 30 40 50 100 290 2519 2519 2519 2519 2519 2519 2519 2519 2519 295 2717 2717 2717 2717 2717 2717 2717 2711 2551 300 2858 2844 2830 2815 2801 2771 2741 2711 2551 305 2858 2844 2830 2815 2800 2771 2741 2710 2551 310 2858 2844 2830 2815 2800 2771 2741 2710 2550 315 2858 2844 2829 2815 2800 2770 2740 2709 2550 320 2858 2844 2829 2815 2800 2770 2740 2709 2550 325 2858 2844 2829 2815 2800 2770 2739 2706 2550 330 2858 2844 2829 2815 2800 2770 2737 2705 2551 335 2858 2844 2829 2815 2800 2768 2737 2705 2551 340 2858 2844 2829 2815 2799 2768 2736 2705 2555 345 2858 2844 2829 2814 2798 2768 2736 2704 2555 350 2858 2844 2829 2814 2798 2767 2736 2704 2555 Table 6: Uncorrected ISLH 10 F/HR for 55 EFPY (pressures in psig)
Temp, °F Heatup Cooldown 75 621 621 80 621 621 85 621 621 90 621 621 95 621 621 100 621 621 105 621 621 110 621 621 115 621 621 120 621 621 125 934 907 130 943 917 135 953 928 140 965 940 145 977 954 150 991 969 155 1006 986 160 1023 1004 165 1042 1025 170 1062 1048 175 1085 1073 180 1110 1101 185 1138 1132 190 1169 1166 195 1203 1203 200 1240 1245 205 1282 1291
U.S. Nuclear Regulatory Commission RA-19-0359 Page 11 Table 6: Uncorrected ISLH 10 F/HR for 55 EFPY (pressures in psig)
Temp, °F Heatup Cooldown 210 1327 1342 215 1378 1398 220 1434 1461 225 1496 1529 230 1564 1603 235 1639 1682 240 1722 1770 245 1815 1867 250 1916 1975 255 2029 2094 260 2153 2225 265 2290 2370 270 2441 2530 275 2609 2707 280 2794 2903 285 2999 3120 290 3225 3359 295 3475 3623 300 3751 3773 305 3811 3773 310 3811 3773 315 3811 3773 320 3811 3773 325 3811 3772 330 3811 3772 335 3811 3772 340 3811 3772 345 3811 3772 350 3811 3772 Note that the existing TS Figure 3.4-2 (Cooldown Limitations) contains a P/T limit curve for a cooldown rate of 100 F/HR. The proposed change (i.e., revised Figure 3.4-2) does not contain a curve for the cooldown rate of 100 F/HR because when the RCS fluid temperature is 290 F or greater, the allowable pressure is above the reactor vessel design pressure of 2485 psig.
Regulatory Issue Summary (RIS) 2014-11 Considerations 10 CFR 50, Appendix G requires that P/T limits be developed to bound all ferritic materials in the RPV. RIS 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components (Reference 3),
clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may define P/T curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure.
U.S. Nuclear Regulatory Commission RA-19-0359 Page 12 For HNP, thickness transitions (i.e., discontinuities) exist at the lower shell to Torus shell weld (lower transition) and in the upper shell plate to intermediate shell weld (upper transition). P/T limits, which include the impact of the structural discontinuities in the HNP reactor vessel shell upper and lower transition regions at 55 EFPY, were developed in accordance with the guidelines of the ASME Code,Section XI, Appendix G, 2007 Edition including Addenda through 2008. These P/T limits for the thickness transition sections of the reactor vessel were then compared against the P/T limits calculated for the reactor vessel beltline shell, nozzle and closure head locations to verify that they were not more limiting than the traditional beltline P/T limits. Duke Energy determined that the P/T limit curves for the transition regions are not more limiting than the traditional beltline P/T limits for any heatup and cooldown rates. Therefore, the transition region P/T limits are bounded by the P/T limits in Tables 4, 5 and 6 above.
Low Temperature Overpressure (LTOP) Setpoint Considerations Duke Energy has performed calculations to demonstrate that for the PORV setpoints in TS Figure 3.4-4, the predicted system pressure overshoots resulting from mass or heat input transients will not exceed the proposed HNP P/T limits presented in Tables 4 and 5 above and reflected in the Attachment 1 figures.
The LTOP analysis calculated a peak pressure of 559 psig (including location adjustment and instrument uncertainty). The calculation evaluated a mass input transient and a heat input transient with a single PORV opening in response to each transient. The calculation assumed the existing PORV setpoints in TS Figure 3.4-4. The LTOP analysis utilizes the new limiting pressures determined for the heatup and cooldown curves at 55 EFPY for the reactor vessel beltline region and vessel flange region to conclude that the vessel peak pressure will not be exceeded during low temperature operations. Therefore, the current PORV low-pressure setpoints in TS Figure 3.4-4 remain acceptable for plant operation to 55 EFPY.
The LTOP analysis also evaluated the enable temperature utilizing the 55 EFPY highest adjusted reference temperature. The current 325°F LTOP enable temperature was evaluated using the ASME Boiler and Pressure Vessel Code 2007 Edition including Addenda through 2008 Section XI Appendix G Article G-2215, and determined to remain acceptable.
- 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements and guidance documents are applicable to the proposed change.
10 CFR 50.36 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36(c)
TSs will include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) LCOs, (3) Surveillance Requirements (SRs), (4) design features; and (5) administrative controls.
HNP LCOs 3.4.9.1 (Applicability: Modes 1, 2 and 3) and 3.4.9.2 (Applicability: Modes 4, 5 and 6 with reactor vessel head on) limit the pressure and temperature changes during RCS heatup
U.S. Nuclear Regulatory Commission RA-19-0359 Page 13 and cooldown (i.e., to the right and below the P/T curves in Figures 3.4-2 and 3.4-3), to prevent non-ductile RPV failure. The proposed change revises the HNP P/T limit curves in TS and reflects that the curves are applicable until 55 EFPY. Based on the determination that the proposed TS Figures 3.4-2 and 3.4-3 are acceptable up to 55 EFPY (see Section 3 above),
Duke Energy concludes that HNP LCOs 3.4.9.1 and 3.4.9.2 will continue to meet the requirements of 10 CFR 50.36(c)(2)(i) with the proposed change.
10 CFR 50.60 Section 50.60 of 10 CFR, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, imposes fracture toughness and material surveillance program requirements, which are set forth in 10 CFR 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements. With the proposed change, HNP meets the requirements set forth in 10 CFR 50, Appendices G and H. Therefore, HNP also satisfies the requirements of 10 CFR 50.60 for the proposed change.
10 CFR 50, Appendix G Appendix G to 10 CFR 50 requires that the P/T limits for the facilitys RPV be at least as conservative as those obtained by following the linear elastic fracture mechanics methodology of Appendix G to Section XI of the ASME Code. Using the ART values, P/T limits curves were determined in accordance with the requirements of 10 CFR 50, Appendix G. Therefore, Duke Energy concludes for the proposed change that the HNP RPV will continue to meet RPV integrity regulatory requirements through 55 EFPY.
10 CFR 50, Appendix H Appendix H to 10 CFR 50 establishes requirements for each facility related to its RPV material surveillance. These regulatory requirements will continue to be met for the proposed change with the surveillance capsule removal schedule prescribed in Section 9.0 of Framatome report ANP-3798NP that was provided to the NRC staff in the Appendix H submittal dated October 23, 2019 (Reference 2).
RG 1.99, Revision 2 RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains guidance on methodologies the NRC considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. This RG was used along with fluence to calculate ART values for the HNP reactor vessel materials at 55 EFPY. Therefore, the proposed change has no effect on how Duke Energy applies RG 1.99, Revision 2 for HNP.
RG 1.190 RG. 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001, describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence. Framatome neutron transport evaluation methodologies (proprietary) utilized for the HNP neutron fluence evaluation followed the guidance of RG 1.190.
U.S. Nuclear Regulatory Commission RA-19-0359 Page 14 Regulatory Issue Summary (RIS) 2014-11 RIS 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Duke Energy appropriately considered RIS 2014-11 by performing analyses to include the impact of structural discontinuities in the HNP RPV shell upper and lower transition regions at 55 EFPY.
The proposed change does not affect plant compliance with any of the above regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.
4.2 Precedent The NRC has previously approved changes similar to the proposed change in this License Amendment Request for other nuclear power plants including:
H.B. Robinson Steam Electric Plant, Unit No. 2: Application dated November 2, 2015 (ADAMS Accession No. ML15307A069); NRC Safety Evaluation dated November 22, 2016 (ADAMS Accession No. ML16285A404).
Browns Ferry Nuclear Plant, Unit 2: Application dated June 19, 2014 (ADAMS Accession No. ML14175A307); NRC Safety Evaluation dated June 2, 2015 (ADAMS Accession No. ML15065A049).
4.3 No Significant Hazards Consideration Determination Analysis Duke Energy Progress, LLC (Duke Energy) requests an amendment to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Renewed Facility Operating License. The proposed change revises HNP Technical Specification (TS) 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System to reflect an update to the pressure and temperature limit curves in Figures 3.4-2 (Reactor Coolant System Cooldown Limitations) and 3.4-3 (Reactor Coolant System Heatup Limitations). The proposed change also reflects that the revised HNP pressure and temperature limit curves in TS Figures 3.4-2 and 3.4-3, as well as the existing power-operated relief valve (PORV) setpoints in Figure 3.4-4, will be applicable until 55 effective full power years (EFPY).
Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:
U.S. Nuclear Regulatory Commission RA-19-0359 Page 15
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises HNP TS to reflect updated pressure and temperature (P/T) limit curves in Figures 3.4-2 and 3.4-3 that are applicable until 55 EFPY. The proposed change also reflects that the existing PORV setpoints are applicable until 55 EFPY. The proposed change does not involve physical changes to the plant or alter the reactor coolant system (RCS) pressure boundary (i.e., there are no changes in operating pressure, materials or seismic loading). The proposed TS Figures 3.4-2 and 3.4-3, with an applicability term of 55 EFPY, provide continued assurance that the fracture toughness of the reactor pressure vessel (RPV) is consistent with analysis assumptions and NRC regulations. The methodology used to develop the proposed P/T limit curves provides assurance that the probability of a rapidly propagating failure will be minimized.
The proposed P/T limit curves, with the applicability term of 55 EFPY, will continue to prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises HNP TS to reflect updated P/T limit curves in Figures 3.4-2 and 3.4-3 that are applicable until 55 EFPY. The proposed change also reflects that the existing PORV setpoints are applicable until 55 EFPY. The proposed change does not affect the design or assumed accident performance of any structure, system or component or introduce any new modes of system operation or failure modes.
Compliance with the proposed P/T limit curves will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change revises HNP TS to reflect updated P/T limit curves in Figures 3.4-2 and 3.4-3 that are applicable until 55 EFPY. The proposed change also reflects that the existing PORV setpoints are applicable until 55 EFPY. HNP complies with applicable regulations (i.e., 10 CFR 50, Appendices G and H) and adheres to Nuclear Regulatory Commission (NRC)-approved methodologies (i.e., Regulatory Guides 1.99 and 1.190) with respect to the proposed P/T limit curves in TS Figures 3.4-2 and 3.4-3 in
U.S. Nuclear Regulatory Commission RA-19-0359 Page 16 order to provide an adequate margin of safety to the conditions at which brittle fracture may occur. The proposed P/T limit curves for HNP, with an applicability term of 55 EFPY, will continue to provide assurance that the P/T limits are not exceeded.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.
4.4 Conclusions In conclusion, based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commissions regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
- 5. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
- 6. REFERENCES
- 1. NRC letter, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Pressure/Temperature Limits (TAC No. MA8642), July 28, 2000 (ADAMS Accession No. ML003736272).
- 2. Duke Energy letter, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z, October 23, 2019 (ADAMS Accession No. ML19296C841).
- 3. Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, October 14, 2014.
U.S. Nuclear Regulatory Commission to RA-19-0359 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System Attachment 1 Technical Specifications Markup
Replace with Insert #1 1----------i 2400 MATERIAL P RCYERT YB ASES I I I I/
- 1SLH I I ControllingMderid -P ldeB 4197-2 I 100Pf/4F CoA'.)ef Content -0.09%
Nickel Content-0.50%
2200 Reg_jdoryG.Jioo-1 .99 Re.1. 2 I RT'°' lrltld-91'f RT K>< c:t 1/4 T -191'f *I RT Kl< d 3/4 T -179'f P res s U'e-T errp,rcture lirrits tote NOT 2000 l:Een ctjus ted for irr; trtPent errors .
T ha!i e errors cre controlled Of the T ectriod S P3dficction Eq..ipnEJ'lt List I I 1800 P rogcm P lent P rocecl.Jre P LP -106. I CJ 1600 in I Cl.
I IIII I
~ 1400 50°F/HR rn J rn I w
I 1200 I
C w
~
g 1000 I
C I 3:
I 800 J J
- 30°F/HR
~
.,,/
600 .....
400 200 0
50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE - °F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMIT ATIO NS - APPLICABLE TO UP TO 36 EFPY SHEARON HARRIS - UNIT l 3/4 4-35 Amendment No. l-00
2400 MATERIAL PROPERTY BASES:
ISLH Controlling Material Plate B4197-2 Copper Content 0.09%
Nickel Content 0.50%
2200 Regulatory Guide 1.99 Rev. 2 RTNDT Initial 91°F RTNDT at 1/4 T 212°F RTNDT at 3/4 T 198°F Pressure-Temperature limits have NOT 2000 been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.
1800 1600 INDICATED PRESSURE - PSIG 1400 50oF/HR 1200 1000 800 30oF/HR 600 400 200 0
50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE - °F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONSAPPLICABLE UP TO 55 EFPY
Replace with Insert #2 I
2400 MATERIAL PROPERTY BASES I I
I I Controlling Material - Plate B4197 -2 Copper Content - 0.09%
I I
ISLH I J 2200 Nickel Content - 0.50%
Regulatory Guide - 1.99 Rev . 2 I I I RT'°' Initial - 91 °F RT..,, at 1/4 T -191°F I I RTNOT at 3/4 T - 179°F I 2000 Pressure-Temperature limtts have NOT I
i been adjusted for instrument errors. These errors are controlled by the Technical I 1800 Specification Equipment List Program. Plant Procedure PLP-106. I I D I J 0 1600 Above 350"F, pressure limtt exceeds 2485 psig setpoinl of pressurizer safely relief valves !Soecificalion 3.4.2.1 ).
I iii D. j I
~ 1400 I
U) )
U) w f 1200 I
C I
...w 50°F/HR I
c3 1000 ,I I cz /
BOO /
/
/
/
v 600 400 200 0
50 100 150 200 250 300 350 INDICATED TEMPERATURE - °F FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS-APPLICABLE UP TO 36 EFPY SHEARON HARRIS - UNIT 1 3/4 4-36 Amendment No. 1oo
MATERIAL PROPERTY BASES:
2400 Controlling Material Plate B4197-2 Copper Content 0.09%
Nickel Content 0.50%
Regulatory Guide 1.99 Rev. 2 2200 RTNDT Initial 91°F RTNDT at 1/4 T 212°F RTNDT at 3/4 T 198°F Pressure-Temperature limits have NOT ISLH been adjusted for instrument errors.
2000 These errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.
1800 Above 350°F, pressure limit exceeds 2485 psig setpoint of pressurizer safety relief valves (Specification 3.4.2.1).
1600 INDICATED PRESSURE - PSIG 1400 1200 50°F/HR 1000 800 600 400 200 0
50 100 150 200 250 300 350 400 INDICATED TEMPERATURE - °F FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONSAPPLICABLE UP TO 55 EFPY
500 ~ - -- --- - -- ------------ ----------- -- - - - - --------- --- -1
~LOW PORv I . - - -- -- ~-- ------*--- ----------- - ------ .
~ - - - --- ------ -
--- - - ---- -- - -------- - ' - ___ _ _ __ --;i
- : -----1
~ - - --;.F--:.----- ---j S'
in
~
t--~+--~-+--~~ ----- ~- -;-_-_-~ _____ _____,_~~~i--==~
~
0 400 +----+----Oi-+---------------C'F----~- ---r-----1 t
(I) t---- - +--- ---+---- -+-- - --'-- - - -I- , ---- - t - - -- - - -i cc I
0 A.
t------'---- --i--- ----'--- -- -' - --
,__ _ _ _ L __ _ _~ - -- --- -- ----- *-- - ------ -
- - - - ~ - - - ' - - - - - --- -- *-- ---*- - -- - - - - - - - -----,
300 L -- --'-- - - - + - - - - -- -- t----- - - - -+----
o 100 200 300 400 MEASURED RCS TEMPERATURE ( F) 0 RCS TEMP (°F) LOW PORV* (psig) HIGH PORV* (psig) 90 400 410 250 400 410 325 440 450
- - - - - - - - - - -Change from 36 to 55
- VALUES BASED ON 36 55 EFPY REACTOR VESSEL DATA INSTRUMENT ERRORS ARE CONTROLLED BY THE TECHNICAL SPECIFICATION EQUIPMENT LIST PROGRAM, PLANT PROCEDURE PLP-106.
FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendment No.-lOO
U.S. Nuclear Regulatory Commission to RA-19-0359 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System Attachment 2 Technical Specifications Bases Markup (Provided for Information Only)
Replace with:
considered to be 0°F for conservatism based on the maximum closure head design specification requirement.
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g.,
pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. These limit lines shall be calculated periodically using methods provided below,
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70°F,
- 4. The pressurizer heatup and cooldown rates shall not exceed 100°F/h and 200°F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 625°F, and
- 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
The fracture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code. The fracture toughness testing of the ferritic materials associated with the replacement reactor vessel head (RRVH) was performed in accordance with the 2001 Edition, with Addenda up to and including 2003, of Section III of the ASME Boiler and Pressure Vessel Code. These properties are then evaluated in accordance with the NRC Standard Review Plan.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 36 effective full power years (EFPY) of service life.
The reactor vessel materials have been tested to determine their initial RT NDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the value of RTNDT, including margin, computed by Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."
Revision 1 SHEARON HARRIS - UNIT 1 B 3/4 4-7 Amendment No. 100
TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS CHARPY INITIAL UPPER SHELF ENERGY Cu Ni TNDT RTNDT TRANSVERSE COMPONENT GRADE HEAT NO (wt.%) (wt.%) (°F) (°F) FT LB Closure Hd. A508,Gr3,CL1 16W84 0.04 0.83 -40 -40 210 Vessel Flange A508, CL2 5302-V1 - - -10 -8 110 Inlet Nozzle " 438B-4 - - -20 -20 169
" " " 438B-5 - - 0 0 128
" " " 438B-6 - - -20 -20 149 Outlet Nozzle " 439B-4 - - -10 -10 151
" " " 439B-5 - - -10 -10 152
" " " 439B-6 - - -10 -10 150 Nozzle Shell A533B,CL1 C0224-1 .12 - -20 -1 90
" " " C0123-1 .12 - 0 42 84 Inter. Shell* " A9153-1 .09 .46 -10 60 83
" "* " B4197-2 .09 .50 -10 91 71 Lower Shell* " C9924-1 .08 .47 -10 54 98
" "* " C9924-2 .08 .47 -20 57 88 Bottom Hd. Torus " A9249-2 - - -40 14 94
" " Dome A9213-2 - - -40 -8 125 Weld (Inter & Lower Shell 4P4784 .05 .91 -20 -20 > 94 Vertical Weld Seams)*
Weld (Inter. to Lower Shell 5P6771 .03 .94 -80 -20 80 Girth Seam)*
The initial RTNDT for the Torus Plate material is a generic value based on the class of material calculated in accordance with Regulatory Guide 1.99 Rev. 2 and is considered proprietary.
Revision 1 SHEARON HARRIS - UNIT 1 B 3/4 4-8 Amendment No. 68
- I I
- ' l N
E 0
C 0
cu
- E 1.0
/\
w Q) 0
- I C i --
Cl> i u:
r I
- --i
! *. I ,
' i I I
- i I '
- I
- '. - - !. . -i--- - -~---* '** ..;- __ ,_ --
1 ~ : ' 1 1
'\ '
!' I 8 13 18 23 28 33 38 Operating Time (EFPY)
FIGURE B 3/4.4- 1 FAST NEUTRON FLUENCE (E> lMeV ) AS A FUNCTION OF FULL POWER SERVICE LIFE SHEARON HARRIS - UNIT 1 B 3/4 4-9
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)
The cooldown and heatup limits of Figures 3.4-2 and 3.4-3 are based upon an adjusted RTNor (initial RTNor plus predicted adjustments for this shift in RTNor plus margin).
In accordance with Regulatory Guide 1.99. Revision 2. the results from the material surveillance program. evaluated according to ASTM El85. may be used to determine ARTNor when two or more sets of credible surveillance data are available. Capsules will be removed and evaluated in accordance with the requirements of ASTM El85-82 and 10 CFR Part 50. Appendix H. The results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule . The cooldown and heatup curves must be recalculated when the ARTNor determined from the surveillance capsule exceeds the calculated ARTNor for the equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various cooldown and heatup rates are calculated using methods derived from Appendix Gin Section XI of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 ~nd ASME Code Case N-640 jfor the reactor vessel controlling ma er1a .
!D
.....-e-le-te---.~ ral method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology . In the calculation procedures for the beltline shell region a Insert:
semielliptical surface defect with a depth of one-quarter of the wall thickness. T. and a length of 3/2T is assumed to exist at the inside of the surface vesse wall as well as at the outside of the vessel wall. A ~emiellipticalk,---l, elete I Replace inside corner J l aw is assumed for the nozzle regions with a depth of one-with : quarter o E=1:JOZzle bel g wall thickness . The inlet nozzle is used in the upper shell calculation procures since the inner radius of this tapered nozzle is larger
~~~~ at the corne inner radius of the more tapered outlet nozzle. The
~r:'!:r:!th 1mensions o t ese postulated cracks. referred to in Appendix G of ASME Section XI as reference flaws . amply exceed the current capabilities of I
inservice inspection techniques. Therefore . the reactor operation limit Insert curves developed for reference crack are conservative and provide sufficient sa e y margins or protection against nonductile failure. To assure that the a radiation embrittlement effects are accounted for in the calculation of the limit curves. the most limiting value of the nil-ductility reference tempera-ture. RTNor, is used and this includes the radiation-induced shift. ARTNor*
corresponding to the end of the period for which cooldown and heatup curves are generated .
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor. K,. for the combined tnermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor. K,R* for the SHEARON HARRIS - UNIT 1 B 3/4 4- 11 Amendment No. 100
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued) metal temperature at that time. KIR is obtained from reference fracture toughness curves defined in the ASME Code. Pressure-temperature limits are developed for the vessel using the K IR curve defined in Appendix A to the ASME Code, as permitted by ASME Code Case N-640. For the Replacement Reactor Vessel Head (RRVH), pressure-temperature limits are developed, using KIR = KIc curve from ASME Section XI, Appendix G, 2007 Edition through 2008 Addenda. For the remaining components of the primary pressure boundary, pressure-temperature limits are based on the KIR curve defined in Appendix G to the ASME Code. The K IR curves are given by the equations:
Vessel regions: KIR = KIC = 33.2 + 20.734 exp[0.02(T-RTNDT)]
KIR = KIc = 33.2 + 2.806 exp [0.02(T-RTNDT + 100°F)] (1a)
Remaining regions:
KIR = KIa = 26.8 + 1.233 exp [0.0145(T-RTNDT + 160°F)] (1b)
Where: KIR is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RT NDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C KIM + KIt KIR (2)
Where: KIM = the stress intensity factor caused by membrane (pressure) stress, KIt = the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RTNDT of the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice leak and hydrostatic (ISLH) test operations.
At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the wall are calculated and then the corresponding thermal stress intensity factor, K IT, for the reference flaw is computed. The pressure stress intensity factors are obtained and allowable pressures are calculated from equation 2.
COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall and the inlet nozzle corner. During cooldown, the controlling location of the flaw is always at the inside surface because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The composite limit curves are developed considering the controlling reactor vessel component, either the beltline shell, RV head flange limit, or the inlet nozzle.
Revision 1 SHEARON HARRIS - UNIT 1 B 3/4 4-12 Amendment No. 100
REACTOR COOLANT SYSTEM BASES PRESSURE /TEMPERATURE LIMI TS (Continued)
The use of the composite cur ve in t he cool down analysis is necessary because control of the cooldown procedure i s ba sed on measurement of reactor coolant temperature. whereas the limiting pressure i s actually dependent on the material temperature at the ti p of t he ass umed flaw During cooldown. the
!Delete I 1/41,,Ji rls ide su rface! location~ s at a higher tern erature than the flui adJE ent'. fo me inside surfa ce.
Insert:
i 10n. 0 course. is no true for fromtheinside the steady-state situation. It follows that. at any given reactor coolant ....su_rt_ac_e ~ ~......
temperature . the ti.T developed during cooldown results in a higher value of KIA at the l/4T locati on for finite cooldown rates than for steady-state operation . Furthermore. if conditions exist such that the increase in KIA exceeds K~ . the calculated allowable pressure during cooldown will be greater than the steady -state value .
The above procedures are needed because there is no direct control on temperature at the l/4T location: therefore. allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period .
HEATUP Three separate cal culati ons are required to determine the limit curves for finite heatup rates . As is done in the cooldown analysis. allowable pressure-temperature relationships are devel oped for steady-state conditi ons as well as finite heatup rate conditions assuming the presence of a l /4T defect at the inside surface. The thermal gradients during heatup produce compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature: therefore . the KIA for the l/4T crack during heatup is lower than the K1R for the l/4T crack during steady-state conditions at the same coolant temperature . During heatup. esQecially at the end of the tran -
sient . conditions may exist such that the effects of compressive thermal stresses and different K1R*s for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore. both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained .
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a l/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface. the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present . These thermal stresses. of course. are aependent on both the rate of SHEARON HARRIS - UNIT 1 B 3/ 4 4- 13 Amendment No . l 00