ML24179A028
| ML24179A028 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/26/2024 |
| From: | Sivaraman M Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| Download: ML24179A028 (1) | |
Text
Post Office Box 2000, Decatur, Alabama 35609-2000
June 26, 2024 10 CFR 50.55a 10 CFR 50.4(a)
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Browns Ferry Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-33 NRC Docket No. 50-259
Subject:
Evaluation of Effects of Out-of-Limits Condition as described in IWB-3720(a)
The Tennessee Valley Authority is submitting the Browns Ferry Nuclear Plant (BFN) evaluation of Pressure-Temperature excursions following a reactor scram on April 24, 2024. The evaluation is contained in the enclosure to this letter and is in accordance with the requirements of 10 CFR 50.55a(a)(2)(xliii).
There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact David Renn, Site Licensing Manager, at (256) 729-2636.
RespectfullyRespectfully,,
ManuManu Sivaraman Sivaraman BFN Site Vice President
Enclosure:
Engineering Assessment of BFN1 P-T Excursions Following April 2024 Scram
cc (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant
ENCLOSURE
Browns Ferry Nuclear Plant Unit 1
Engineering Assessment of BFN1 P-T Excursions Following April 2024 Scram
See Enclosed
FILE NO.: 2400586.301
PROJECT NO.: 2400586
Quality Program Type: Nuclear Commercial
CALCULATION PACKAGE
PROJECT NAME:
CONTRACT NO.:
PO 7622947
CLIENT: PLANT:
Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant (BFN)
CALCULATION TITLE:
Engineering Assessment of BFN1 P-T Excursions Following April 2024 Scram
Document Affected Project Manager Preparer(s) &
Revision Pages Revision Description Approval Checker(s)
Signature & Date Signatures & Date 0 1 - 25 Initial Issue Prepared By:
A-1 - A-6 B-1 Daniel B. Patten See Page-2
DBP 5/5/2024 Reviewed By:
See Page-2
© 2024 Structural Integrity Associates, Inc. All rights reserved. No part of this document or the related files may be reproduced or transmitted in any form, without the prior written permission of Structural Integrity Associates, Inc.
Scope A: P-T Surveillance Data Review
Task
Description:
Prepared By: Reviewed By:
Finite Element Modeling, Pressure and Thermal Stress Predictions Daniel B. Patten Mark J. Jaeger DBP 5/05/2024 MJJ 5/05/2024
Scope B: Assessment of Bottom Head Curve-1 Excursion
Task
Description:
Prepared By: Reviewed By:
Finite Element Modeling and Thermal Stress Predictions Tyler D. Novotny Tim D. Gilman TDN 5/05/2024 TDG 5/05/2024 Heat Transfer Coefficients for Modeling
Keith R. Evon Tim D. Gilman KRE 5/05/2024 TDG 5/05/2024
Fracture Mechanics Assessment, Compliance with ASME Code Section-XI Dilip Dedhia, PhD DD 5/05/2024
Nat Cofie NC 5/05/2024 Sam Ranganath SR 5/05/2024
Scope C: Evaluation of Excess Ramp Rates
Task
Description:
Prepared By: Reviewed By:
Evaluation of Alternate Heatup/Cooldown Rates for P-T Curve Applicability Daniel B. Denis Jianxin Wang DBD 5/05/2024 JW 5/05/2024
2400586.301.R0 Page 2 of 24
F0306-01R4
TABLE OF CONTENTS 1.0 Introduction........................................................................................................................ 4 2.0 Technical Approach........................................................................................................... 5 Scope B: Bottom Head (BH) Curve-1 Excursion....................................................................... 5 FEA Stress Analysis.............................................................................................................. 5 Fracture Mechanics Evaluation............................................................................................. 8 Scope-C: Excess Ramp Rates Methodology.......................................................................... 16 Scope C Background........................................................................................................... 16 Scope C Approach............................................................................................................... 17 3.0 Results of Analysis........................................................................................................... 19 Scope-A: Surveillance Data Review........................................................................................ 19 Scope-B: BH Curve -1 Excursion............................................................................................. 19 Scope-C: Excess Ramp Rates................................................................................................ 19 4.0 Observations/Conclusions............................................................................................... 23 4.1 Scope A Conclusion..................................................................................................... 23 4.2 Scope B Conclusion..................................................................................................... 23 4.3 Scope C Conclusion..................................................................................................... 23 5.0 References....................................................................................................................... 23 APPENDIX A : Review of Surveillance Data............................................................................. A-1 APPENDIX B : Supporting Files................................................................................................ B-1
LIST OF TABLES Table 1: Bottom Head Tabular Values for 200°F/hr Ramp Rate Curve (K It of 17.7 ksiinch)... 20
LIST OF FIGURES Figure 1: As Modeled Dimensions with Conservative Values (6.125in or 8.00in as Applicable). 6 Figure 2: Field Measured vs. As -Modeled Transient-01 OD Temperatures................................. 7 Figure 3: ANSYS, ID Applied (Wetted Surface) Transient Temperatures.................................... 8 Figure 4: ANSYS Different Paths Used for the Fracture Mechanics Evaluation........................ 11 Figure 5: ANSYS Comparison of (2 KIP + KIT) vs. KIc for Path 1 (Pr Str pR/2t)........................... 12 Figure 6: ANSYS Comparison of (2 KIP + KIT) vs. KIc for Path 1 (Pressure Stress = pR/t)......... 13 Figure 7: ANSYS Comparison of (2 KIP + KIT) vs. KIc for Path 2................................................. 14 Figure 8: ANSYS Comparison of (2 KIP + KIT) vs. KIc for Path 3................................................. 15 Figure 9: Ramp Rate (left) and P -T Curve Compliance (right) for RPV Bottom Head Region.... 19 Figure 10: Pressure-Temperature Limit Curve of Bottom Head with 200°F/hr Ramp Rate with 25 EFPY Curve 1 and Plotted Temperature D ata....................................................................... 21 Figure 11: Extended Beltline Zone for 50 EFPY for BFN1.......................................................... 22
2400586.301.R0 Page 3 of 24
F0306- 01R4
1.0 INTRODUCTION
On Wednesday, April 24, 2024, at approximately 22:15, Browns Ferry Unit 1 (BFN1) experienced a failed bushing in the main transformer, which imparted a turbine trip and subsequently initiated a reactor scram. Following the scram, BFN1 experienced a loss of feedwater pumps (LOFP) and recirculation pumps, which in turn caused the high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and control rod drive (CRD )
systems to activate and inject water into the reactor pressure vessel (RPV) to maintain level.
HPCI and RCIC only injected for a few minutes, but CRD continued to inject approximately 80 gpm directly to the bottom head of the vessel for several hours.This in turn resulted in temperatures of the RPV bottom head decreasing significantly with the remainder of the RPV still at nominal pressure. Later in the shutdown, temperatures in the bottom head increased rapidly, exceeding specified heatup rate limits.
In the process of completing its standard RPV pressure and temperature (P -T) surveillance monitoring, TVA operators noted potential excursions beyond the established limits.
Specifically, RPV bottom head temperatures may have exceeded the Curve-1 P -T limits, and cooldown/heatup rates may have exceeded the 100 °F/hr limit in multiple regions.
TVA requested assistance from Structural Integrity Associates, Inc. (SI) to evaluate the surveillance data, and assess whether the potential excursions remained bounded by ASME Code limits throughout the event. The purpose of this calculation is to summarize SIs review of the associated data and document separate evaluation(s) that were performed to verify Code compliance. These scopes are broken out as follows:
A. P-T Surveillance Data Review :
Evaluation of surveillance progression as documented by TVA. Results are characterized independently in Appendix A - as illustrated in Figure A-4 and Figure A-2 therein, the RPV bottom head region experienced an excursion beyond the specified Curve-1 limits and subsequently heated up at a ramp rate in excess of 100 °F/hr. Upon further review, no other RPV regions experienced P-T curve or ramp rate excursions.
This review formed the basis for the two additional scopes that follow.
B. Assessment of Bottom Head Curve-1 Excursion:
Finite element analysis (FEA) and fracture mechanics (FM) to demonstrate ASME Code compliance for the high-pressure, low-temperature state observed several hours after the initial scram event.
C. Evaluation of Excess Ramp Rates:
Alternate calculation of P-T limit curves using modified heatup/cooldown rates to demonstrate sufficient margin, for the reheat period in excess of 100°F/hr.
2400586.301.R0 Page 4 of 24
F0306-01R4
2.0 TECHNICAL APPROACH
Scope B: Bottom Head (BH) Curve-1 Excursion
FEA Stress Analysis
ASME Code of Record, 1965 Edition based on the BFN stress report Reference [5, PDF 6, 1252].
The lower head hemisphere is taken to be SA-302 Grade B (Mn-1/2Mo) or SA-508 Cl. 2 (C -
1/2Mo-Cr -Ni) based on References [5][6][Spec 21A1111 in 7]. Both component materials have identical material properties for the thermal analysis (elastic modulus, mean coefficient of thermal expansion, diffusivity, specific heat, density, and poisons ratio), based on the 1965 edition of the ASME code. The skirt is assumed to be SA106 Gr. C, which is acceptable as the skirt is not the subject of this evaluation and is only modeled for completeness. Thermal diffusivity and specific heat are obtained from more recent ASME code, 2017 ASME Section II, Part D, for completeness.
The inside diameter of the bottom head was taken from Reference [8 ] as 251 inches to the base metal. Cladding, which is not modeled, is 3/16. The bottom head thicknesses is interpreted from Reference [5, PDF 1267, 1269]. The skirt radius and thickness are interpreted from Reference [5] to be as shown in Figure 1.
2400586.301.R0 Page 5 of 24
F0306- 01R4
6.125
125.5
1.25 8.00
113.75
Figure 1: As Modeled Dimensions with Conservative Values (6.125in or 8.00in as Applicable)
The inside surface heat transfer coefficient, film coefficient, was taken as 500 BTU/hr-ft 2-°F based on the original design analysis for BFN Unit 1 vessel shell [5, PDF 1264].The outside surface was assumed to be insulated, with a film coefficient of 0.2 BTU/hr -ft² -°F and an ambient containment temperature of 100°F.
Transient-01 is modeled to mimic the as -measured temperature provided by BFN for the bottom head, specifically temperature '56-30'. The OD temperatures were interactively checked and then compared against the BFN provided data. Figure 2 illustrates the OD temperature comparison of the field measured and as -modeled transients. Figure 3 illustrates the ANSYS
[14] applied ID temperature versus time.
Transient-02 is modeled as specified by the BFN Unit 1 thermal cycle diagram, GE Drawing 729E762 [9 ], for region C for the Loss of FW Pumps event, in which HPCI is defined to inject a total of 3 consecutive times. Figure 3 illustrates the ANSYS applied ID temperature versus time.
2400586.301.R0 Page 6 of 24
F0306-01R4
Figure 2: Field Measured vs. As-Modeled Transient-01 OD Temperatures
2400586.301.R0 Page 7 of 24
F0306- 01R4
Figure 3: ANSYS, ID Applied (Wetted Surface) Transient Temperatures
Fracture Mechanics Evaluation
Pressure-temperature (P -T) limit curves for pressure test (Curve A), and heatup/cooldown conditions for core not critical (Curve B) and core critical operation (Curve C) for up to 25 EFPY of operation are contained in the Browns Ferry Unit 1 plant Technical Specifications [3]. These curves were developed using the GE PTLR method [4]. The P -T curves (Curve A for the pressure test and Curves B and C for heatup and cooldown with core not critical and core critical conditions). The P-T curves incorporate the requirements of ASME Code Section XI Appendix G and 10CFR50 Appendix G. The key criterion is that the Applied K for the postulated one quarter-thickness (1/4t) flaw (including a structural factor of 2 on pressure and 1 on thermal stress) is less than the available KIC value. The postulated semi-elliptic flaw has a depth equal to 0.25 times the thickness and length equal to1.5 times the thickness (i.e., aspect ratio a/L=1/6).
Following the pressure - temperature conditions described above, an evaluation of the allowable P-T curve conditions was performed to demonstrate that the requirement of a structural factor of 2 on pressure and 1 on thermal stress are maintained during the transient following the scram for the bottom head region using the actual transient measured temperature and pressure data.
The Code required margin for the bottom head c urve P-T limits was confirmed using the actual stresses and temperatures determined from the ANSYS model, the TIFFANY code [ 15] and Excel spreadsheets that have been independently benchmarked against the ASME Section XI, Appendix G method as described below. The bottom thermocouple measurements were used to benchmark the calculated through-wall temperatures of the bottom head metal. The methods and inputs used to demonstrate that the Appendix G margins were maintained (i.e., structural
2400586.301.R0 Page 8 of 24
F0306- 01R4
factor of 2 on KIP (pressure) and 1 on KIT (thermal) using the fracture mechanics analyses methods as summarized below.
The steps in the analyses were as follows:
Step 1: Determine the fluid temperature, T, vs. time based on the actual thermal transient condition as described above. The through-wall temperatures in the bottom head were calculated at each time in the transient to determine the through-wall stresses and temperatures.
Step 2: For an assumed 1/4t reference flaw at the limiting location(s), the thermal stresses at the crack tip were extracted from the ANSYS results and the applied thermal stress intensity, KIT, was determined using the formula in Equation (1). A survey of the stress conditions in the vessel was performed at three locations (or paths) in the bottom head as shown in Figure 4 to be sure that the limiting (or bounding) condition in the bottom head has been evaluated. These paths were then used in the subsequent fracture mechanics analyses.
(1)
Step 3: For each of these paths and at each point in time during the transient, the static initiation fracture toughness, KIC, was computed at the 1/4t depth crack tip temperatures using the following Equation (2) from [13]
(2) K Ic =33 ART.2+20.734exp[0.02(T)]
Where: KIC = the lower bound static initiation critical fracture toughness (ksiin).
T = the metal temperature at the tip of the postulated 1/4 through-wall flaw (°F).
ART = RTndt for the lower head base material since the bottom head is not affected by fluence.
Step 4: For the bottom head region, the compliance with E quation (3) was demonstrated for postulated flaws at different bottom head locations.
2 KIP + KIT < KIc (3)
The value for KIP was calculated using the time dependent stresses and geometry -
dependent functions for the assumed reference flaws at each stress path with aspect ratio of a/L=1/6 and oriented normal to the largest primary membrane stress in the vessel head. Evaluation was performed for the three paths shown in Figure 4.
For Path 3 an additional stress concentration factor (SCF) of 2.0 was included in the membrane stress equation to account for stress concentrations at the holes for the CRD nozzles. The value of 2.0 for equal biaxial tensile loading around a hole is well established for the stress state around a hole based on superposition of the Kirschs solution.
2400586.301.R0 Page 9 of 24
F0306-01R4
The allowable stress intensity (or margin) was calculated for each point in time during the transient condition through each stress path. The residual margin can be expressed as the ratio of KIC/(2KIP + KIT). In all cases where this ratio exceeds 1.0 throughout the entire transient, this quantity assures that there is no exceedance of the required Appendix G margin. The results are shown graphically in Figure 5, Figure 6, Figure 7, and Figure 8, for the three paths evaluated.
2.1.1.1 Design Inputs 1.Bottom Head RTNDT: 56°F [10]
- 2. RPV Dimensions:
RPV inside radius: 125.0 inches [10]
RPV shell thickness: 6.125 inches [10]
Bottom head inside radius: 125.0 inches [10]
Bottom head shell thickness: 8.00 inches [10]
- 3. Maximum Heat-up / Cool-down Rate: +/-100°F/hr
2.1.1.2 Discussion,Results, and Conclusions - Scope B The through-wall thermal stresses were calculated by FEA at approximately 600 time steps. The calculated crack tip stress intensity factors were calculated at each of those time steps for a 1/4 t flaw with length/depth ratio of 6. The FEA calculated stresses consisted of the meridional and hoop components. The larger of these stresses were used in the calculation of K IT. KIP due to pressure at the corresponding step was added. As discussed above, a structural factor of 2.0 was applied to the pressure stress and 1.0 to the thermal stresses. In addition, a stress concentration factor (SCF) of 2.0 was also applied to the pressure stresses for Path 3 to account for the penetrations at the bottom head.
The 1/4t depth axial crack model was used for Paths 1 and 2, and the 1/4t depth circumferential crack model was used for Path 3.The paths are shown in Figure 4.
2400586.301.R0 Page 10 of 24
F0306- 01R4
Figure 4: ANSYS Different Paths Used for the Fracture Mechanics Evaluation
The comparison of the required (2K IP + KIT) values to meet the Appendix G criterion and the available toughness are shown for the three paths in Figure 5 through Figure 8.
For Path 1, which is at the juncture of the cylindrical shell and the spherical shell, the pressure hoop stress is pR/t in the cylindrical section and pR/2t in the spherical section (where p is internal pressure and t is the thickness). The finite element model did not include the cylindrical shell. Pending the FEM results with model that includes the cylindrical section, two plots are made for Path 1 using the pressure stress in the cylinder and spherical shell, respectively.
(Figure 5 and Figure 6). It is expected that the actual pressure stress will be between the values of pR/t in the cylindrical section and pR/2t in the spherical section. For the assumed stress of pR/2t corresponding to the spherical shell, the results (shown in Figure 5) show more margin. As expected, the results in Figure 6 (for the assumed stress of pR/t) are conservative. However, in both cases, the available toughness exceeds the requir ed K values, and the Appendix G criterion is met.
For Path 2, the FEM results are equal to the pR/2t as it is in the spherical section. The comparison of the available toughness and the required K value is shown in Figure 7, Again, the available toughness exceeds the required K values, and the Appendix G criterion is met.
For Path 3, the pressure stress is multiplied by the SCF of 2 to account for the hole in the shell for the CRD penetration. For Path 3, the FEM results are equal to the pR/2t value as it is in the spherical section. The comparison of the available toughness and the required K value is shown
2400586.301.R0 Page 11 of 24
F0306- 01R4
in Figure 8, Again, the available toughness exceeds the required K values, and the Appendix G criterion is met.
Sensitivity Analysis
The value of 2.0 for SCF used for Path 3 is based on a single hole in a plate with equal biaxial loading. Since there are multiple penetrations, the actual Kt might be slightly higher. To address this, a sensitivity analysis was performed to determine how high the value of SCF could be while still maintaining the Appendix G requirements. The sens itivity analysis showed that the SCF could be as high as 2.5 and still meet the Appendix G criteria. This confirms that there is added margin to address the effect of multiple penetrations.
In summary, the ASME Appendix G requirement shown in Equation 3 is met for all three paths.
Path 1 - assuming pressure stress of pR/(2t).
Figure 5: ANSYS Comparison of (2KIP + KIT) vs. KIC for Path 1 (Pr Str pR/2t)
2400586.301.R0 Page 12 of 24
F0306- 01R4
Path 1 - assuming pressure stress of pR/t.
Figure 6: ANSYS Comparison of (2KIP + KIT) vs. KIC for Path 1 (Pressure Stress = pR/t)
2400586.301.R0 Page 13 of 24
F0306- 01R4
Figure 7: ANSYS Comparison of (2KIP + KIT) vs. KIC for Path 2
2400586.301.R0 Page 14 of 24
F0306- 01R4
Figure 8: ANSYS Comparison of (2KIP + KIT) vs. KIC for Path 3
2400586.301.R0 Page 15 of 24
F0306- 01R4
Scope-C: Excess Ramp Rates Methodology
Scope C Background
Browns Ferry Unit 1 (BFN1) operational pressure -temperature limits curves were developed for 25 effective full power years (EFPY) by General Electric Hitachi (GEH). These curves provide guidance to bound the operation of the power plant, specifically for the protection of the reactor pressure vessel (RPV) from brittle fracture. These curves are defined by the ASME Boiler &
Pressure Vessel Code (BPVC) in Section XI Appendix G and further governed by regulations in 10 CFR 50 Appendix G. As such, these curv es are based on linear elastic fracture mechanics (LEFM) to ensure that plants are operated in a manner that protects the vessel from rapid propagation of a crack and failure. The flaw evaluation methodology is based on an assumed flaw of one-quarter of the thickness (1/4T) of the reactor pressure vessel.
In this approach, a stress intensity factor (K I) must remain below the static fracture toughness (KIC), a material property of base metal and weld materials. The available fracture toughness is defined by the initial RTNDT, a test-based property defined by NB -2300 of ASME BPVC Section III (for Design of Nuclear Power Plants). The fracture toughness can be reduced over the lifetime of a reactors operation by neutron irradiation, which defines a new value, the Adjusted Reference Temperature (ART) for the vessel material, and this defines a value for KIC that decreases with increasing neutron irradiation. (For the purposes of the current discussion of the bottom head, the material of the bottom head does not receive sufficient neutron irradiation to affect the material property for K IC, therefore, the initial RTNDT is also the same as the end-of-life ART.)
Besides the RTNDT (or ART as appropriate), the KIC value of a material is affected by the current operating temperature of a material, such that increasing temperature increases the value of K IC for a given value of RTNDT (or ART). The applied KI is affected both by the pressure in the vessel (increasing pressure increases the pressure contribution to K I) and by the thermal ramp rate (higher rate increases the thermal contribution to K I).
In overall summary, the vessel must be sufficiently warmed before pressurization for startup and also, during a shutdown, the pressure must be reduced before the v essel is allowed to cool. The specific parameters of these heatup and cooldown cycles are defined by pressure-temperature limit curves, which provide guidance and boundaries to retain structural stability and not have occurrence of rapidly propagating failure due to unstable brittle crack extension.
The pressure-temperature limit curves (also called P -T curves) are defined by two applicability criteria. One is a time-based criterion for plant age (generally defined in EFPY) that is a proxy for the accumulated embrittlement and shift in the K IC value for various materials in the reactor vessel. The other criterion is the temperature ramp rate, which provides a maximum thermal stress effect on the vessel. If either the limitation of the plant age or the te mperature ramp rate of the curves is exceeded, an evaluation for the specific reactor conditions and transients must be made to establish if a risk to the structural stability of the reactor occurred. Typically, the temperature ramp rate for the components in the RPV are limited to 100°F per hour (although other ramp rates can be utilized if appropriate analysis is performed). For P -T Limit Curve
2400586.301.R0 Page 16 of 24
F0306- 01R4
generation, curves are generated for various components in the RPV, and the limiting curve provides the overall composite curve that defines operational limitations for the unit.
In spring 2024, Browns Ferry Unit 1 (BFN1) had a scram event that continued to shutdown and during a portion of the operational event, a ramp rate of approximately 185°F/hr was experienced by the bottom head of the RPV. Figure A -2 in Appendix A shows the ramp rate for various components that were monitored during the scram event.
Within the BFN1 Technical Specifications, there is a Curve 1 that defines limitations for the bottom head of the RPV at 100°F per hour. Although that ramp rate was exceeded for a portion of the event, there is margin within the material to tolerate a higher ramp rate for operation. The purpose of this evaluation is to define a higher ramp rate curve for the bottom head of the BFN1 RPV that envelopes the rates experienced during the scram event and subsequently provide confirmation that the component was not at-risk for brittle failure or exceedance of the composite Curve B for Core Not Critical operation.
Scope C Approach
Structural Integrity Associates has previously prepared pressure-temperature limit curves in support of BFN1 and its subsequent license renewal [10]. The calculation employed the methodology of an NRC-approved topical report for generation of boiling water reactor (BWR) pressure-temperature limits [11]. The current approach will be to take the data for BFN1 and utilize the approved methodology approach for the pressure -temperature limit curves [11] for the bottom head component, but instead of a ramp rate of 100°F/hr (provided in the current Technical Specifications curves), a bounding ramp rate of 200°F/hr will be utilized.
The newly developed curve will show acceptability of the ramp rate exceedance event.
Although the higher ramp rate will increase the thermal stress contribution to the applied stress intensity factor, the bottom head material retains sufficient margin in its K IC value not to be negatively affected by the increased ramp rate for brittle crack stability.
Furthermore the prior RTNDT (and equivalent ART) for the bottom head of 56°F shall be utilized
[10], and because of the low neutron irradiation levels below 1 x 10 17 n/cm2 (E > 1.0 MeV), the RTNDT value remains valid for utilization of pressure-temperature limit curves in a time-independent (not dependent on limiting EFPY of the plant) manner.
Figure 11 shows the extended beltline for the BFN1 RPV, with the green zone showing the axial area around the core mid-plane with fluence greater than 1 x 10 17 n/cm2 (E > 1.0 MeV) for 50 EFPY (assumed EFPY for 80 years of plant operation) [12]. The fluence evaluation provides confirmation that no embrittlement shift from neutron irradiation is required for evaluation in the bottom head, which is well below the extended beltline region.
Justification of 200°F/hr Ramp Rate Versus 100°F/hr Ramp Rate
The pressure-temperature limits curves are defined at a particular ramp rate. The ramp rate relates to a thermal contribution to the overall applied stress intensity factor. The pressure-temperature limits for BWRs are generally defined by the more limiting cooldown event that induces stress on the 1/4T location (versus the 3/4T location for the heatup event).
2400586.301.R0 Page 17 of 24
F0306- 01R4
The below equation defines the thermal contribution to the stress intensity factor from Appendix G of the ASME Code,Section XI [13]:
= 0.953 103 2.5 Where:
KIT = Thermal contribution to stress intensity factor, CR = Cooling rate in °F/hr, and t = thickness of the material in inches.
Thermal contribution to stress intensity factor, K IT, for 100°F/hr is 8.85 ksiinch while the value for 200°F/hr is 17.7 ksiinch. The newly defined curve replaces the existing 100°F/hr Curve 1 for operating events that contain heatup or cooldown rates up to 200°F/hr. The pressure contribution to stress intensity factor is dependent on the pressure of the water inventory within the vessel and is independent of the thermal ramp rate.
The 200°F/hr curve (shown in Figure 10) shifts down and to the right on the graph and captures additional design margin, while being more restrictive in operating space relative to the 100°F/hr Curve 1. The combined applied stress intensity factor (from pressure and thermal contributions) is compared to the margin within a component of the K IC (for a given operating temperature).
2400586.301.R0 Page 18 of 24
F0306- 01R4
3.0 RESULTS OF ANALYSIS
Scope-A: Surveillance Data Review
Appendix A contains a complete summary of SIs review of surveillance data provided by TVA for the 4/24 scram and ensuing cooldown. As documented therein, two issues were observed for the bottom head region: an excursion beyond the Curve-1 P -T limits, and a heatup where ramp rates exceeded 180 °F/hr.These concerns were evaluated in detail herein, with results of those supplementary analyses presented in the following sections.
Figure 9: Ramp Rate (left) and P-T Curve Compliance (right) for RPV Bottom Head Region
Scope-B: BH Curve-1 Excursion
As shown in Figure 5 through Figure 8, the ASME Appendix G requirement shown in Equation 3 are met for all three representative paths.
Scope-C: Excess Ramp Rates
Figure 10 shows the dotted line 200°F/hr pressure-temperature limit curve plotted with the existing 25 EFPY Curve 1 (for bottom head) currently within the Browns Ferry Unit 1 Technical Specifications. The data from the scram event, specifically from thermocouples 56-29, 56-30, and 56-31 (during the time of the ramp rate exceedance event) have also been plotted.
The new 200°F/hr line is positioned lower to the right than Curve 1, showing a more limiting curve than the 100°F/hr Curve 1. The new curve provides margin to the existing Curve 2 (Curve 2400586.301.R0 Page 19 of 24
F0306- 01R4
B) for Core Not Critical operation. Tabular data for the 200°F/hr curves are contained within Table 1. Furthermore, all data associated with the ramp rate event (thermocouple information from 56-29, 56-30, and 56-31) are positioning to the lower right of the 200°F/hr limit curve, providing assurance of protection from brittle fracture.
Table 1: Bottom Head Tabular Values for 200°F/hr Ramp Rate Curve (KIt of 17.7 ksiinch)
Gage Fluid KIP P-T Curve Temperature KIC (Pressure Contribution to Temperature P-T Curve Pressure Stress Intensity Factor)
°F ksi*in1/2 ksi*in1/2 °F psig 83.1 68.9 25.6 83.1 0.0 83.1 68.9 25.6 83.1 330.9 87.1 71.8 27.1 87.1 352.0 91.1 75.0 28.7 91.1 374.8 95.1 78.5 30.4 95.1 399.5 99.1 82.3 32.3 99.1 426.2 103.1 86.4 34.3 103.1 455.2 107.1 90.8 36.6 107.1 486.6 111.1 95.6 39.0 111.1 520.6 115.1 100.8 41.6 115.1 557.5 119.1 106.4 44.4 119.1 597.4 123.1 112.5 47.4 123.1 640.6 127.1 119.2 50.7 127.1 687.4 131.1 126.3 54.3 131.1 738.2 135.1 134.1 58.2 135.1 793.1 139.1 142.5 62.4 139.1 852.7 143.1 151.6 66.9 143.1 917.2 147.1 161.4 71.9 147.1 987.0 151.1 172.1 77.2 151.1 1062.7 155.1 183.7 83.0 155.1 1144.7 159.1 196.2 89.3 159.1 1233.5 163.1 209.8 96.0 163.1 1329.8 167.1 224.5 103.4 167.1 1434.0 171.1 240.4 111.4 171.1 1546.9 175.1 257.7 120.0 175.1 1669.2
2400586.301.R0 Page 20 of 24
F0306-01R4
Curve B - Core Not Critical, All Components
& Bottom Head
BH 200°F/hr Curve Curve 1 25 EFPY 56-29 Data 56-30 Data 56-31 Data 1300
1200
1100
1000
900
800
700
600
500
400
300
200
100
0 0 50 100 150 200 250 300 Minimum Reactor Vessel Metal Temperature (° F)
Figure 10: Pressure-Temperature Limit Curve of Bottom Head with 200°F/hr Ramp Rate with 25 EFPY Curve 1 and Plotted Temperature Data
2400586.301.R0 Page 21 of 24
F0306- 01R4 Figure 11: Extended Beltline Zone for 50 EFPY for BFN1
2400586.301.R0 Page 22 of 24
F0306- 01R4 4.0 OBSERVATIONS/CONCLUSIONS
4.1 Scope A Conclusion
Appendix A presents the results of the measured temperature and pressure data, and is used in the subsequent Scopes B and C. No specific conclusions are applicable.
4.2 Scope B Conclusion
The calculated crack tip stress intensity factors for pressure and thermal loading with appropriate structural factors (2KIP + KIT)were calculated for a 1/4 t flaw and were all shown to be less than the allowable KIC values.
4.3 Scope C Conclusion
Based on the evaluation, the bottom head components that experienced greater than 100°F/hr temperature ramp rates (approximately 180 °F/hr) have no risk of brittle fracture and satisfy the pressure-temperature limit curve defined for a 200°F/hr temperature ramp rate. The 200°F/hr curve captures additional margin available for the component to experience higher temperature ramp rates.
The applied ramp rate did not apply any degradation to the components or impair ment to their safe and reliable operation.
5.0 REFERENCES
- 1. BFN Procedure 1-SR -3.4.9.1(1), "Reactor Heatup and Cooldown Rate Monitoring," Revision 14, April 2015, SI File No. 2400586.201.
Subject:
Copy of the actual SR with data -
Reactor Heatup and Cooldown Rate Monitoring, Received 5/1/2024 at 5:22pm EDT, Including Attachment 1 -SR -3.4.9.1 with data.pdf, SI File N o. 2400586.2 01.
Subject:
RE: DatAWare Points for P-T Evaluation, Received 4/29/2024 at 6:51pm MDT, Including Attachment U1 Vessel Data for Structural Integrity Associates.xlsx, SI File No. 2400586.2 01.
Subject:
RE: SI Requested Information 5/1/24 @ 2200, request 4, Received 5/1/2024 at 9:40pm and 10:03pm MDT, Including Attachments 56-8 and 56-35.xlsx and 56 -23 and 56-26.xlsx, SI File No. 2400586. 201.
- 5. Summary Stress Report for General Electric - NED TVA #1 & #2 Reactor, General Electric Order No. 205-55577, B&W Contract No. 610-0127-51, M05421 Stress Report, September 1970, SI File No. 2001191.209.
- 6. SI Calculation Package 2100318.302, Past Operability Evaluation of the Shroud Support Access Hole Covers.
- 7. General Electric Purchase Specification 21A1111, Reactor Pressure Vessel, SI File No.
2100318.204
- 8. Babcock and Wilcox Drawing 122860E, Shell Segment Assembly Course #1 & #4, SI File No.
1700558-202.
- 9. General Electric Drawing 729E762, Reactor Thermal Cycles, SI File No. 1700558.212.
- 10. Structural Integrity Calculation, 2200107.306, Revision 1, BFN Units 1, 2, and 3 Pressure-Temperature (P-T) Limit Curves Calculation.
- 11. Licensing Topical Report (LTR) BWROG-TP -11 -022-A (SIR 044), Revision 1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, August 2013, ADAMS Accession No. ML13277A557.
2400586.301.R0 Page 23 of 24
F0306- 01R4
- 12. TransWare Enterprises Report, BFN-FLU -001-R -004, Revision 0, Browns Ferry Nuclear Plant Unit 1 Reactor Pressure Vessel Fluence Evaluation - Subsequent License Renewal, August 2022. SI File Number 2200107.201.
- 13. ASME Boiler & Pressure Vessel Code,Section XI, Division 1, Rules for Inspection and Testing of Components of Light-Water Cooled Plants, 2019 Edition.
- 14. ANSYS Mechanical APDL (UP20170403), Release 18.1, SAS IP, Inc.
- 15. SI-TIFFANY 3.2, Structural Integrity Associates, April 2020.
2400586.301.R0 Page 24 of 24
F0306- 01R4 APPENDIX A:
REVIEW OF SURVEILLANCE DATA
2400586.301.R0 Page A-1 of A-6
F0306- 01R4 BFNs process for monitoring of reactor heatup and cooldown evolutions is governed by procedure 1-SR-3.4.9.1(1) [1 ]. For the 4/24 scram, a pen-and-ink copy of the procedure as completed by BFN operations was provided in [2 ]. The procedure appears to denote several excursions during which specified ramp rates (heatup/cooldown) and/or P -T limit curves were exceeded. A variety of existing instruments are used to make these comparisons, although only a subset of the sensors are available in BFNs data historian.
Historian data was provided by TVA in an XLSX-compatible format [3, 4]. Refer to Figure A-1 for a markup schematic of the subset of instruments used in this evaluation. All highlighted tags are evaluated for ramp rate; tags highlighted blue are subsequently assessed for P-T curve compliance.
Tags highlighted green are optional points that are not included in TVAs surveillance procedure [1]
but are evaluated herein to ensure that the maximum ramp rate is appropriately characterized.
Figure A-1: Temperature Instruments Utilized for Analysis Figure A-2 plots all highlighted tags from Figure A-1 to illustrate temperature progression during the shutdown f or various regions of interest. In addition to the highlighted tags, the steam dome region and RHR heat exchanger inlet indications are plotted for consistency with Attachment 1 of [1 ]. For each group of instruments, a bounding cooldown/heatup rate was calculated as a 30-minute moving 2400586.301.R0 Page A-2 of A-6
F0306-01R4 average (lines shown as -----). Use of a 30-minute average is conservative compared to the Code-required 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> measurement.As shown in Figure A-2, most regions remained within the specified
<100 °F/hr ramp rate limit. The exceptions to this convention were:
- The RPV bottom head region ( instruments 56-29, 56-30, and 56-31), where a ramp rate of
~180 °F/hr was observed during a period of reheating which began shortly after 08:00 on 4/25.
It is believed that this reheating occurred due to resumption of circulating flow, thus mixing the cold water in the bottom head (from HPCI injection and CRD ingress, presumably <100 °F at the time) with higher -temperature water in the upper vessel (>400°F at the time).
- Instruments 74-32 and 74-43 on the RHR heat exchanger inlet locations experienced a rapid increase and then decrease in temperature at approximately 11:00 on 4/25, presumably in concert with initiation of shutdown cooling. The rate of change during both increase and decrease was in excess of 200 °F/hr. However, given the short duration of the event and proximity to changing conditions, it is not believed that the RPV experienced similar fluid transient conditions. The maximum change in temperature during this period was approximately 140 °F. Assessment of any significance of this event should be addressed separately by the fatigue monitoring program.
Figure A-3 plots all available tags within the bottom head region to compare ramp rates at all locations over time. As shown in the plot, the maximum temperature rate during heatup occurred for instrument 56-31 at approximately 09:00 on 4/25, and remained less than 200 °F/hr. The maximum cooldown temperature rate occurred for instrument 56-8 on the bottom head drain line, and was briefly in excess of 200 °F/hr. However, this rate occurred shortly after the reactor water cleanup (RWCU) system was placed back in service, and is indicative of cold water flowing through a relatively thin-wall pipe as opposed to the RPV. After a short period of significant cooling, this instrument signal matched the trends for other bottom head sensors. Therefore, the maximum heatup/cooldown rate applicable to the RPV was less than 200 °F/hr during the period of interest.
Figure A-4 plots the tags highlighted blue in Figure A-1 to compare measured temperatures by region against the applicable P-T curves. As shown in the plot, the bottom head region exceeded ( ventured left of) the Curve-1 limits. At the furthest-left point, RPV pressure was approximately 950 psig while bottom head temperature was approximately 110°F. However, at that time the ramp rate was relatively low (<10 °F/hr). Curve -1 was first exceeded at approximately 943 psig and 140°F, at which point the ramp rate was approximately 20 °F/hr. The maximum ramp rate experienced by the bottom head during the event was approximately 70 °F/hr, which occurred prior to exceeding the limit curves.
Based on the assessment herein, there were tw o observations of interest during the 4/24 scram and ensuing cooldown, both associated with the bottom head region. These items are summarized below and evaluated in detail elsewhere in this calculation.
- 1. With the RPV still at pressure, metal temperatures decreased such that values exceed the Curve-1 limits from [1 ] (as shown in Figure A-4 ).
- 2. Following depressurization, metal temperatures increased quickly as shown in Figure A-2 (at a rate of approximately 180 °F/hr) presumably in concert with resumption of some sort of circulating flow. These rates exceed the 100 °F/hr limit stated in [1 ].
From the figures herein, the beltline instruments (e.g., 56-17) remained at hot conditions (>500°F) even as the bottom head region (e.g., 56-23 and all lower locations) were cooled. The exact point of the cold/hot interface is not known since the instruments at 56-21 and 56 -22 could not be retrieved.
However, in comparing the instrument locations to the map of fluence locations for the extended beltline region in Figure 11, it appears unlikely that a location in the embrittled zone was subjected to low temperatures.
2400586.301.R0 Page A-3 of A-6
F0306- 01R4 Figure A-2: Temperature Profiles and Rates during Shutdown Period
2400586.301.R0 Page A-4 of A-6
F0306- 01R4 Figure A-3: Temperature Profiles and Ramp Rates for all Bottom Head Instruments 2400586.301.R0 Page A-5 of A-6
F0306- 01R4 Figure A-4: Comparison of RPV Temperature Regions to P-T Limit Curves
2400586.301.R0 Page A-6 of A-6
F0306- 01R4 APPENDIX B:
SUPPORTING FILES
Supporting Files Comment
- 1. 2400586.301.Ramp.R0.xlsx Excel file contains calculations for BFN Unit 1 for 200°F/hr ramp rate.
2400586.301.R0 Page B-1 of B-1
F0306-01R4